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Sample records for high dose neutron

  1. BH3105 type neutron dose equivalent meter of high sensitivity

    International Nuclear Information System (INIS)

    Ji Changsong; Zhang Enshan; Yang Jianfeng; Zhang Hong; Huang Jiling

    1995-10-01

    It is noted that to design a neutron dose meter of high sensitivity is almost impossible in the frame of traditional designing principle--'absorption net principle'. Based on a newly proposed principle of obtaining neutron dose equi-biological effect adjustment--' absorption stick principle', a brand-new neutron dose-equivalent meter with high neutron sensitivity BH3105 has been developed. Its sensitivity reaches 10 cps/(μSv·h -1 ), which is 18∼40 times higher than one of foreign products of the same kind and is 10 4 times higher than that of domestic FJ342 neutron rem-meter. BH3105 has a measurement range from 0.1μSv/h to 1 Sv/h which is 1 or 2 orders wider than that of the other's. It has the advanced properties of gamma-resistance, energy response, orientation, etc. (6 tabs., 5 figs.)

  2. High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert

    2018-04-30

    Project Objective: The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiated microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations

  3. Cation disorder in high-dose, neutron-irradiated spinel

    International Nuclear Information System (INIS)

    Sickafus, K.E.; Larson, A.C.; Yu, N.; Nastasi, M.; Hollenberg, G.W.; Garner, F.A.; Bradt, R.C.

    1994-08-01

    The objective of this effort is to determine whether MgAl 2 O 4 spinel is a suitable ceramic for fusion applications. Here, the crystal structures of MgAl 2 O 4 spinel single crystals irradiated to high neutron fluences [>5·10 26 n/m 2 (E n > 0.1 MeV)] were examined by neutron diffraction. Crystal structure refinement of the highest dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by ∼ 20% while increasing by ∼ 8% on octahedral sites. Since the neutron scattering length for Mg is considerably larger than for Al, this results is consistent with site exchange between Mg 2+ ions on tetrahedral sites and Al 3+ ions on octahedral sites. Least-squares refinements also indicated that, in all irradiated samples, at least 35% of Mg 2+ and Al 3+ ions in the crystal experienced disordering replacements. This retained dpa on the cation sublattices is the largest retained damage ever measured in an irradiated spinel material

  4. High dose neutron irradiation damage in beryllium as blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

    2001-11-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

  5. High dose neutron irradiation damage in beryllium as blanket material

    International Nuclear Information System (INIS)

    Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

    2001-01-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

  6. The neutron dose equivalent around high energy medical electron linear accelerators

    Directory of Open Access Journals (Sweden)

    Poje Marina

    2014-01-01

    Full Text Available The measurement of neutron dose equivalent was made in four dual energy linear accelerator rooms. Two of the rooms were reconstructed after decommissioning of 60Co units, so the main limitation was the space. The measurements were performed by a nuclear track etched detectors LR-115 associated with the converter (radiator that consist of 10B and with the active neutron detector Thermo BIOREM FHT 742. The detectors were set at several locations to evaluate the neutron ambient dose equivalent and/or neutron dose rate to which medical personnel could be exposed. Also, the neutron dose dependence on collimator aperture was analyzed. The obtained neutron dose rates outside the accelerator rooms were several times smaller than the neutron dose rates inside the accelerator rooms. Nevertheless, the measured neutron dose equivalent was not negligible from the aspect of the personal dosimetry with almost 2 mSv a year per person in the areas occupied by staff (conservative estimation. In rooms with 15 MV accelerators, the neutron exposure to the personnel was significantly lower than in the rooms having 18 MV accelerators installed. It was even more pronounced in the room reconstructed after the 60Co decommissioning. This study confirms that shielding from the neutron radiation should be considered when building vaults for high energy linear accelerators, especially when the space constraints exist.

  7. Neutron dose to patients treated with high-energy medical accelerators

    International Nuclear Information System (INIS)

    McGinley, P.H.

    2001-01-01

    The neutron dose equivalent received by patients treated with high energy x-ray beams was measured in this research. A total of 13 different medical accelerators were evaluated in terms of the neutron dose equivalent in the patient plane and at the beam center. The neutron dose equivalent at the beam center was found to ranged from 0.02 to 9.4 mSv per Sv of x-ray dose and values from 0.029 to 2.58 mSv per Sv of x-ray were measured in the patient plane. It was concluded that the neutron levels meet the International Electrotechnical Commission standard for the patient plane. It was also concluded that when intensity modulated radiation treatment is conducted the neutron dose equivalent received by the patient will increase by a factor of 2 to 10. (author)

  8. Dose conversion coefficients for high-energy photons, electrons, neutrons and protons

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2005-01-01

    Dose conversion coefficients for photons, electrons and neutrons based on new ICRP recommendations were cited in the ICRP Publication 74, but the energy ranges of these data were limited and there are no data for high energy radiations produced in accelerator facilities. For the purpose of designing the high intensity proton accelerator facilities at JAERI, the dose evaluation code system of high energy radiations based on the HERMES code was developed and the dose conversion coefficients of effective dose were evaluated for photons, neutrons and protons up to 10 GeV, and electrons up to 100 GeV. The dose conversion coefficients of effective dose equivalent were also evaluated using quality factors to consider the consistency between radiation weighting factors and Q-L relationship. The effective dose conversion coefficients obtained in this work were in good agreement with those recently evaluated by using FLUKA code for photons and electrons with all energies, and neutrons and protons below 500 MeV. There were some discrepancy between two data owing to the difference of cross sections in the nuclear reaction models. The dose conversion coefficients of effective dose equivalents for high energy radiations based on Q-L relation in ICRP Publication 60 were evaluated only in this work. The previous comparison between effective dose and effective dose equivalent made it clear that the radiation weighting factors for high energy neutrons and protons were overestimated and the modification was required. (author)

  9. Dose conversion coefficients for high-energy photons, electrons, neutrons and protons

    CERN Document Server

    Sakamoto, Y; Sato, O; Tanaka, S I; Tsuda, S; Yamaguchi, Y; Yoshizawa, N

    2003-01-01

    In the International Commission on Radiological Protection (ICRP) 1990 Recommendations, radiation weighting factors were introduced in the place of quality factors, the tissue weighting factors were revised, and effective doses and equivalent doses of each tissues and organs were defined as the protection quantities. Dose conversion coefficients for photons, electrons and neutrons based on new ICRP recommendations were cited in the ICRP Publication 74, but the energy ranges of theses data were limited and there are no data for high energy radiations produced in accelerator facilities. For the purpose of designing the high intensity proton accelerator facilities at JAERI, the dose evaluation code system of high energy radiations based on the HERMES code was developed and the dose conversion coefficients of effective dose were evaluated for photons, neutrons and protons up to 10 GeV, and electrons up to 100 GeV. The dose conversion coefficients of effective dose equivalent were also evaluated using quality fact...

  10. Dose levels due to neutrons in the vicinity of high energy medical accelerators

    International Nuclear Information System (INIS)

    McGinley, P.H.; Wood, M.; Sohrabi, M.; Mills, M.; Rodriguez, R.

    1976-01-01

    High energy photons are generated for use in radiation therapy by the decelleration of electrons in metal targets. Fast neutrons are also generated as a result of (γ, n) and (e, e'n) interactions in the target, beam compensator filter, and collimator material. In this work the adsorbed dose to neutrons was measured at the center of a 10 x 10 cm photon beam and 5 cm outside of the beam edge for a number of treatment units. Dose levels due to slow and fast neutrons were also established outside of the treatment rooms and a Bonner sphere neutron spectrometer system was employed to determine the neutron energy spectrum due to stray neutron radiation at each accelerator. For the linac it was found that the neutron dose at the beam center was 0.0039% of the photon dose and values of 0.049% and 0.053% were observed for the Allis Chalmers betatron and the Brown Boveri Betatron. Dose equivalent rates in the range of 0.3 to 22.5 mrem/hr were measured for points outside the treatment rooms when the accelerators were operated at a photon dose rate of 100 rad/min at the treatment position

  11. Monitor units are not predictive of neutron dose for high-energy IMRT

    Directory of Open Access Journals (Sweden)

    Hälg Roger A

    2012-08-01

    Full Text Available Abstract Background Due to the substantial increase in beam-on time of high energy intensity-modulated radiotherapy (>10 MV techniques to deliver the same target dose compared to conventional treatment techniques, an increased dose of scatter radiation, including neutrons, is delivered to the patient. As a consequence, an increase in second malignancies may be expected in the future with the application of intensity-modulated radiotherapy. It is commonly assumed that the neutron dose equivalent scales with the number of monitor units. Methods Measurements of neutron dose equivalent were performed for an open and an intensity-modulated field at four positions: inside and outside of the treatment field at 0.2 cm and 15 cm depth, respectively. Results It was shown that the neutron dose equivalent, which a patient receives during an intensity-modulated radiotherapy treatment, does not scale with the ratio of applied monitor units relative to an open field irradiation. Outside the treatment volume at larger depth 35% less neutron dose equivalent is delivered than expected. Conclusions The predicted increase of second cancer induction rates from intensity-modulated treatment techniques can be overestimated when the neutron dose is simply scaled with monitor units.

  12. Development of Real-Time Measurement of Effective Dose for High Dose Rate Neutron Fields

    International Nuclear Information System (INIS)

    Braby, L. A.; Reece, W. D.; Hsu, W. H.

    2003-01-01

    Studies of the effects of low doses of ionizing radiation require sources of radiation which are well characterized in terms of the dose and the quality of the radiation. One of the best measures of the quality of neutron irradiation is the dose mean lineal energy. At very low dose rates this can be determined by measuring individual energy deposition events, and calculating the dose mean of the event size. However, at the dose rates that are normally required for biology experiments, the individual events can not be separated by radiation detectors. However, the total energy deposited in a specified time interval can be measured. This total energy has a random variation which depends on the size of the individual events, so the dose mean lineal energy can be calculated from the variance of repeated measurements of the energy deposited in a fixed time. We have developed a specialized charge integration circuit for the measurement of the charge produced in a small ion chamber in typical neutron irradiation experiments. We have also developed 4.3 mm diameter ion chambers with both tissue equivalent and carbon walls for the purpose of measuring dose mean lineal energy due to all radiations and due to all radiations except neutrons, respectively. By adjusting the gas pressure in the ion chamber, it can be made to simulate tissue volumes from a few nanometers to a few millimeters in diameter. The charge is integrated for 0.1 seconds, and the resulting pulse height is recorded by a multi channel analyzer. The system has been used in a variety of photon and neutron radiation fields, and measured values of dose and dose mean lineal energy are consistent with values extrapolated from measurements made by other techniques at much lower dose rates. It is expected that this technique will prove to be much more reliable than extrapolations from measurements made at low dose rates because these low dose rate exposures generally do not accurately reproduce the attenuation and

  13. Neutron dose measurements with the GSI ball at high energy accelerators

    International Nuclear Information System (INIS)

    Fehrenbacher, G.; Gutermuth, F.; Radon, T.; Kozlova, E.

    2005-01-01

    Full text: At high energy particle accelerators the production of neutron radiation dominates radiation protection. For the radiation survey at accelerators there is a need for reliable detection systems (passive radiation monitors), which can measure the dose for a wide range of neutron energies independently on the beam pulse structure of the produced radiation. In this work a passive neutron dosemeter for the measurement of the ambient dose equivalent is presented. The dosemeter is suitable for measurements of the emerging neutron radiation at accelerators for the whole energy range up to about 10 GeV. The dosemeter consists of a polyethylene sphere, TL elements (pairs of TLD600/700) and an additional lead layer (PE/Pb) in neutron fields at high energy accelerators is investigated in this work. Results of dose measurements which were performed in realistic neutron fields at the high energy accelerator SPS at CERN (CERF facility) and in Cave A at the heavy ion synchrotron SIS at GSI are presented. The results of these measurements are compared with the expected dose values from the neutron spectra determined for the measurement positions at CERF and in Cave A (FLUKA) and with the dosemeter response derived by the calculated response functions (FLUKA) folded with the neutron spectra. The comparisons show that the additional lead layer in the PE/Pb-sphere improves significantly the response of the dosemeter. The response of the PE/Pb-sphere is 40 to 50 % higher at CERF and Cave A in comparison to the bare PE-sphere. At CERF the dose values of the PE/Pb-sphere is about 25 % lower than the expected dose value, whilst for Cave A, a rather good agreement was found (2 % deviation). (author)

  14. Development of Real-Time Measurement of Effective Dose for High Dose Rate Neutron Fields

    CERN Document Server

    Braby, L A; Reece, W D

    2003-01-01

    Studies of the effects of low doses of ionizing radiation require sources of radiation which are well characterized in terms of the dose and the quality of the radiation. One of the best measures of the quality of neutron irradiation is the dose mean lineal energy. At very low dose rates this can be determined by measuring individual energy deposition events, and calculating the dose mean of the event size. However, at the dose rates that are normally required for biology experiments, the individual events can not be separated by radiation detectors. However, the total energy deposited in a specified time interval can be measured. This total energy has a random variation which depends on the size of the individual events, so the dose mean lineal energy can be calculated from the variance of repeated measurements of the energy deposited in a fixed time. We have developed a specialized charge integration circuit for the measurement of the charge produced in a small ion chamber in typical neutron irradiation exp...

  15. Determination of dose components in mixed gamma neutron fields by use of high pressure ionization chambers

    International Nuclear Information System (INIS)

    Golnik, N.; Pliszczynski, T.; Wysocka, A.; Zielczynski, M.

    1985-01-01

    The two ionization chamber method for determination of dose components in mixed γ-neutron field has been improved by increasing gas pressure in the chambers up to some milions pascals. Advantages of high pressure gas filling are the followings: 1) significant reduction of the ratio of neutron-to gamma sensitivity for the hydrogen-free chamber, 2) possibility of sensitivity correction for both chambers by application of appropriate voltage, 3) high sensitivity for small detectors. High-pressure, pen-like ionization chambers have been examined in fields of different neutron sources: a TE-chamber, filled with 0.2 MPa of quasi-TE-gas and a conductive PTFE chamber, filled with 3.1 MPa of CO 2 . The ratio of neutron-to-gamma sensitivity for the PTFE chamber, operated at electrical field strength below 100 V/cm, has not exceeded 0.01 for neutrons with energy below 8 MeV. Formula is presented for calculation of this ratio for any high-pressure, CO 2 -filled ionization chamber. Contribution of gamma component to total tissue dose in the field of typical neutron sources has been found to be 3 to 70%

  16. High dose effect of gamma and neutrons on the N-JFET electronic components

    International Nuclear Information System (INIS)

    Assaf, Jamal-Eddin

    2006-11-01

    Two types of N-JFET components have been irradiated by high doses of thermal neutrons and gamma rays up to 2000x10 12 n/cm 2 and 1000 kGy, respectively. The static tests show a decrease of the g m and I d s parameters. The behaviour of electronic noise on the output was the principal dynamic test after irradiation. The result of this test gives an increase of the noise with radiation dose increasing. The noise was described as the Equivalent Noise of Charge (ENC) at the output of the measurements set-up. The quantities and the qualities of the noise depend on the N-JEET type and the type of radiation (neutrons or gamma). Other tests were carried out like the relaxation or recovery phenomena after radiation, and the superposed effects of gamma and neutrons.(author)

  17. Dose determination of Neutron contamination in radiothrapy rooms equiped with high energy linear accelerators

    International Nuclear Information System (INIS)

    Shweikani, R.; Anjak, O.

    2014-03-01

    Radiotherapy represents the most widely spread technique to control and treat cancer. To increase the treatment efficiency, high-energy linear accelerators are used. However, applying high energy photon beams leads to a non-negligible dose of neutrons contaminating therapeutic beams. A high-energy (23 MV) linear accelerator (Varian 21EX) was studied. The CR-39 nuclear track detectors (NTDs) were used to study the variation of fast neutron relative intensities around a linear accelerator high energy photon beam and to determined the its variation on the patient plane at 0, 50, 100, 150 and 200 cm from the center of the photon beam was. By increasing the distance from the center of the X-ray beam towards the periphery, the photoneutron dose equivalent decreased rapidly for the fields. Photoneutron intensity and distributions at isocenter level with the field sizes of 40*40 cm'2 at SSD=100cm around 23 MV photon beam using Nuclear Track Detectors were determined. The advantages of CR-39 NTD s over active detectors: 1- there is no pulse pileup problem. 2- no photon interference with neutron measurement. 3- no electronics are required. 4 - less prone to noise and interference. The photoneutron intensities were rapidly decreased as we move away from the isocenter of linear accelerators. As the use of simulation software MCNP match in the results we have obtained through direct measurements and the modeling results using the code MCNP (author).

  18. Effects of high neutron doses and duration of the chemical etching on the optical properties of CR-39

    International Nuclear Information System (INIS)

    Sahoo, G.S.; Tripathy, S.P.; Paul, S.; Sharma, S.C.; Joshi, D.S.; Gupta, A.K.; Bandyopadhyay, T.

    2015-01-01

    Effects of the duration of chemical etching on the transmittance, absorbance and optical band gap width of the CR-39 (Polyallyl diglycol carbonate) detectors irradiated to high neutron doses (12.7, 22.1, 36.0 and 43.5 Sv) were studied. The neutrons were produced by bombardment of a thick Be target with 12 MeV protons of different fluences. The unirradiated and neutron-irradiated CR-39 detectors were subjected to a stepwise chemical etching at 1 h intervals. After each step, the transmission spectra of the detectors were recorded in the range from 200 to 900 nm, and the absorbances and optical band gap widths were determined. The effect of the etching on the light transmittance of unirradiated detectors was insignificant, whereas it was very significant in the case of the irradiated detectors. The dependence of the optical absorbance on the neutron dose is linear at short etching periods, but exponential at longer ones. The optical band gap narrows with increasing etching time. It is more significant for the irradiated dosimeters than for the unirradiated ones. The rate of the narrowing of the optical band gap with increasing neutron dose increases with increasing duration of the etching. - Highlights: • The variation of optical properties of CR-39 at very high neutron dose is analyzed. Etching process is found to play a crucial role for change in optical properties of neutron-irradiated CR-39. • The optical absorbance varies linearly at lower dose, at very high dose absorbance saturation occurs. The dose at which saturation absorbance is observed shifts towards lower neutron dose with increase in etching time. • The rate of decrease in optical band gap with respect to neutron dose is found to be more at higher etching durations

  19. Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Saleh, Tarik A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Eftink, Benjamin P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, the synergies between ' and fine-scale and moderate-scale cavity formation is investigated.

  20. High-Dose Neutron Detector Development Using 10B Coated Cells

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzlova, Daniela [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-08

    During FY16 the boron-lined parallel-plate technology was optimized to fully benefit from its fast timing characteristics in order to enhance its high count rate capability. To facilitate high count rate capability, a novel fast amplifier with timing and operating properties matched to the detector characteristics was developed and implemented in the 8” boron plate detector that was purchased from PDT. Each of the 6 sealed-cells was connected to a fast amplifier with corresponding List mode readout from each amplifier. The FY16 work focused on improvements in the boron-10 coating materials and procedures at PDT to significantly improve the neutron detection efficiency. An improvement in the efficiency of a factor of 1.5 was achieved without increasing the metal backing area for the boron coating. This improvement has allowed us to operate the detector in gamma-ray backgrounds that are four orders of magnitude higher than was previously possible while maintaining a relatively high counting efficiency for neutrons. This improvement in the gamma-ray rejection is a key factor in the development of the high dose neutron detector.

  1. In vivo transcriptome modulation after low dose of high energy neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Amendola, R; Fratini, E; Piscitelli, M; Sallustio, D E [ENEA, BAS BIOTEC MED, Roma (Italy); Angelone, M; Pillon, M [ENEA, FUS TEC, Frascati (Italy); Chiani, F; Licursi, V; Negri, R [Universita La Sapienza, Roma (Italy). Dip. Biologia Cellulare e dello Sviluppo

    2007-07-01

    Complete text of publication follows. Objective: This project aims to the identification of an hypothetical transcriptome modulation of mouse peripheral blood lymphocytes and skin after exposure to high energy neutron in vivo. Positive candidate genes isolated from mice in in vivo experiments will be selected and evaluated for both radioprotection issues dealing with cosmic ray exposure, and for biomedical issues mainly for low doses and non-cancer effects. Methods: High energy neutron irradiation is performed at the ENEA Frascati, neutron generator facilities (FNG), specifically dedicated to biological samples. FNG is a linear electrostatic accelerator that produces up to 1.0 x 10{sup 11} n/s 14 MeV neutrons via the D-T nuclear reaction. The dose-rate applied for this study is of 0.7 cGy/min. The functional genomic approach has been performed on six animals for each experimental points: un-irradiated; 20 cGy, 6 hours and 24 hours delayed time after exposure. Preliminarily, a pool of total RNA is evaluated on commercial micro-arrays containing large collections of mus musculus cDNAs. Statistical filtering and functional clustering of the data is carried out using dedicated software packages. Results: Candidate genes are selected on the basis of responsiveness to 20 cGy of exposure, with a defined temporal regulation. We plan to organize a systematic screen focused on genes responding to our selection criteria, in in vivo mouse experiments, and correlate their differential expression to the human counterparts. A specific cross species database will be created with all the functional information available in standardized format (MIAME: minimal information about micro-arrays experiments). Conclusions: A lack of information on in vivo experiments is still evident for low doses exposure, especially for neutron of cosmic interest. Individual susceptibility, extensive number of animals to be processed, lack of standardization methodologies are among problems to be solved

  2. Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

    Science.gov (United States)

    Jiao, Z.; Michalicka, J.; Was, G. S.

    2018-04-01

    Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  4. An investigation of methods for neutron dose measurement in high temperature irradiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Kosako, Toshisou; Sugiura, Nobuyuki [Tokyo Univ. (Japan); Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan)] [and others

    2000-10-01

    The Japan Atomic Energy Research Institute (JAERI) has been conducting the innovative basic research on high temperature since 1994, which is a series of high temperature irradiation studies using the High Temperature Engineering Test Reactor (HTTR). 'The Task Group for Evaluation of Irradiation Dose under High Temperature Radiation' was founded in the HTTR Utilization Research Committee, which is the promoting body of the innovative basic research. The present report is a summary of investigation which has been made by the Task Group on the present status and subjects of research and development of neutron detectors in high temperature irradiation fields, in view of contributing to high temperature irradiation research using the HTTR. Detectors investigated here in the domestic survey are the following five kinds of in-core detectors: 1) small fission counter, 2) small fission chamber, 3) self-powered detector, 4) activation detector, and 5) optical fiber. In addition, the research and development status in Russia has been investigated. The present report will also be useful as nuclear instrumentation of high temperature gas-cooled reactors. (author)

  5. Desorption of tritium and helium from high dose neutron irradiated beryllium

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

    2007-08-01

    The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

  6. 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

  7. Thermal conductivity of beryllium under low temperature high dose neutron irradiation

    International Nuclear Information System (INIS)

    Chakin, V.P.; Latypov, R.N.; Suslov, D.N.; Kupriyanov, I.B.

    2004-01-01

    Thermal conductivity of compact beryllium of several Russian grades such as TE-400, TE-56, TE-30, TIP and DIP differing in the production technology, grain size and impurity content has been investigated. The thermal diffusivity of beryllium was measured on the disks in the initial and irradiated conditions using the pulse method in the range from room temperature to 200degC. The thermal conductivity was calculated using the table values for the beryllium thermal capacity. The specimens and beryllium neutron source fragments were irradiation in the SM reactor at 70degC and 200degC to a neutron fluence of (0.5-11.4)·10 22 cm -2 (E>0.1 MeV) and in the BOR-60 reactor at 400degC to 16·10 22 cm -2 (E>0.1MeV), respectively. The low-temperature irradiation leads to the drop decrease of the beryllium thermal conductivity and the effect depends on the irradiation parameters. The paper analyses the effect of irradiation parameters (temperature, neutron fluence), measurement temperature and structural factors on beryllium conductivity. The experiments have revealed that the short time post-irradiation annealing at high temperature results in partial reduction of the thermal conductivity of irradiated beryllium. (author)

  8. Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

    International Nuclear Information System (INIS)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

    2007-01-01

    Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to ∼ 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

  9. 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 ...

  10. Characterization of the Microstructure in Recrystallized Zircaloy-2 Cladding Irradiated to a High Neutron Dose

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2003-04-01

    The objectives of the present project were to determine if there is anything in the microstructure of highly irradiated Zircaloy-2 which may make the material fracture in a brittle manner. Samples were taken from three different locations on a fuel rod which had been irradiated for 12 years. The displacement doses were estimated to be 1.4, 9 and 28 dpa. Specimens for electron microscopy were prepared with two different orientation called axial and radial. In the axial orientation the electron beam goes parallel with the basal plane and diffraction conditions can be arranged so that dislocations with a Burgers' vectors become invisible. In the low dose specimen only a-component damage was present and all second phase particles were crystalline. In both the high and intermediate dose samples there was c-component damage present with a slightly higher amount in the high dose sample. The particles of the Zr(Cr,Fe) 2 type were generally amorphous in these samples and the Fe-content of the particles was highly reduced. The hydride structures were similar in all samples. The hydrides were often precipitated in parallel in the same grain and chains of hydrides were seen which ran from grain to grain. No population of small hydrides were observed except from surface hydrides formed during specimen preparation. It was concluded from the investigation that there is nothing in the microstructure which may make the material in the high dose state subject to a purely mechanically induced fast brittle cracking

  11. High energy neutron radiography

    International Nuclear Information System (INIS)

    Gavron, A.; Morley, K.; Morris, C.; Seestrom, S.; Ullmann, J.; Yates, G.; Zumbro, J.

    1996-01-01

    High-energy spallation neutron sources are now being considered in the US and elsewhere as a replacement for neutron beams produced by reactors. High-energy and high intensity neutron beams, produced by unmoderated spallation sources, open potential new vistas of neutron radiography. The authors discuss the basic advantages and disadvantages of high-energy neutron radiography, and consider some experimental results obtained at the Weapons Neutron Research (WNR) facility at Los Alamos

  12. High dose effects in neutron irradiated face-centered cubic metals

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors

  13. High energy neutron dosimeter

    International Nuclear Information System (INIS)

    Rai, K.S.F.

    1994-01-01

    A device for measuring dose equivalents in neutron radiation fields is described. The device includes nested symmetrical hemispheres (forming spheres) of different neutron moderating materials that allow the measurement of dose equivalents from 0.025 eV to past 1 GeV. The layers of moderating material surround a spherical neutron counter. The neutron counter is connected by an electrical cable to an electrical sensing means which interprets the signal from the neutron counter in the center of the moderating spheres. The spherical shape of the device allows for accurate measurement of dose equivalents regardless of its positioning. 2 figures

  14. The impact of ICRP 60 recommendations on the dose equivalent in low- and high energy neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Jakes, J; Schraube, H [GSF-Forschungszentrum Neuberg, D-85758 Oberschleissheim (Germany). Inst. fuer Strahlenschutz

    1996-12-31

    The objectives of this study was to determine the impact of the increased risk factors for neutrons after ICRP 60 on the operational dose equivalent quantities at a few neutron fields selected with the respect to cover the broad variety of neutron spectra: (1) Cadarache calibration assembly, with average neutron energy around 0.6 MeV, designed to simulate realistic neutron spectra at workplaces. This assembly is basically composed of an almost spherical {sup 238}U converter irradiated by 14.6 MeV neutrons from an accelerator target, placed at its center, and a scattering chamber consisting of a cylindrical polyethylene duct and a series of additional shieldings; (2) Neutron spectra at exposed workplaces in nuclear power plants; (3) Moderated spectra of {sup 252}Cf fission source; (4) Neutron spectra behind a shielding made of the iron (the average energy 5.,89 MeV) and concrete (the average energy 46.51 MeV), respectively; (5) Cosmic rays induced neutron spectra measured on the top of the Zugspitze (2968 m) where there is the average neutron energy around 40 MeV. From the derived neutron spectra, the mean quality factors and conversion factors h after ICRP 21 and ICRP 60, respectively, were calculated. The dose equivalent conversion factors were taken for the region below 20 MeV, and the energy region above 20 MeV. The results show that the operational quantities were affected predominately in the low energy fields, where the changes are given by a factor of 1,3 for the neutron fields given above. As has been expected, the impact of the new recommendations depends on the shape of the neutron spectra. Therefore, this factor can be much higher in the fields where the intermediate energy region is dominant, which is the case of moderated and scattered spectra at some places in the nuclear power plant and around containers with the spent fuel elements. (J.K.) 9 refs.

  15. Presentation of a semiempirical method for the calculation of doses due to neutrons and capture gamma rays inside high energy accelerators rooms

    International Nuclear Information System (INIS)

    Larcher, A.M.; Bonet Duran, S.M.

    1998-01-01

    Full text: Medical electron accelerators operating above 10 MeV produce radiation beams that are contaminated with neutrons. Therefore, shielding design for high energy accelerator rooms must consider the neutron component of the radiation field. In this paper a semiempirical method is presented to calculate doses due to neutrons and capture gamma rays inside the room and the maze. The calculation method is based on the knowledge of the neutron yield Q (neutrons/Gy of photons at isocenter) and the average energy of the primary beam of neutrons Eo (MeV). The method constitutes an appropriate tool for shielding facilities evaluation. The accuracy of the method has been contrasted with data obtained from the literature and an excellent correlation among the calculations and the measured values was achieved. In addition, the method has been used in the verification of experimental data corresponding to a 15 MeV linear accelerator installed in the country with similar results. (author) [es

  16. Equivalent-spherical-shield neutron dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.

    1988-01-01

    Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab

  17. Neutron Dose Measurement Using a Cubic Moderator

    International Nuclear Information System (INIS)

    Sheinfeld, M.; Mazor, T.; Cohen, Y.; Kadmon, Y.; Orion, I.

    2014-01-01

    The Bonner Sphere Spectrometer (BSS), introduced In July 1960 by a research group from Rice University, Texas, is a major approach to neutron spectrum estimation. The BSS, also known as multi-sphere spectrometer, consists of a set of a different diameters polyethylene spheres, carrying a small LiI(Eu) scintillator in their center. What makes this spectrometry method such widely used, is its almost isotropic response, covering an extraordinary wide range of energies, from thermal up to even hundreds of MeVs. One of the most interesting and useful consequences of the above study is the 12'' sphere characteristics, as it turned out that the response curve of its energy dependence, have a similar shape compared with the neutron's dose equivalent as a function of energy. This inexplicable and happy circumstance makes it virtually the only monitoring device capable providing realistic neutron dose estimates over such a wide energy range. However, since the detection mechanism is not strictly related to radiation dose, one can expect substantial errors when applied to widely different source conditions. Although the original design of the BSS included a small 4mmx4mmO 6LiI(Eu) scintillator, other thermal neutron detectors has been used over the years: track detectors, activation foils, BF3 filled proportional counters, etc. In this study we chose a Boron loaded scintillator, EJ-254, as the thermal neutron detector. The neutron capture reaction on the boron has a Q value of 2.78 MeV of which 2.34 MeV is shared by the alpha and lithium particles. The high manufacturing costs, the encasement issue, the installation efficiency and the fabrication complexity, led us to the idea of replacing the sphere with a cubic moderator. This article describes the considerations, as well as the Monte-Carlo simulations done in order to examine the applicability of this idea

  18. NEUTRON AND PHOTON DOSE MAPPING OF A DD NEUTRON GENERATOR.

    Science.gov (United States)

    Metwally, Walid A; Taqatqa, Osama A; Ballaith, Mohammed M; Chen, Allan X; Piestrup, Melvin A

    2017-11-01

    Neutron generators are an excellent tool that can be effectively utilized in educational institutions for applications such as neutron activation analysis, neutron radiography, and profiling and irradiation effects. For safety purposes, it is imperative that appropriate measures are taken in order to minimize the radiation dose from such devices to the operators, students and the public. This work presents the simulation and measurement results for the neutron and photon dose rates in the vicinity of the neutron generator installed at the University of Sharjah. A very good agreement is found between the simulated and measured dose rates. All of the public dose constraints were found to be met. The occupational dose constraint was also met after imposing a 200 cm no entry zone around the generator room. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  19. A neutron dose equivalent meter at CAEP

    International Nuclear Information System (INIS)

    Tian Shihai; Lu Yan; Wang Heyi; Yuan Yonggang; Chen Xu

    2012-01-01

    The measurement of neutron dose equivalent has been a widespread need in industry and research. In this paper, aimed at improving the accuracy of neutron dose equivalent meter: a neutron dose counter is simulated with MCNP5, and the energy response curve is optimized. The results show that the energy response factor is from 0.2 to 1.8 for neutrons in the energy range of 2.53×10 -8 MeV to 10 MeV Compared with other related meters, it turns that the design of this meter is right. (authors)

  20. Induction and disappearance of G2 chromatid breaks in lymphocytes after low doses of low LET γ - rays and high LET fast neutrons

    International Nuclear Information System (INIS)

    Vral, Anne; Thierens, Hubert; Baeyens, Ans; De Ridder, Leo

    2001-01-01

    In view of the potential importance of the G2 assay for detecting chromosomal radiosensitivity and possible predisposition to cancer the need to elucidate the mechanism underlying the formation of chromatid breaks, observed with the G2 assay after low dose irradiation, has been recognised. In this study we irradiated blood samples of 4 healthy donors with low LET γ-rays and high LET neutrons, which initially produce the same number of dsb but of a different quality. By means of the G2 assay, we determined the number of chromatid breaks induced by γ-rays and neutrons and compared the kinetics of chromatid break rejoining for radiations of different quality. In a first set of experiments a dose-response curve for the formation of chromatid breaks was carried out for γ-rays and neutrons with doses ranging between 0.1 and 0.5 Gy. In a second set of experiments the kinetics of chromatid break formation and disappearance was investigated after a dose of 0.5 Gy using post-irradiation times ranging between 0.5 h and 3.5 h. For the highest dose of 0.5 Gy the number of isochromatid breaks were also scored. No significant differences in the number of chromatid breaks were observed between low LET γ-rays and high LET neutrons for the 4 donors at any of the doses given. The dose response curves for the formation of chromatid breaks are linear for both radiation qualities and RBE values equal to one were obtained. Scoring of isochromatid breaks at the highest dose of 0.5 Gy revealed that high LET neutrons are however more effective at inducing isochromatid breaks (RBE of 6.2). The rejoining experiments further showed that the kinetics of disappearance of chromatid breaks following irradiation with low LET γ-rays or high-LET neutrons are not significantly different. T 1/2 0.92 h for γ-rays and t 1/2 = 0.84 h for neutrons were obtained. In conclusion, our results show that at low doses of radiation the induction as well as the disappearance of G2 chromatid breaks is LET

  1. Dose prescription in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Gupta, N.M.S.; Gahbauer, R.A.; Blue, T.E.; Wambersie, A.

    1994-01-01

    The purpose of this paper is to address some aspects of the many considerations that need to go into a dose prescription in boron neutron capture therapy (BNCT) for brain tumors; and to describe some methods to incorporate knowledge from animal studies and other experiments into the process of dose prescription. Previously, an algorithm to estimate the normal tissue tolerance to mixed high and low linear energy transfer radiations in BNCT was proposed. The authors have developed mathematical formulations and computational methods to represent this algorithm. Generalized models to fit the central axis dose rate components for an epithermal neutron field were also developed. These formulations and beam fitting models were programmed into spreadsheets to simulate two treatment techniques which are expected to be used in BNCT: a two-field bilateral scheme and a single-field treatment scheme. Parameters in these spreadsheets can be varied to represent the fractionation scheme used, the 10 B microdistribution in normal tissue, and the ratio of 10 B in tumor to normal tissue. Most of these factors have to be determined for a given neutron field and 10 B compound combination from large animal studies. The spreadsheets have been programmed to integrate all of the treatment-related information and calculate the location along the central axis where the normal tissue tolerance is exceeded first. This information is then used to compute the maximum treatment time allowable and the maximum tumor dose that may be delivered for a given BNCT treatment. The effect of different treatment variables on the treatment time and tumor dose has been shown to be very significant. It has also been shown that the location of D max shifts significantly, depending on some of the treatment variables-mainly the fractionation scheme used. These results further emphasize the fact that dose prescription in BNCT is very complicated and nonintuitive. 11 refs., 6 figs., 3 tabs

  2. Neutron absorbed dose in a pacemaker CMOS

    International Nuclear Information System (INIS)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L.

    2012-01-01

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10 -17 Gy per neutron emitted by the source. (Author)

  3. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, 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); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-06-15

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

  4. Cosmic radiation dose in aircraft - a neutron track etch detector

    Energy Technology Data Exchange (ETDEWEB)

    Vukovic, B.; Radolic, V.; Miklavcic, I.; Poje, M.; Varga, M. [Department of Physics, University of Osijek, 31000 Osijek, P.O. Box 125, Gajev trg 6 (Croatia); Planinic, J. [Department of Physics, University of Osijek, 31000 Osijek, P.O. Box 125, Gajev trg 6 (Croatia)], E-mail: planinic@ffos.hr

    2007-12-15

    Cosmic radiation bombards us at high altitude by ionizing particles. The radiation environment is a complex mixture of charged particles of solar and galactic origin, as well as of secondary particles produced in interaction of the galactic cosmic particles with the nuclei of atmosphere of the Earth. The radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. The non-neutron component of cosmic radiation dose aboard ATR 42 and A 320 aircrafts (flight level of 8 and 11 km, respectively) was measured with TLD-100 (LiF:Mg,Ti) detectors and the Mini 6100 semiconductor dosimeter. The estimated occupational effective dose for the aircraft crew (A 320) working 500 h per year was 1.64 mSv. Other experiments, or dose rate measurements with the neutron dosimeter, consisting of LR-115 track detector and boron foil BN-1 or 10B converter, were performed on five intercontinental flights. Comparison of the dose rates of the non-neutron component (low LET) and the neutron one (high LET) of the radiation field at the aircraft flight level showed that the neutron component carried about 50% of the total dose. The dose rate measurements on the flights from the Middle Europe to the South and Middle America, then to Korea and Japan, showed that the flights over or near the equator region carried less dose rate; this was in accordance with the known geomagnetic latitude effect.

  5. Cosmic radiation dose in aircraft - a neutron track etch detector

    International Nuclear Information System (INIS)

    Vukovic, B.; Radolic, V.; Miklavcic, I.; Poje, M.; Varga, M.; Planinic, J.

    2007-01-01

    Cosmic radiation bombards us at high altitude by ionizing particles. The radiation environment is a complex mixture of charged particles of solar and galactic origin, as well as of secondary particles produced in interaction of the galactic cosmic particles with the nuclei of atmosphere of the Earth. The radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. The non-neutron component of cosmic radiation dose aboard ATR 42 and A 320 aircrafts (flight level of 8 and 11 km, respectively) was measured with TLD-100 (LiF:Mg,Ti) detectors and the Mini 6100 semiconductor dosimeter. The estimated occupational effective dose for the aircraft crew (A 320) working 500 h per year was 1.64 mSv. Other experiments, or dose rate measurements with the neutron dosimeter, consisting of LR-115 track detector and boron foil BN-1 or 10B converter, were performed on five intercontinental flights. Comparison of the dose rates of the non-neutron component (low LET) and the neutron one (high LET) of the radiation field at the aircraft flight level showed that the neutron component carried about 50% of the total dose. The dose rate measurements on the flights from the Middle Europe to the South and Middle America, then to Korea and Japan, showed that the flights over or near the equator region carried less dose rate; this was in accordance with the known geomagnetic latitude effect

  6. 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...

  7. Neutron dose and energy spectra measurements at Savannah River Plant

    International Nuclear Information System (INIS)

    Brackenbush, L.W.; Soldat, K.L.; Haggard, D.L.; Faust, L.G.; Tomeraasen, P.L.

    1987-08-01

    Because some workers have a high potential for significant neutron exposure, the Savannah River Plant (SRP) contracted with Pacific Northwest Laboratory (PNL) to verify the accuracy of neutron dosimetry at the plant. Energy spectrum and neutron dose measurements were made at the SRP calibrations laboratory and at several other locations. The energy spectra measurements were made using multisphere or Bonner sphere spectrometers, 3 He spectrometers, and NE-213 liquid scintillator spectrometers. Neutron dose equivalent determinations were made using these instruments and others specifically designed to determine dose equivalent, such as the tissue equivalent proportional counter (TEPC). Survey instruments, such as the Eberline PNR-4, and the thermoluminescent dosimeter (TLD)-albedo and track etch dosimeters (TEDs) were also used. The TEPC, subjectively judged to provide the most accurate estimation of true dose equivalent, was used as the reference for comparison with other devices. 29 refs., 43 figs., 13 tabs

  8. On the mechanisms of action of high and superhigh doses of γ-quanta and neutrons of the central nervous system

    International Nuclear Information System (INIS)

    Lavrova, G.A.; Pushkareva, T.V.; Nikanorova, N.G.; Sverdlov, A.G.

    1984-01-01

    Wistar rats were exposed to γ-quanta ( 60 Co, 100-400 Gy) and neutrons (100-200 Gy). The C. N. S. syndrome was shown to develop the neuron lesions being relatively minor. With a reference to this index, the RBE of neutrons was 1.75. The permeability of blood vessels of the brain was decreased during the development of the C. N. S. -syndrome. Disturbances in the functional state of neurons and interneuron reactions, including the changes in the mediator metabolism (these changes were manifested by the increased GABA content of the cerebellum), play, in all appearance, an important role in the impairment of the brain function under the effect of high doses of neutrons and γ-quanta

  9. Experimental evaluation of neutron dose in radiotherapy patients: Which dose?

    Energy Technology Data Exchange (ETDEWEB)

    Romero-Expósito, M., E-mail: mariateresa.romero@uab.cat; Domingo, C.; Ortega-Gelabert, O.; Gallego, S. [Grup de Recerca en Radiacions Ionizants (GRRI), Departament de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Sánchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009 (Spain); Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41009 (Spain)

    2016-01-15

    Purpose: The evaluation of peripheral dose has become a relevant issue recently, in particular, the contribution of secondary neutrons. However, after the revision of the Recommendations of the International Commission on Radiological Protection, there has been a lack of experimental procedure for its evaluation. Specifically, the problem comes from the replacement of organ dose equivalent by the organ-equivalent dose, being the latter “immeasurable” by definition. Therefore, dose equivalent has to be still used although it needs the calculation of the radiation quality factor Q, which depends on the unrestricted linear energy transfer, for the specific neutron irradiation conditions. On the other hand, equivalent dose is computed through the radiation weighting factor w{sub R}, which can be easily calculated using the continuous function provided by the recommendations. The aim of the paper is to compare the dose equivalent evaluated following the definition, that is, using Q, with the values obtained by replacing the quality factor with w{sub R}. Methods: Dose equivalents were estimated in selected points inside a phantom. Two types of medical environments were chosen for the irradiations: a photon- and a proton-therapy facility. For the estimation of dose equivalent, a poly-allyl-diglicol-carbonate-based neutron dosimeter was used for neutron fluence measurements and, additionally, Monte Carlo simulations were performed to obtain the energy spectrum of the fluence in each point. Results: The main contribution to dose equivalent comes from neutrons with energy higher than 0.1 MeV, even when they represent the smallest contribution in fluence. For this range of energy, the radiation quality factor and the radiation weighting factor are approximately equal. Then, dose equivalents evaluated using both factors are compatible, with differences below 12%. Conclusions: Quality factor can be replaced by the radiation weighting factor in the evaluation of dose

  10. Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

    Energy Technology Data Exchange (ETDEWEB)

    Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

    2002-07-01

    Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

  11. Low doses of neutrons induce changes in gene expression

    International Nuclear Information System (INIS)

    Woloschak, G.E.; Chang-Liu, C.M.; Panozzo, J.; Libertin, C.R.

    1993-01-01

    Studies were designed to identify genes induced following low-dose neutron but not following γ-ray exposure in fibroblasts. Our past work had shown differences in the expression of β-protein kinase C and c-fos genes, both being induced following γ-ray but not neutron exposure. We have identified two genes that are induced following neutron, but not γ-ray, exposure: Rp-8 (a gene induced by apoptosis) and the long terminal repeat (LTR) of the human immunodeficiency (HIV). Rp-8 mRNA induction was demonstrated in Syrian hamster embryo fibroblasts and was found to be induced in cells exposed to neutrons administered at low (0.5 cGy/min) and at high dose rate (12 cGy/min). The induction of transcription from the LTR of HIV was demonstrated in HeLa cells bearing a transfected construct of the chloramphenicol acetyl transferase (CAT) gene driven by the HIV-LTR promoter. Measures of CAT activity and CAT transcripts following irradiation demonstrated an unresponsiveness to γ rays over a broad range of doses. Twofold induction of the HIV-LTR was detected following neutron exposure (48 cGy) administered at low (0.5 cGy/min) but not high (12 cGy/min) dose rates. Ultraviolet-mediated HIV-LTR induction was inhibited by low-dose-rate neutron exposure

  12. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C. Y.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: candy_borja@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    The absorbed dose due to neutrons by a Complementary Metal Oxide Semiconductor (CMOS) has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes a patient that must be treated by radiotherapy with a linear accelerator; the pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. When the Linac is working in Bremsstrahlung mode an undesirable neutron field is produced due to photoneutron reactions; these neutrons could damage the CMOS putting the patient at risk during the radiotherapy treatment. In order to estimate the neutron dose in the CMOS a Monte Carlo calculation was carried out where a full radiotherapy vault room was modeled with a W-made spherical shell in whose center was located the source term of photoneutrons produced by a Linac head operating in Bremsstrahlung mode at 18 MV. In the calculations a phantom made of tissue equivalent was modeled while a beam of photoneutrons was applied on the phantom prostatic region using a field of 10 x 10 cm{sup 2}. During simulation neutrons were isotropically transported from the Linac head to the phantom chest, here a 1 {theta} x 1 cm{sup 2} cylinder made of polystyrene was modeled as the CMOS, where the neutron spectrum and the absorbed dose were estimated. Main damages to CMOS are by protons produced during neutron collisions protective cover made of H-rich materials, here the neutron spectrum that reach the CMOS was calculated showing a small peak around 0.1 MeV and a larger peak in the thermal region, both connected through epithermal neutrons. (Author)

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

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo; Aizawa, Otohiko; Nozaki, Tetsuya

    1981-01-01

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

  14. Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

    International Nuclear Information System (INIS)

    Field, Kevin G.; Yang, Ying; Allen, Todd R.; Busby, Jeremy T.

    2015-01-01

    Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling with mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were used to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibiting unique defect sink characteristics depending on their local structure. Such interactions were found to be consistent across all doses investigated and to have larger global implications, including precipitation of Ni–Si clusters near different grain boundary types

  15. The development of BH3105E type neutron dose-equivalent meter

    International Nuclear Information System (INIS)

    Ji Changsong; Wang Tingting; Zhang Shuheng; Tan Baozeng

    2011-01-01

    A new BH3105E Type Neutron Dose-equivalent Meter has been developed. The 'multi-stick' ab- sorption method is used for thermal -14 MeV neutron equal dose-equivalent detection, what gives a high neutron sensitivity of 5 cps/μSv · h-1. RS-232 interface is accepted for signal communication (authors)

  16. SU-E-T-611: Photon and Neutron Peripheral Dose Ratio for Low (6 MV) and High (15 MV) Energy for Treatment Selection

    Energy Technology Data Exchange (ETDEWEB)

    Irazola, L; Sanchez-Doblado, F [Departamento de Fisiologia Medica y Biofisica, Universidad de Seville (Spain); Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Seville (Spain); Terron, J; Ortiz-Seidel, M [Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Seville (Spain); Departamento de Fisiologia Medica y Biofisica, Universidad de Seville (Spain); Sanchez-Nieto, B [Instituto de Fisica, Pontificia Universidad Catolica de Chile, Santiago (Chile)

    2015-06-15

    Purpose: Differences between radiotherapy techniques and energies, can offer improvements in tumor coverage and organs at risk preservation. However, a more complete decision should include peripheral doses delivered to the patient. The purpose of this work is the balance of photon and neutron peripheral doses for a prostate case solved with 6 different treatment modalities. Methods: Inverse and Forward IMRT and 3D-CRT in 6 and 15 MV for a Siemens Primus linac, using the same CT data set and contours. The methodology described in [1], was used with the TNRD thermal neutron detector [2] for neutron peripheral dose estimation at 7 relevant organs (colon, esophagus, stomach, liver, lung, thyroid and skin). Photon doses were estimated for these organs by terms of the algorithm proposed in [3]. Plans were optimized with the same restrictions and limited to 30 segments in the Inverse case. Results: A similar photon peripheral dose was found comparing 6 and 15 MV cases with slightly higher values of (1.9 ± 1.6) % in mean, for the 6 MV cases. Neutron presence when using 15 MV, represents an increase in peripheral dose of (18 ± 17) % in average. Due to the higher number of MU used in Inverse IMRT, an increasing of (22 ± 3) % in neutron dose is found related to Forward and 3D-CRT plans. This corresponds to photon doses within 44 and 255 mSv along the organs, for a dose prescription of 68 Gy at the isocenter. Conclusion: Neutron and photon peripheral doses for a prostate treatment planified in 6 different techniques have been analyzed. 6 MV plans are slightly more demanding in terms of photon peripheral doses. Inverse technique in 15 MV has Result to be the most demanding one in terms of total peripheral doses, including neutrons and photons.

  17. Long distance elementary measurement of the radiation dose ratio produced by neutron activation

    International Nuclear Information System (INIS)

    Zhou Changgeng; Lou Benchao; Wu Chunlei; Hu Yonghong; Li Yan

    2009-04-01

    The working principle and the structure and performances of a long distance controllable individual radiation dose ratio instrument are described. The radiation dose ratio produced by neutron activation is elementarily measured by using this instrument in the neutron generator hall with high neutron yield. When neutron yield arrives to 2 x 10 11 s -1 , the radiation dose ratio produced by neutron activation is 99.9 μSv/h in 1 h after the generator being stopped. The radiation dose ratio is reduced to 24.4 μSv/h in 39 h after the generator being stopped. When neutron yield is 3.2 x 10 10 s -1 , the radiation dose ratio produced by neutron activation is 21.9 μSv/h in 36 min, after the generator being stopped. The measurement results may provide reference for physical experimenters and neutron generator operators. (authors)

  18. Experimental method research on neutron equal dose-equivalent detection

    International Nuclear Information System (INIS)

    Ji Changsong

    1995-10-01

    The design principles of neutron dose-equivalent meter for neutron biological equi-effect detection are studied. Two traditional principles 'absorption net principle' and 'multi-detector principle' are discussed, and on the basis of which a new theoretical principle for neutron biological equi-effect detection--'absorption stick principle' has been put forward to place high hope on both increasing neutron sensitivity of this type of meters and overcoming the shortages of the two traditional methods. In accordance with this new principle a brand-new model of neutron dose-equivalent meter BH3105 has been developed. Its neutron sensitivity reaches 10 cps/(μSv·h -1 ), 18∼40 times higher than that of all the same kinds of meters 0.23∼0.56 cps/(μSv·h -1 ), available today at home and abroad and the specifications of the newly developed meter reach or surpass the levels of the same kind of meters. Therefore the new theoretical principle of neutron biological equi-effect detection--'absorption stick principle' is proved to be scientific, advanced and useful by experiments. (3 refs., 3 figs., 2 tabs.)

  19. TU-F-CAMPUS-T-01: Dose and Energy Spectra From Neutron Induced Radioactivity in Medical Linear Accelerators Following High Energy Total Body Irradiation

    International Nuclear Information System (INIS)

    Keehan, S; Taylor, M; Franich, R; Smith, R; Dunn, L; Kron, T

    2015-01-01

    Purpose: To assess the risk posed by neutron induced activation of components in medical linear accelerators (linacs) following the delivery of high monitor unit 18 MV photon beams such as used in TBI. Methods: Gamma spectroscopy was used to identify radioisotopes produced in components of a Varian 21EX and an Elekta Synergy following delivery of photon beams. Dose and risk estimates for TBI were assessed using dose deliveries from an actual patient treatment. A 1 litre spherical ion chamber (PTW, Germany) has been used to measure the dose at the beam exit window and at the total body irradiation (TBI) treatment couch following large and small field beams with long beam-on times. Measurements were also made outside of the closed jaws to quantify the benefit of the attenuation provided by the jaws. Results: The radioisotopes produced in the linac head have been identified as 187 W, 56 Mn, 24 Na and 28 Al, which have half-lives from between 2.3 min to 24 hours. The dose at the beam exit window following an 18 MV 2197 MU TBI beam delivery was 12.6 µSv in ten minutes. The dose rate at the TBI treatment couch 4.8 m away is a factor of ten lower. For a typical TBI delivered in six fractions each consisting of four beams and an annual patient load of 24, the annual dose estimate for a staff member at the treatment couch for ten minutes is 750 µSv. This can be further reduced by a factor of about twelve if the jaws are closed before entering the room, resulting in a dose estimate of 65 µSv. Conclusion: The dose resulting from the activation products for a representative TBI workload at our clinic of 24 patients per year is 750 µSv, which can be further reduced to 65 µSv by closing the jaws

  20. Seed irradiation with continuously increasing doses of thermal neutrons

    International Nuclear Information System (INIS)

    Uhlik, J.; Pfeifer, M.; Pittermann, P.

    1977-01-01

    In the 'Raman' pea cv. the biological activity of thermal neutrons was investigated after irradiation of a 780 mm column of seeds for 3000 and 4167 seconds with a flux of 5.607 x 10 9 n.cm -2 per second. For different fractions of the seed column the average density of the neutron flux was calculated. It was proved that for the described method of seed irradiation it was sufficient to determine only the dose approaching the lethal dose. If a sufficiently high column of seeds is used part of the column of seeds will be irradiated with the optimum range of doses. The advantages of the suggested method of irradiation are not only smaller time and technological requirements resulting from the need for the determination of only the critical lethal dose of radiation by means of inhibition tests performed with seedlings, but also a simpler irradiation procedure. The suggested method of irradiation is at least nine times cheaper. (author)

  1. A 'hybrid' neutron area survey instrument for the determination of neutron dose quantities in the workplace

    International Nuclear Information System (INIS)

    Tanner, R.J.; Jenkins, R.; Lowe, T.; Silvie, J.; Joyce, M.J.; Winsby, A.; Molinos, C.

    2005-01-01

    Full text: Neutron survey instruments are used routinely to determine the dose rates in areas where persons may be occupationally exposed. With a few exceptions, these instruments generally use a proportional counter with a high thermal neutron response located in a moderating sphere of CH 2 . The moderating sphere in such designs contains a thermal neutron absorber to reduce the over-response to thermal and intermediate energy neutrons. However, the commercially available examples of such instruments tend to have strongly energy dependent ambient dose equivalent response characteristics. In particular, they often over-respond in the energy range between 1 eV and 10 keV. A prototype of a novel design has been produced that uses seven detectors located in a moderating sphere of CH 2 , six near the surface to detect thermal and epithermal neutrons, and one in the centre to detect fast neutrons. This has been characterized using a combination of MCNP modelling and measurements to produce an instrument that has improved energy dependence of response characteristics. Additionally, the use of seven detectors offers direction and field hardness information. The design and calibration of the instrument are described and its response in workplaces calculated. (author)

  2. 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)

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  4. Estimate of absorbed dose received by individuals irradiated with neutrons

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Mauricio, C.L.P.

    1995-01-01

    An innovating methodology is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the methodology here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μGy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author). 5 refs., 1 fig., 4 tabs

  5. Intercomparison of high energy neutron personnel dosimeters

    International Nuclear Information System (INIS)

    McDonald, J.C.; Akabani, G.; Loesch, R.M.

    1993-03-01

    An intercomparison of high-energy neutron personnel dosimeters was performed to evaluate the uniformity of the response characteristics of typical neutron dosimeters presently in use at US Department of Energy (DOE) accelerator facilities. It was necessary to perform an intercomparison because there are no national or international standards for high-energy neutron dosimetry. The testing that is presently under way for the Department of Energy Laboratory Accreditation Program (DOELAP) is limited to the use of neutron sources that range in energy from about 1 keV to 2 MeV. Therefore, the high-energy neutron dosimeters presently in use at DOE accelerator facilities are not being tested effectively. This intercomparison employed neutrons produced by the 9 Be(p,n) 9 B interaction at the University of Washington cyclotron, using 50-MeV protons. The resulting neutron energy spectrum extended to a maximum of approximately 50-MeV, with a mean energy of about 20-MeV. Intercomparison results for currently used dosimeters, including Nuclear Type A (NTA) film, thermoluminescent dosimeter (TLD)-albedo, and track-etch dosimeters (TEDs), indicated a wide variation in response to identical doses of high-energy neutrons. Results of this study will be discussed along with a description of plans for future work

  6. Method and apparatus for determining the dose value of neutrons

    International Nuclear Information System (INIS)

    Burgkhardt, B.; Piesch, E.

    1976-01-01

    A method is provided for determining the dose value of neutrons leaving a body as thermal and intermediate neutrons after having been scattered in the body. A first dose value of thermal and intermediate neutrons is detected on the surface of the body by means of a first detector for neutrons which is shielded against thermal and intermediate neutrons not emerging from the body. A second detector is used to measure a second dose value of the thermal and intermediate neutrons not emerging from the body. A first correction factor based on the first and second values is obtained from a calibration diagram and is applied to the first dose value to determine a first corrected first dose value. 21 Claims, 6 Drawing Figures

  7. Neutrons in active proton therapy. Parameterization of dose and dose equivalent

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Uwe; Haelg, Roger A. [Univ. of Zurich (Switzerland). Dept. of Physics; Radiotherapy Hirslanden AG, Aarau (Switzerland); Lomax, Tony [Paul Scherrer Institute, Villigen (Switzerland). Center for Proton Therapy

    2017-08-01

    One of the essential elements of an epidemiological study to decide if proton therapy may be associated with increased or decreased subsequent malignancies compared to photon therapy is an ability to estimate all doses to non-target tissues, including neutron dose. This work therefore aims to predict for patients using proton pencil beam scanning the spatially localized neutron doses and dose equivalents. The proton pencil beam of Gantry 1 at the Paul Scherrer Institute (PSI) was Monte Carlo simulated using GEANT. Based on the simulated neutron dose and neutron spectra an analytical mechanistic dose model was developed. The pencil beam algorithm used for treatment planning at PSI has been extended using the developed model in order to calculate the neutron component of the delivered dose distribution for each treated patient. The neutron dose was estimated for two patient example cases. The analytical neutron dose model represents the three-dimensional Monte Carlo simulated dose distribution up to 85 cm from the proton pencil beam with a satisfying precision. The root mean square error between Monte Carlo simulation and model is largest for 138 MeV protons and is 19% and 20% for dose and dose equivalent, respectively. The model was successfully integrated into the PSI treatment planning system. In average the neutron dose is increased by 10% or 65% when using 160 MeV or 177 MeV instead of 138 MeV. For the neutron dose equivalent the increase is 8% and 57%. The presented neutron dose calculations allow for estimates of dose that can be used in subsequent epidemiological studies or, should the need arise, to estimate the neutron dose at any point where a subsequent secondary tumour may occur. It was found that the neutron dose to the patient is heavily increased with proton energy.

  8. Thermal neutron dose calculation in synovium membrane for BNCS

    International Nuclear Information System (INIS)

    Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.

    2006-01-01

    A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values. (author)

  9. Out-of-field doses and neutron dose equivalents for electron beams from modern Varian and Elekta linear accelerators.

    Science.gov (United States)

    Cardenas, Carlos E; Nitsch, Paige L; Kudchadker, Rajat J; Howell, Rebecca M; Kry, Stephen F

    2016-07-08

    Out-of-field doses from radiotherapy can cause harmful side effects or eventually lead to secondary cancers. Scattered doses outside the applicator field, neutron source strength values, and neutron dose equivalents have not been broadly investigated for high-energy electron beams. To better understand the extent of these exposures, we measured out-of-field dose characteristics of electron applicators for high-energy electron beams on two Varian 21iXs, a Varian TrueBeam, and an Elekta Versa HD operating at various energy levels. Out-of-field dose profiles and percent depth-dose curves were measured in a Wellhofer water phantom using a Farmer ion chamber. Neutron dose was assessed using a combination of moderator buckets and gold activation foils placed on the treatment couch at various locations in the patient plane on both the Varian 21iX and Elekta Versa HD linear accelerators. Our findings showed that out-of-field electron doses were highest for the highest electron energies. These doses typically decreased with increasing distance from the field edge but showed substantial increases over some distance ranges. The Elekta linear accelerator had higher electron out-of-field doses than the Varian units examined, and the Elekta dose profiles exhibited a second dose peak about 20 to 30 cm from central-axis, which was found to be higher than typical out-of-field doses from photon beams. Electron doses decreased sharply with depth before becoming nearly constant; the dose was found to decrease to a depth of approximately E(MeV)/4 in cm. With respect to neutron dosimetry, Q values and neutron dose equivalents increased with electron beam energy. Neutron contamination from electron beams was found to be much lower than that from photon beams. Even though the neutron dose equivalent for electron beams represented a small portion of neutron doses observed under photon beams, neutron doses from electron beams may need to be considered for special cases.

  10. Does fast-neutron radiotherapy merely reduce the radiation dose

    International Nuclear Information System (INIS)

    Ando, Koichi

    1984-01-01

    We examined whether fast-neutron radiotherapy is superior to low-LET radiotherpy by comparing the relationship between cell survival and tumor control probabilities after exposure of tumor-bearing (species) to the two modalities. Analysis based on TCD 50 assay and lung colony assay indicated that single dose of fast neutron achieved animal cures at higher survival rates than other radiation modalities including single and fractionated γ-ray doses, fractionated doses of fast neutron, and the mixed-beam scheme with a sequence of N-γ-γ-γ-N. We conclude that fast-neutron radiotherapy cured animal tumors with lower cell killing rates other radiation modalities. (author)

  11. Impact of the Revised 10 CFR 835 on the Neutron Dose Rates at LLNL

    International Nuclear Information System (INIS)

    Radev, R.

    2009-01-01

    In June 2007, 10 CFR 835 (1) was revised to include new radiation weighting factors for neutrons, updated dosimetric models, and dose terms consistent with the newer ICRP recommendations. A significant aspect of the revised 10 CFR 835 is the adoption of the recommendations outlined in ICRP-60 (2). The recommended new quantities demand a review of much of the basic data used in protection against exposure to sources of ionizing radiation. The International Commission on Radiation Units and Measurements has defined a number of quantities for use in personnel and area monitoring (3,4,5) including the ambient dose equivalent H*(d) to be used for area monitoring and instrument calibrations. These quantities are used in ICRP-60 and ICRP-74. This report deals only with the changes in the ambient dose equivalent and ambient dose rate equivalent for neutrons as a result of the implementation of the revised 10 CFR 835. In the report, the terms neutron dose and neutron dose rate will be used for convenience for ambient neutron dose and ambient neutron dose rate unless otherwise stated. This report provides a qualitative and quantitative estimate of how much the neutron dose rates at LLNL will change with the implementation of the revised 10 CFR 835. Neutron spectra and dose rates from selected locations at the LLNL were measured with a high resolution spectroscopic neutron dose rate system (ROSPEC) as well as with a standard neutron rem meter (a.k.a., a remball). The spectra obtained at these locations compare well with the spectra from the Radiation Calibration Laboratory's (RCL) bare californium source that is currently used to calibrate neutron dose rate instruments. The measurements obtained from the high resolution neutron spectrometer and dose meter ROSPEC and the NRD dose meter compare within the range of ±25%. When the new radiation weighting factors are adopted with the implementation of the revised 10 CFR 835, the measured dose rates will increase by up to 22%. The

  12. A Study on the Neutron Dose Distribution in Case of 10 MV X-rays Radiotherapy

    International Nuclear Information System (INIS)

    Park, Cheol Soo; Shin, Seong Soo; Lim, Cheong Hwan; Jung, Hong Ryang

    2008-01-01

    This study is to measure the radiation dose of neutrons generated by the particle accelerator during X-ray (photon) treatment with a neutron detection method by using CR-39, and to research how the generation of neutrons may incur problems associated with radiation doses for patient treatment when using high energy photons for cancer treatment as a clinical application. The findings are summarized as follows : The results showed that average 0.35 mSv was measured with exposure of 1 Gy photon in case of fast neutron, 0.65 mSv with exposure of 2 Gy photon, 1.82 mSv exposure of 5 Gy, 0.26 mSv with exposure of 1 Gy photon in case of thermal neutron, 0.56 mSv with exposure of 2 Gy photon, and 1.23 mSv with exposure of 5 Gy of photon. By measuring the occurrence of neutron by using Wedge Filter, it has been confirmed that the occurrence of neutrons increased when using Wedge Filter. The results also showed that more neutrons were detected over the existing experiments when using an SRS Cone requiring high doses of radiation. Total 2.85 mSv neutrons were found on the average with exposure of 5 Gy photon in case of fast neutron and 1.37 mSv neutrons were found on the average with exposure of 5 Gy photon in case of thermal neutron. During the general treatment, about 1.6 times more neutrons over 5 Gy photon were found in case of fast neutron and about 1.12 time more neutrons over 5 Gy photon were found in case of thermal neutron.

  13. Alterations in water and electrolyte absorption in the rat colon following neutron irradiation: influence of neutron component and irradiation dose.

    Science.gov (United States)

    Dublineau, I; Ksas, B; Joubert, C; Aigueperse, J; Gourmelon, P; Griffiths, N M

    2002-12-01

    To study the absorptive function of rat colon following whole-body exposure to neutron irradiation, either to the same total dose with varying proportion of neutrons or to the same neutron proportion with an increasing irradiation dose. Different proportions of neutron irradiation were produced from the reactor SILENE using a fissile solution of uranium nitrate (8, 47 and 87% neutron). Water and electrolyte fluxes were measured in the rat in vivo under anaesthesia by insertion into the descending colon of an agarose gel cylinder simulating the faeces. Functional studies were completed by histological analyses. In the first set of experiments, rats received 3.8 Gy with various neutron percentages and were studied from 1 to 14 days after exposure. In the second set of experiments, rats were exposed to increasing doses of irradiation (1-4Gy) with a high neutron percentage (87%n) and were studied at 4 days after exposure. The absorptive capacity of rat colon was diminished by irradiation at 3-5 days, with a nadir at 4 days. The results demonstrate that an increase in the neutron proportion is associated with an amplification of the effects. Furthermore, a delay in the re-establishment of normal absorption was observed with the high neutron proportion (87%n). A dose-dependent reduction of water absorption by rat colon was also observed following neutron irradiation (87%n), with a 50% reduction at 3 Gy. Comparison of this dose-effect curve with the curve obtained following gamma (60)Co-irradiation indicates an RBE of 2.2 for absorptive colonic function in rat calculated at 4 days after exposure.

  14. The experimental method for neutron dose-equivalent detection

    International Nuclear Information System (INIS)

    Ji Changsong

    1992-01-01

    A new method, for getting neutron dose-equivalent Cd rode absorption method is described. The method adopts Cd-rode-swarm buck absorption, which greatly improved the neutron sensitivity and simplified the adjustment method. By this method, the author has developed BH3105 model neutron dose equivalent meter, the sensitivity of this instrument reach 10 cps/μSvh -1 . γ-ray depression rate reaches 4000:1, the measurement range is 0.1 μSv/h-10 6 μSv/h. The energy response is good (from thermal neutron-14 MeV neutron), this instrument can be used to measure the dose equivalent of the neutron areas

  15. Dose inhomogeneities for photons and neutrons near interfaces

    International Nuclear Information System (INIS)

    Broerse, J. J.; Zoetelief, J.

    2004-01-01

    Perturbations of charged particle equilibrium (CPE) at interfaces of materials of different atomic composition can lead to considerable differences in the energy deposition by photons and neutrons. Specific examples of these interface perturbations are encountered during irradiation of body cavities and soft tissue adjacent to bone or metallic implants and irradiation of cells in monolayer on the bottom of culture dishes. Another example is the build-up of CPE at air-tissue interfaces, referred to in radiotherapy as the skin sparing effect. For photon irradiation excess production of secondary electrons in high-Z materials, such as glass, bone or gold, will induce appreciably higher doses and decreased cell survival compared to the equilibrium situation. The energy dissipation of fast neutrons in biological materials occurs through recoil protons, heavy recoil nuclei and products of nuclear reactions. Owing to the large contribution from recoil protons to the neutron kerma, the hydrogen content of the biological material mainly determines the energy deposition. For neutron irradiation of cells in monolayer, CPE can be established or deliberately avoided by mounting tissue-equivalent plastic or carbon discs in front of the cells, respectively. This approach makes it possible to distinguish the biological effects of the low- and high-LET radiation components. (authors)

  16. High energy neutron generator

    International Nuclear Information System (INIS)

    Barjon, R.; Breynat, G.

    1987-01-01

    This patent describes a generator of fast neutrons only slightly contaminated by neutrons of energy less than 15 MeV, comprising a source of charged particles of energy equal to at least 15 MeV, a target made of lithium deuteride, and means for cooling the target. The target comprises at least two elements placed in series in the path of the charged particles and separated from each other, the thickness of each of the elements being selected as a function of the average energy of the charged particles emitted from the source and the energy of the fast neutrons to be generated such that neutrons of energy equal to at least 15 MeV are emitted in the forward direction in response to the bombardment of the target from behind by the charged particles. The target cooling means comprises means for circulating between and around the elements a gas which does not chemically react with lithium deuteride

  17. 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)

  18. 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.

  19. Study on method of dose estimation for the Dual-moderated neutron survey meter

    International Nuclear Information System (INIS)

    Zhou, Bo; Li, Taosheng; Xu, Yuhai; Gong, Cunkui; Yan, Qiang; Li, Lei

    2013-01-01

    In order to study neutron dose measurement in high energy radiation field, a Dual-moderated survey meter in the range from 1 keV to 300 MeV mean energies spectra has been developed. Measurement results of some survey meters depend on the neutron spectra characteristics in different neutron radiation fields, so the characteristics of the responses to various neutron spectra should be studied in order to get more reasonable dose. In this paper the responses of the survey meter were calculated under different neutron spectra data from IAEA of Technical Reports Series No. 318 and other references. Finally one dose estimation method was determined. The range of the reading per H*(10) for the method estimated is about 0.7–1.6 for the neutron mean energy range from 50 keV to 300 MeV. -- Highlights: • We studied a novel high energy neutron survey meter. • Response characteristics of the survey meter were calculated by using a series of neutron spectra. • One significant advantage of the survey meter is that it can provide mean energy of radiation field. • Dose estimate deviation can be corrected. • The range of corrected reading per H*(10) is about 0.7–1.6 for the neutron fluence mean energy range from 0.05 MeV to 300 MeV

  20. Development of a neutron personal dose equivalent detector

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.; Momose, T.; Nunomiya, T.; Aoyama, K.

    2007-01-01

    A new neutron-measuring instrument that is intended to measure a neutron personal dose equivalent, H p (10) was developed. This instrument is composed of two parts: (1) a conventional moderator-based neutron dose equivalent meter and (2) a neutron shield made of borated polyethylene, which covers a backward hemisphere to adjust the angular dependence. The whole design was determined on the basis of MCNP calculations so as to have response characteristics that would generally match both the energy and angular dependencies of H p (10). This new instrument will be a great help in assessing the reference values of neutron H p (10) during field testing of personal neutron dosemeters in workplaces and also in interpreting their readings. (authors)

  1. Neutron doses to personnel from a 24 MeV betatron

    International Nuclear Information System (INIS)

    Beckham, W.A; Entwistle, R.F.

    1987-01-01

    Neutrons are produced by bombardment of most materials by high-energy photons. Because the x-ray shielding around high-energy x-ray generators may not have been designed with neutrons in mind there may be unexpected contributions to the radiation doses of staff working in the immediate vicinity. Neutron fluxes in the working area close to an Allis-Chalmers 24 MeV betatron have been measured using a lithium-6-loaded scintillator and the dose rates calculated. Hazard of staff has been found to be low; typical dose-equivalent rates in occupied areas range from 0.0042 to 0.012 mrem/hour. The flux of fast neutrons in the treatment room was found to be essentially zero. Measurements of neutron flux may be routinely performed using the scintillation detector (NE 912) described, and could usefully form part of the acceptance protocol for any new accelerator

  2. Fast Neutron Dose Distribution in a Linac Radiotherapy Facility

    International Nuclear Information System (INIS)

    Al-Othmany, D.Sh.; Abdul-Majid, S.; Kadi, M.W.

    2011-01-01

    CR-39 plastic detectors were used for fast neutron dose mapping in the radiotherapy facility at King AbdulAziz University Hospital (KAUH). Detectors were calibrated using a 252 Cf neutron source and a neutron dosimeter. After exposure chemical etching was performed using 6N NaOH solution at 70 degree C. Tracks were counted using an optical microscope and the number of tracks/cm 2 was converted to a neutron dose. 15 track detectors were distributed inside and outside the therapy room and were left for 32 days. The average neutron doses were 142.3 mSv on the accelerator head, 28.5 mSv on inside walls, 1.4 mSv beyond the beam shield, and 1 mSv in the control room

  3. Photon and neutron doses of the personnel using moisture and density measurement devices

    Energy Technology Data Exchange (ETDEWEB)

    Carinou, E.; Papadomarkaki, E.; Tritakis, P.; Hourdakis, C.I.; Kamenopoulou, V. [Greek Atomic Energy Commission, Agia Paraskevi, Attiki, 60092 (Greece)

    2006-07-01

    The objective of this study is to present the evolution of the photon doses received by the workers who use mobile devices for measuring the moisture and the density in various materials and to estimate the neutron doses. The workers employed in more than 30 construction companies in Greece were 76 in 2004. The devices used for that purpose incorporate a {sup 137}Cs source for density measurements and an {sup 241}Am-Be source for moisture measurements of soil, asphalt or concrete. Photon and neutron measurements were performed occasionally during the on site inspections. The results of the measurements showed that the photon and neutron dose rates were not negligible. The workers were monitored for photon radiation using film badges (Kodak Type 2, Holder NRPB type) till the year 2000 and then TLD badges issued by the Greek Atomic Energy Commission (GAEC), on a monthly basis. Since the neutron dose rates measured by a rem-meter were not so high, no neutron dosemeters were issued for them. Their personal dose equivalent data for photons are kept in the National Dose Registry Information System (N.D.R.I.S.) in G.A.E.C. and were used for statistical analysis for the period from 1997 till 2004. As far as the neutrons are concerned, a Monte Carlo code was used to simulate the measuring devices and the working positions in order to calculate the neutron individual doses. (authors)

  4. Photon and neutron doses of the personnel using moisture and density measurement devices

    International Nuclear Information System (INIS)

    Carinou, E.; Papadomarkaki, E.; Tritakis, P.; Hourdakis, C.I.; Kamenopoulou, V.

    2006-01-01

    The objective of this study is to present the evolution of the photon doses received by the workers who use mobile devices for measuring the moisture and the density in various materials and to estimate the neutron doses. The workers employed in more than 30 construction companies in Greece were 76 in 2004. The devices used for that purpose incorporate a 137 Cs source for density measurements and an 241 Am-Be source for moisture measurements of soil, asphalt or concrete. Photon and neutron measurements were performed occasionally during the on site inspections. The results of the measurements showed that the photon and neutron dose rates were not negligible. The workers were monitored for photon radiation using film badges (Kodak Type 2, Holder NRPB type) till the year 2000 and then TLD badges issued by the Greek Atomic Energy Commission (GAEC), on a monthly basis. Since the neutron dose rates measured by a rem-meter were not so high, no neutron dosemeters were issued for them. Their personal dose equivalent data for photons are kept in the National Dose Registry Information System (N.D.R.I.S.) in G.A.E.C. and were used for statistical analysis for the period from 1997 till 2004. As far as the neutrons are concerned, a Monte Carlo code was used to simulate the measuring devices and the working positions in order to calculate the neutron individual doses. (authors)

  5. Fast neutron radiation inactivation of Bacillus subtilis: Absorbed dose determination

    International Nuclear Information System (INIS)

    Song Lingli; Zheng Chun; Ai Zihui; Li Junjie; Dai Shaofeng

    2011-01-01

    In this paper, fast neutron inactivation effects of Bacillus subtilis were investigated with fission fast neutrons from CFBR-II reactor of INPC (Institute of Nuclear Physics and Chemistry) and mono-energetic neutrons from the Van de Graaff accelerator at Peking University. The method for determining the absorbed dose in the Bacillus subtilis suspension contained in test tubes is introduced. The absorbed dose, on account of its dependence on the volume and the form of confined state, was determined by combined experiments and Monte Carlo method. Using the calculation results of absorbed dose, the fast neutron inactivation effects on Bacillus subtilis were studied. The survival rates and absorbed dose curve was constructed. (authors)

  6. Nominal effective radiation doses delivered during clinical trials of boron neutron capture therapy

    International Nuclear Information System (INIS)

    Capala, J.; Diaz, A.Z.; Chanana, A.D.

    1997-01-01

    Boron neutron capture therapy (BNCT) is a binary system that, in theory, should selectively deliver lethal, high linear energy transfer (LET) radiation to tumor cells dispersed within normal tissues. It is based on the nuclear reaction 10-B(n, α)7-Li, which occurs when the stable nucleus of boron-10 captures a thermal neutron. Due to the relatively high cross-section of the 10-B nucleus for thermal neutron capture and short ranges of the products of this reaction, tumor cells in the volume exposed to thermal neutrons and containing sufficiently high concentration of 10-B would receive a much higher radiation dose than the normal cells contained within the exposed volume. Nevertheless, radiation dose deposited in normal tissue by gamma and fast neutron contamination of the neutron beam, as well as neutron capture in nitrogen, 14-N(n,p)14-C, hydrogen, 1-H(n,γ)2-H, and in boron present in blood and normal cells, limits the dose that can be delivered to tumor cells. It is, therefore, imperative for the success of the BNCT the dosed delivered to normal tissues be accurately determined in order to optimize the irradiation geometry and to limit the volume of normal tissue exposed to thermal neutrons. These are the major objectives of BNCT treatment planning

  7. High energy (42-66 MeV reactions) fast neutron dose optimization studies in the head and neck, thorax, upper abdomen, pelvis and extremities

    International Nuclear Information System (INIS)

    Griffin, T.W.; Laramore, G.E.; Maor, M.H.; Hendrickson, F.R.; Parker, R.G.; Davis, L.W.

    1990-01-01

    550 Patients were entered into a set of dose-searching studies designed to determine normal tissue tolerances to high energy (42-66 MeV reactions) fast neutrons delivered in 12 equal fractions over 4 weeks. Patients were stratified by treatment facility and then randomized to receive 16, 18 or 20 Gy for tumors located in the upper abdomen or pelvis, and 18, 20 or 22 Gy for tumors located in the head and neck, thorax or extremities. Following completion of the randomized protocols, additional patients were studied at the 20.4 Gy level in the head and neck, thorax and pelvis. Normal tissue effect scoring was accomplished using the RTOG-EORTC acute and late normal tissue effect scales. Acute Grade 3+ toxicity rates in the head and neck were 19 per cent for 20/20.4 Gy and 20 per cent for 22 Gy. Time adjusted late toxicity rates in the head and neck at 12 months were 15 per cent for 20/20.4 Gy and 0 per cent for 22 Gy. The 18 Gy treatment arm of the head and neck protocol was dropped early in the study after only two patients were accrued. For cases treated in the thorax, acute Grade 3+ toxicity rates were 6 per cent for 18 Gy, 15 per cent for 20/20.4 Gy and 7 per cent for 22 Gy. Late toxicity rates at 12 months were 0 per cent for 18 Gy, 11 per cent for 20/20.4 Gy and 18 per cent for 22 Gy. Acute Grade 3+ toxicity rates in the upper abdomen were 0 per cent for 16 Gy, 18 per cent for 18 Gy and 12 per cent for 20 Gy. There were no Grade 3+ late toxicities in the upper abdomen. In the pelvis acute Grade 3+ toxicity rates were 0 per cent for 16 Gy, 3 per cent for 18 Gy and 3 per cent for 20/20.4 Gy. Late Grade 3+ toxicities at 24 months were 20 per cent for 16 Gy, 5 per cent for 18 Gy and 24 per cent for 20/20.4 Gy. In the extremities, acute Grade 3+ toxicity rates were 7 per cent for 20 Gy and 21 per cent for 22 Gy, while at 12 months, late Grade 3+ toxicity rates were 14 and 35 per cent respectively. The 18 Gy treatment arm of the extremities protocol was dropped early

  8. Absorbed dose conversion coefficients for embryo and foetus in neutron fields

    International Nuclear Information System (INIS)

    Chen, J.

    2007-01-01

    The Monte Carlo code MCNPX has been used to determine mean absorbed doses to the embryo and foetus when the mother is exposed to neutron fields. There are situations, such as on-board aircraft, where high-energy neutrons are often peaked in top down (TOP) direction. In addition to previous publications for standard irradiation geometries, this study provides absorbed dose conversion coefficients for the embryo of 8 weeks and the foetus of 3, 6 or 9 months at TOP irradiation geometry. The conversion coefficients are compared with the coefficients in isotropic irradiation (ISO). With increasing neutron energies, the conversion coefficients in TOP irradiation become dominant. A set of conversion coefficients is constructed from the higher value in either ISO or TOP irradiation at a given neutron energy. In cases where the irradiation geometry is not adequately known, this set of conversion coefficients can be used in a conservative dose assessment for embryo and foetus in neutron fields. (authors)

  9. Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

    Energy Technology Data Exchange (ETDEWEB)

    Cecilia, A.; Baccaro, S.; Cemmi, A. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Colli, V.; Gambarini, G. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy); Rosi, G. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Scolari, L. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy)

    2004-07-01

    A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF{sub 2}:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH{sub 3}CH(NH{sub 2})COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT. (authors)

  10. 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)

  11. Use of prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent

    Energy Technology Data Exchange (ETDEWEB)

    Priyada, P.; Sarkar, P.K., E-mail: pradip.sarkar@manipal.edu

    2015-06-11

    The possibility of using measured prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent is explored theoretically. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of a high density polyethylene cylinder to emit prompt gammas from interaction of neutrons with the nuclei of hydrogen and carbon present in polyethylene. The neutron energy dependent responses of hydrogen and carbon nuclei are combined appropriately to match the energy dependent neutron fluence to ambient dose equivalent conversion coefficients. The proposed method is tested initially with simulated spectra and then validated using experimental measurements with an Am–Be neutron source. Experimental measurements and theoretical simulations have established the feasibility of estimating neutron ambient dose equivalent using measured neutron induced prompt gammas emitted from polyethylene with an overestimation of neutron dose at very low energies. - Highlights: • A new method for estimating H{sup ⁎}(10) using prompt gamma emissions from HDPE. • Linear combination of 2.2 MeV and 4.4 MeV gamma intensities approximates DCC (ICRP). • Feasibility of the method was established theoretically and experimentally. • The response of the present technique is very similar to that of the rem meters.

  12. 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.

  13. Rapid Measurement of Neutron Dose Rate for Transport Index

    International Nuclear Information System (INIS)

    Morris, R.L.

    2000-01-01

    A newly available neutron dose equivalent remmeter with improved sensitivity and energy response has been put into service at Rocky Flats Environmental Technology Site (RFETS). This instrument is being used to expedite measurement of the Transport Index and as an ALARA tool to identify locations where slightly elevated neutron dose equivalent rates exist. The meter is capable of measuring dose rates as low as 0.2 μSv per hour (20 μrem per hour). Tests of the angular response and energy response of the instrument are reported. Calculations of the theoretical instrument response made using MCNPtrademark are reported for materials typical of those being shipped

  14. High sensitivity MOSFET-based neutron dosimetry

    International Nuclear Information System (INIS)

    Fragopoulou, M.; Konstantakos, V.; Zamani, M.; Siskos, S.; Laopoulos, T.; Sarrabayrouse, G.

    2010-01-01

    A new dosemeter based on a metal-oxide-semiconductor field effect transistor sensitive to both neutrons and gamma radiation was manufactured at LAAS-CNRS Laboratory, Toulouse, France. In order to be used for neutron dosimetry, a thin film of lithium fluoride was deposited on the surface of the gate of the device. The characteristics of the dosemeter, such as the dependence of its response to neutron dose and dose rate, were investigated. The studied dosemeter was very sensitive to gamma rays compared to other dosemeters proposed in the literature. Its response in thermal neutrons was found to be much higher than in fast neutrons and gamma rays.

  15. Development of highly effective neutron shields and neutron absorbing materials

    International Nuclear Information System (INIS)

    Tsuda, K.; Matsuda, F.; Taniuchi, H.; Yuhara, T.; Iida, T.

    1993-01-01

    A wide range of materials, including polymers and hydrogen-occluded alloys that might be usable as the neutron shielding material were examined. And a wide range of materials, including aluminum alloys that might be usable as the neutron-absorbing material were examined. After screening, the candidate material was determined on the basis of evaluation regarding its adaptabilities as a high-performance neutron-shielding and neutron-absorbing material. This candidate material was manufactured for trial, after which material properties tests, neutron-shielding tests and neutron-absorbing tests were carried out on it. The specifications of this material were thus determined. This research has resulted in materials of good performance; a neutron-shielding material based on ethylene propylene rubber and titanium hydride, and a neutron-absorbing material based on aluminum and titanium hydride. (author)

  16. Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

    Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

  17. Transportable, Low-Dose Active Fast-Neutron Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

  18. Research activities on dosimetry for high energy neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The external dosimetry research group of JAERI has been calculating dose conversion coefficients for high-energy radiations using particle transport simulation codes. The group has also been developing radiation dose measurement techniques for high-energy neutrons in collaboration with some university groups. (author)

  19. Validation of dose planning calculations for boron neutron capture therapy using cylindrical and anthropomorphic phantoms

    Energy Technology Data Exchange (ETDEWEB)

    Koivunoro, Hanna; Seppaelae, Tiina; Uusi-Simola, Jouni; Merimaa, Katja; Savolainen, Sauli [Department of Physics, POB 64, FI-00014 University of Helsinki (Finland); Kotiluoto, Petri; Seren, Tom; Auterinen, Iiro [VTT Technical Research Centre of Finland, Espoo, POB 1000, FI-02044 VTT (Finland); Kortesniemi, Mika, E-mail: hanna.koivunoro@helsinki.f [HUS Helsinki Medical Imaging Center, University of Helsinki, POB 340, FI-00029 HUS (Finland)

    2010-06-21

    In this paper, the accuracy of dose planning calculations for boron neutron capture therapy (BNCT) of brain and head and neck cancer was studied at the FiR 1 epithermal neutron beam. A cylindrical water phantom and an anthropomorphic head phantom were applied with two beam aperture-to-surface distances (ASD). The calculations using the simulation environment for radiation application (SERA) treatment planning system were compared to neutron activation measurements with Au and Mn foils, photon dose measurements with an ionization chamber and the reference simulations with the MCNP5 code. Photon dose calculations using SERA differ from the ionization chamber measurements by 2-13% (disagreement increased along the depth in the phantom), but are in agreement with the MCNP5 calculations within 2%. The {sup 55}Mn(n,{gamma}) and {sup 197}Au(n,{gamma}) reaction rates calculated using SERA agree within 10% and 8%, respectively, with the measurements and within 5% with the MCNP5 calculations at depths >0.5 cm from the phantom surface. The {sup 55}Mn(n,{gamma}) reaction rate represents the nitrogen and boron depth dose within 1%. Discrepancy in the SERA fast neutron dose calculation (of up to 37%) is corrected if the biased fast neutron dose calculation option is not applied. Reduced voxel cell size ({<=}0.5 cm) improves the SERA calculation accuracy on the phantom surface. Despite the slight overestimation of the epithermal neutrons and underestimation of the thermal neutrons in the beam model, neutron calculation accuracy with the SERA system is sufficient for reliable BNCT treatment planning with the two studied treatment distances. The discrepancy between measured and calculated photon dose remains unsatisfactorily high for depths >6 cm from the phantom surface. Increasing discrepancy along the phantom depth is expected to be caused by the inaccurately determined effective point of the ionization chamber.

  20. Fast neutron dose equivalent rates in heavy ion target areas

    International Nuclear Information System (INIS)

    Fulmer, C.B.; Butler, H.M.; Ohnesorge, W.F.; Mosko, S.W.

    1978-01-01

    At heavy ion accelerators, personnel access to areas near the target is sometimes important for successful performance of experiments. Radiation levels determine the amount of time that can be spent in these areas without exceeding maximum permissible exposures. Inasmuch as the fast neutrons contribute the major part of the Rem dose rates in these areas, knowledge of the fast neutron levels is important for planning permissive entry to target areas. Fast neutron dose rates were measured near thick medium mass targets bombarded with beams of C, N, O, and Ne ions. beam energies ranged from 3 to 16 MeV/amu. Dose rates (mrem/h) 1 meter from the target 90 degrees from the beam direction range from approx. 0.05 at MeV/amu to approx. 50 at 16 MeV/amu. These data should be helpful in planning permissive entry to heavy ion target areas

  1. Fast neutron dose equivalent rates in heavy ion target areas

    Energy Technology Data Exchange (ETDEWEB)

    Fulmer, C.B.; Butler, H.M.; Ohnesorge, W.F.; Mosko, S.W.

    1978-01-01

    At heavy ion accelerators, personnel access to areas near the target is sometimes important for successful performance of experiments. Radiation levels determine the amount of time that can be spent in these areas without exceeding maximum permissible exposures. Inasmuch as the fast neutrons contribute the major part of the Rem dose rates in these areas, knowledge of the fast neutron levels is important for planning permissive entry to target areas. Fast neutron dose rates were measured near thick medium mass targets bombarded with beams of C, N, O, and Ne ions. beam energies ranged from 3 to 16 MeV/amu. Dose rates (mrem/h) 1 meter from the target 90 degrees from the beam direction range from approx. 0.05 at MeV/amu to approx. 50 at 16 MeV/amu. These data should be helpful in planning permissive entry to heavy ion target areas.

  2. Estimation dose of secondary neutrons in proton therapy

    International Nuclear Information System (INIS)

    Urban, T.

    2014-01-01

    Most of proton therapy centers for cancer treatment are still based on the passive scattering, in some of them there is system of the active scanning installed as well. The aim of this study is to compare secondary neutron doses in and around target volumes in proton therapy for both treatment techniques and for different energies and profile of incident proton beam. The proton induced neutrons have been simulated in the very simple geometry of tissue equivalent phantom (imitate the patient) and scattering and scanning nozzle, respectively. In simulations of the scattering nozzle, different types of scattering filters and brass collimators have been used as well. 3D map of neutron doses in and around the chosen/potential target volume in the phantom/patient have been evaluated and compared in the context of the dose deposited in the target volume. Finally, the simulation results have been compared with published data. (author)

  3. Wide-range neutron dose determination with CR-39

    International Nuclear Information System (INIS)

    Arneja, A.R.; Waker, A.J.

    1995-01-01

    Optical density measurements of CR-30 irradiated with 252 Cf neutrons and chemically etched with 6.5 N KOH solution have been used to determine neutron absorbed doses between 0.1 and 10 Gy. Optimum etching conditions will depend upon the absorbed dose. Since it is not always possible to know the range of absorbed dose on a CR-39 dosemeter collected from personnel and area monitor stations in a criticality accident situation, a three-step two-hour chemical etch at 60 o C has been found to be appropriate. If after a total of six hours of chemical etching the optical density is found to be below 0.04 for 500 nm light (transmission > 90%) then further treatment in the form of electrochemical etching can be carried out to determine the lower absorbed dose. In this manner, absorbed doses below 0.1 Gy can be determined by counting tracks over a unit area. (author)

  4. Personnel neutron dose assessment upgrade: Volume 2, Field neutron spectrometer for health physics applications

    International Nuclear Information System (INIS)

    Brackenbush, L.W.; Reece, W.D.; Miller, S.D.

    1988-07-01

    Both the (ICRP) and the (NCPR) have recommended an increase in neutron quality factors and the adoption of effective dose equivalent methods. The series of reports entitled Personnel Neutron Dose Assessment Upgrade (PNL-6620) addresses these changes. Volume 1 in this series of reports (Personnel Neutron Dosimetry Assessment) provided guidance on the characteristics, use, and calibration of personnel neutron dosimeters in order to meet the new recommendations. This report, Volume 2: Field Neutron Spectrometer for Health Physics Applications describes the development of a portable field spectrometer which can be set up for use in a few minutes by a single person. The field spectrometer described herein represents a significant advance in improving the accuracy of neutron dose assessment. It permits an immediate analysis of the energy spectral distribution associated with the radiation from which neutron quality factor can be determined. It is now possible to depart from the use of maximum Q by determining and realistically applying a lower Q based on spectral data. The field spectrometer is made up of two modules: a detector module with built-in electronics and an analysis module with a IBM PC/reg sign/-compatible computer to control the data acquisition and analysis of data in the field. The unit is simple enough to allow the operator to perform spectral measurements with minimal training. The instrument is intended for use in steady-state radiation fields with neutrons energies covering the fission spectrum range. The prototype field spectrometer has been field tested in plutonium processing facilities, and has been proven to operate satisfactorily. The prototype field spectrometer uses a 3 He proportional counter to measure the neutron energy spectrum between 50 keV and 5 MeV and a tissue equivalent proportional counter (TEPC) to measure absorbed neutron dose

  5. Age-dependent conversion coefficients for organ doses and effective doses for external neutron irradiation

    International Nuclear Information System (INIS)

    Nishizaki, Chihiro; Endo, Akira; Takahashi, Fumiaki

    2006-06-01

    To utilize dose assessment of the public for external neutron irradiation, conversion coefficients of absorbed doses of organs and effective doses were calculated using the numerical simulation technique for six different ages (adult, 15, 10, 5 and 1 years and newborn), which represent the member of the public. Calculations were performed using six age-specific anthropomorphic phantoms and a Monte Carlo radiation transport code for two irradiation geometries, anterior-posterior and rotational geometries, for 20 incident energies from thermal to 20 MeV. Effective doses defined by the 1990 Recommendation of ICRP were calculated from the absorbed doses in 21 organs. The calculated results were tabulated in the form of absorbed doses and effective doses per unit neutron fluence. The calculated conversion coefficients are used for dose assessment of the public around nuclear facilities and accelerator facilities. (author)

  6. Time-Dependent Neutron and Photon Dose-Field Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wooten, Hasani Omar [Georgia Inst. of Technology, Atlanta, GA (United States)

    2005-08-01

    A unique tool is developed that allows the user to model physical representations of complicated glovebox facilities in two dimensions and determine neutral-particle flux and ambient dose-equivalent fields throughout that geometry. The Pandemonium code, originally designed to determine flux and dose-rates only, is improved to include realistic glovebox geometries, time-dependent source and detector positions, time-dependent shielding thickness calculations, time-integrated doses, a representative criticality accident scenario based on time-dependent reactor kinetics, and more rigorous photon treatment. A primary benefit of this work has been an extensive analysis and improvement of the photon model that is not limited to the application described in this thesis. The photon model has been extended in energy range to 10 MeV to include photons from fission and new photon buildup factors have been included that account for the effects of photon buildup at slant-path thicknesses as a function of angle, where the mean free path thickness has been preserved. The overall system of codes is user-friendly and it is directly applicable to facilities such as the plutonium facility at Los Alamos National Laboratory, where high-intensity neutron and photon emitters are regularly used. The codes may be used to determine a priori doses for given work scenarios in an effort to supply dose information to process models which will in turn assist decision makers on ensuring as low as reasonably achievable (ALARA) compliance. In addition, coupling the computational results of these tools with the process model visualization tools will help to increase worker safety and radiological safety awareness.

  7. Secondary neutron doses received by patients of different ages during intracranial proton therapy treatments

    International Nuclear Information System (INIS)

    Sayah, R.

    2012-01-01

    Proton therapy is an advanced radiation therapy technique that allows delivering high doses to the tumor while saving the healthy surrounding tissues due to the protons' ballistic properties. However, secondary particles, especially neutrons, are created during protons' nuclear reactions in the beam-line and the treatment room components, as well as inside the patient. Those secondary neutrons lead to unwanted dose deposition to the healthy tissues located at distance from the target, which may increase the secondary cancer risks to the patients, especially the pediatric ones. The aim of this work was to calculate the neutron secondary doses received by patients of different ages treated at the Institut Curie-centre de Protontherapie d'Orsay (ICPO) for intracranial tumors, using a 178 MeV proton beam. The treatments are undertaken at the new ICPO room equipped with an IBA gantry. The treatment room and the beam-line components, as well as the proton source were modeled using the Monte Carlo code MCNPX. The obtained model was then validated by a series of comparisons between model calculations and experimental measurements. The comparisons concerned: a) depth and lateral proton dose distributions in a water phantom, b) neutron spectrometry at one position in the treatment room, c) ambient dose equivalents at different positions in the treatment room and d) secondary absorbed doses inside a physical anthropomorphic phantom. A general good agreement was found between calculations and measurements, thus our model was considered as validated. The University of Florida hybrid voxelized phantoms of different ages were introduced into the MCNPX validated model, and secondary neutron doses were calculated to many of these phantoms' organs. The calculated doses were found to decrease as the organ's distance to the treatment field increases and as the patient's age increases. The secondary doses received by a one year-old patient may be two times higher than the doses

  8. 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

  9. Low-dose neutron dose response of zebrafish embryos obtained from the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility

    International Nuclear Information System (INIS)

    Ng, C.Y.P.; Kong, E.Y.; Konishi, T.; Kobayashi, A.; Suya, N.; Cheng, S.H.; Yu, K.N.

    2015-01-01

    The dose response of embryos of the zebrafish, Danio rerio, irradiated at 5 h post fertilization (hpf) by 2-MeV neutrons with ≤100 mGy was determined. The neutron irradiations were made at the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility in the National Institute of Radiological Sciences (NIRS), Chiba, Japan. A total of 10 neutron doses ranging from 0.6 to 100 mGy were employed (with a gamma-ray contribution of 14% to the total dose), and the biological effects were studied through quantification of apoptosis at 25 hpf. The responses for neutron doses of 10, 20, 25, and 50 mGy approximately fitted on a straight line, while those for neutron doses of 0.6, 1 and 2.5 mGy exhibited neutron hormetic effects. As such, hormetic responses were generically developed by different kinds of ionizing radiations with different linear energy transfer (LET) values. The responses for neutron doses of 70 and 100 mGy were significantly below the lower 95% confidence band of the best-fit line, which strongly suggested the presence of gamma-ray hormesis. - Highlights: • Neutron dose response was determined for embryos of the zebrafish, Danio rerio. • Neutron doses of 0.6, 1 and 2.5 mGy led to neutron hormetic effects. • Neutron doses of 70 and 100 mGy accompanied by gamma rays led to gamma-ray hormesis

  10. Occupational dose due to neutrons in medical linear accelerators

    International Nuclear Information System (INIS)

    Larcher, Ana M.; Bonet Duran, Stella M.; Lerner, Ana M.

    2000-01-01

    This paper describes a semi-empirical method to calculate the occupational dose due to neutrons and capture gamma rays in medical linear accelerators. It compares theoretical dose values with measurements performed in several 15 MeV medical accelerators installed in the country. Good agreement has been found between calculations made using the model and dose measurements, except for those accelerator rooms in which the maze length was shorter than the postulated tenth value distance. For those cases the model seems to overestimate neutron dose. The results demonstrate that the semi-empirical model is a good tool for quick and conservative shielding calculations for radiation protection purposes. Nevertheless, it is necessary to continue with the measurements in order to perform a more accurate validation of the model. (author)

  11. Device for measuring the dose rate of pulsed neutrons

    International Nuclear Information System (INIS)

    Klett, A.

    2009-01-01

    The author presents a new apparatus, developed in collaboration by Berthold Technologies and the German company DESY, allowing neutron pulsed fields to be measured. It is based on the activation by high energy neutrons of carbon 12 present in the sensor materials, and on the decay of short life radionuclides produced by this activation. The detection principle and system are briefly presented

  12. Verification of an effective dose equivalent model for neutrons

    International Nuclear Information System (INIS)

    Tanner, J.E.; Piper, R.K.; Leonowich, J.A.; Faust, L.G.

    1991-10-01

    Since the effective dose equivalent, based on the weighted sum of organ dose equivalents, is not a directly measurable quantity, it must be estimated with the assistance of computer modeling techniques and a knowledge of the radiation field. Although extreme accuracy is not necessary for radiation protection purposes, a few well-chosen measurements are required to confirm the theoretical models. Neutron measurements were performed in a RANDO phantom using thermoluminescent dosemeters, track etch dosemeters, and a 1/2-in. (1.27-cm) tissue equivalent proportional counter in order to estimate neutron doses and dose equivalents within the phantom at specific locations. The phantom was exposed to bare and D 2 O-moderated 252 Cf neutrons at the Pacific Northwest Laboratory's Low Scatter Facility. The Monte Carlo code MCNP with the MIRD-V mathematical phantom was used to model the human body and calculate organ doses and dose equivalents. The experimental methods are described and the results of the measurements are compared to the calculations. 8 refs., 3 figs., 3 tabs

  13. Measurements and calculations of neutron spectra and neutron dose distribution in human phantoms

    International Nuclear Information System (INIS)

    Palfalvi, J.

    1984-11-01

    The measurement and calculation of the radiation field around and in a phantom, with regard to the neutron component and the contaminating gamma radiation, are essential for radiation protection and radiotherapy purposes. The final report includes the development of the simple detector system, automized detector measuring facilities and a computerized evaluating system. The results of the depth dose and neutron spectra experiments and calculations in a human phantom are given

  14. Dose-effect relationship of apoptosis induced by fission-neutron in murine thymocytes

    International Nuclear Information System (INIS)

    Yuan Bin; Li Liang; Xue Wencheng; Sun Jianmin; Wang Baoqin

    2000-01-01

    Objective: To investigate the effectiveness of high LET fission-neutron to induce apoptosis in murine thymocytes and to compare it with that of low LET 60 Co γ-ray. Methods: Apoptosis induction was studied qualitatively by light and transmission electron microscopy and DNA gel electrophoresis,also quantitatively by flow cytometry(FCM) and diphenylamine (DPA)methods. Results: DNA ladders of murine thymocytes were detectable, the typical apoptosis of thymocytes could be observed morphologically by means of light and electron microscopy at 6 h after fission-neutron irradiation with doses ranging from 0.5 to 5.0 Gy, meanwhile the percentages of apoptosis increased with increasing doses. After exposure to γ-rays with doses ranging from 1.0 to 30 Gy, the experimental results were similar to those from neutron radiation. The incidence of apoptosis peaked at about 20 Gy, the percentages did not increase further when doses increased. Conclusion: Apoptosis of murine thymocytes can be induced when mice are exposed to either fission-neutron (0.5-5.0 Gy) or to γ-ray (1-30 Gy). Although the relationship between apoptosis and radiation doses is similar, the percentage of apoptosis induced by neutron irradiation is higher than that induced by γ-irradiation. The RBE values of fission-neutron for inducing apoptosis murine thymocytes are 2.09 (by FCM method) and 2.37 (by DPA method), respectively. These results also suggest that fission-neutron-induced murine immune tissue is more severe than that induced by γ-rays at several hours post-irradiation and this might be the basis for heavy damage to immune tissues induced by fission-neutron-irradiation in later period

  15. Dose Calibration of the ISS-RAD Fast Neutron Detector

    Science.gov (United States)

    Zeitlin, C.

    2015-01-01

    The ISS-RAD instrument has been fabricated by Southwest Research Institute and delivered to NASA for flight to the ISS in late 2015 or early 2016. ISS-RAD is essentially two instruments that share a common interface to ISS. The two instruments are the Charged Particle Detector (CPD), which is very similar to the MSL-RAD detector on Mars, and the Fast Neutron Detector (FND), which is a boron-loaded plastic scintillator with readout optimized for the 0.5 to 10 MeV energy range. As the FND is completely new, it has been necessary to develop methodology to allow it to be used to measure the neutron dose and dose equivalent. This talk will focus on the methods developed and their implementation using calibration data obtained in quasi-monoenergetic (QMN) neutron fields at the PTB facility in Braunschweig, Germany. The QMN data allow us to determine an approximate response function, from which we estimate dose and dose equivalent contributions per detected neutron as a function of the pulse height. We refer to these as the "pSv per count" curves for dose equivalent and the "pGy per count" curves for dose. The FND is required to provide a dose equivalent measurement with an accuracy of ?10% of the known value in a calibrated AmBe field. Four variants of the analysis method were developed, corresponding to two different approximations of the pSv per count curve, and two different implementations, one for real-time analysis onboard ISS and one for ground analysis. We will show that the preferred method, when applied in either real-time or ground analysis, yields good accuracy for the AmBe field. We find that the real-time algorithm is more susceptible to chance-coincidence background than is the algorithm used in ground analysis, so that the best estimates will come from the latter.

  16. Neutron spectrometry and determination of neutron ambient dose equivalents in different LINAC radiotherapy rooms

    International Nuclear Information System (INIS)

    Domingo, C.; Garcia-Fuste, M.J.; Morales, E.; Amgarou, K.; Terron, J.A.; Rosello, J.; Brualla, L.; Nunez, L.; Colmenares, R.; Gomez, F.; Hartmann, G.H.; Sanchez-Doblado, F.; Fernandez, F.

    2010-01-01

    A project has been set up to study the effect on a radiotherapy patient of the neutrons produced around the LINAC accelerator head by photonuclear reactions induced by photons above ∼8 MeV. These neutrons may reach directly the patient, or they may interact with the surrounding materials until they become thermalised, scattering all over the treatment room and affecting the patient as well, contributing to peripheral dose. Spectrometry was performed with a calibrated and validated set of Bonner spheres at a point located at 50 cm from the isocenter, as well as at the place where a digital device for measuring neutrons, based on the upset of SRAM memories induced by thermal neutrons, is located inside the treatment room. Exposures have taken place in six LINAC accelerators with different energies (from 15 to 23 MV) with the aim of relating the spectrometer measurements with the readings of the digital device under various exposure and room geometry conditions. The final purpose of the project is to be able to relate, under any given treatment condition and room geometry, the readings of this digital device to patient neutron effective dose and peripheral dose in organs of interest. This would allow inferring the probability of developing second malignancies as a consequence of the treatment. Results indicate that unit neutron fluence spectra at 50 cm from the isocenter do not depend on accelerator characteristics, while spectra at the place of the digital device are strongly influenced by the treatment room geometry.

  17. High precision thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Radeka, V.; Schaknowski, N.A.; Smith, G.C.; Yu, B. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    Two-dimensional position sensitive detectors are indispensable in neutron diffraction experiments for determination of molecular and crystal structures in biology, solid-state physics and polymer chemistry. Some performance characteristics of these detectors are elementary and obvious, such as the position resolution, number of resolution elements, neutron detection efficiency, counting rate and sensitivity to gamma-ray background. High performance detectors are distinguished by more subtle characteristics such as the stability of the response (efficiency) versus position, stability of the recorded neutron positions, dynamic range, blooming or halo effects. While relatively few of them are needed around the world, these high performance devices are sophisticated and fairly complex, their development requires very specialized efforts. In this context, we describe here a program of detector development, based on {sup 3}He filled proportional chambers, which has been underway for some years at the Brookhaven National Laboratory. Fundamental approaches and practical considerations are outlined that have resulted in a series of high performance detectors with the best known position resolution, position stability, uniformity of response and reliability over time, for devices of this type.

  18. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    International Nuclear Information System (INIS)

    Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

    2013-01-01

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18 O(p, n) 18 F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18 F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18 F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H 2 18 O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×10 6 %

  19. Neutron organ dose and the influence of adipose tissue

    Science.gov (United States)

    Simpkins, Robert Wayne

    Neutron fluence to dose conversion coefficients have been assessed considering the influences of human adipose tissue. Monte Carlo code MCNP4C was used to simulate broad parallel beam monoenergetic neutrons ranging in energy from thermal to 10 MeV. Simulated Irradiations were conducted for standard irradiation geometries. The targets were on gender specific mathematical anthropomorphic phantoms modified to approximate human adipose tissue distributions. Dosimetric analysis compared adipose tissue influence against reference anthropomorphic phantom characteristics. Adipose Male and Post-Menopausal Female Phantoms were derived introducing interstitial adipose tissue to account for 22 and 27 kg additional body mass, respectively, each demonstrating a Body Mass Index (BMI) of 30. An Adipose Female Phantom was derived introducing specific subcutaneous adipose tissue accounting for 15 kg of additional body mass demonstrating a BMI of 26. Neutron dose was shielded in the superficial tissues; giving rise to secondary photons which dominated the effective dose for Incident energies less than 100 keV. Adipose tissue impact on the effective dose was a 25% reduction at the anterior-posterior incidence ranging to a 10% increase at the lateral incidences. Organ dose impacts were more distinctive; symmetrically situated organs demonstrated a 15% reduction at the anterior-posterior Incidence ranging to a 2% increase at the lateral incidences. Abdominal or asymmetrically situated organs demonstrated a 50% reduction at the anterior-posterior incidence ranging to a 25% increase at the lateral incidences.

  20. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1982-11-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone TE-solutions, mineral oil and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. the OAR's measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. It is recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. A table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  1. Neutron activation analysis for calibration of phosphorus implantation dose

    International Nuclear Information System (INIS)

    Paul, Rick L.; Simons, David S.

    2001-01-01

    A feasibility study was undertaken to determine if radiochemical neutron activation analysis (RNAA) can be used to certify the retained dose of phosphorus implanted in silicon, with the goal of producing a phosphorus SRM. Six pieces of silicon, implanted with a nominal phosphorus dose of 8.5x10 14 atoms·cm -2 were irradiated at a neutron flux of 1.05x10 14 cm -2 ·s -1 . The samples were mixed with carrier, dissolved in acid, the phosphorus isolated by chemical separation, and 32 P measured using a beta proportional counter. A mean phosphorus concentration of (8.35±0.20)x10 14 atoms·cm -2 (uncertainty=1 standard deviation) was determined for the six samples, in agreement with the nominal implanted dose

  2. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H p (10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. (authors)

  3. Neutron and photon dose assessment in Indus accelerator complex

    International Nuclear Information System (INIS)

    Verma, Dimple; Haridas Nair, G.; Bandopadhyay, Tapas; Tripathy, R.M.; Pal, Rupali; Bakshi, A.K.; Palani Selvam, T.; Datta, D.

    2016-02-01

    Indus Accelerator Complex (IAC) consists of 20 MeV Microtron, 450/550 MeV Booster, 450 MeV Indus-1 and 2.5 GeV Indus-2 storage rings. The radiation environment in Indus Accelerator Complex comprises of bremsstrahlung photons, electrons, positrons, photo neutrons and muons, out of which, bremsstrahlung photons are the major constituent of the prompt radiation. Major problem faced for on-line detection of neutrons is their severely pulsed nature. In the present study, measurement of neutron and photon dose rates in Indus Accelerator Complex was carried out using passive dosimeters such as CR-39 solid state nuclear track detector (SSNTD) and CaSO 4 :Dy Teflon disc, 6 LiF:Mg,Ti (TLD 600) and 7 LiF:Mg,Ti (TLD 700) based thermo luminescent (TL) detectors. The report describes the details of the measurement and discusses the results. (author)

  4. Verification of an effective dose equivalent model for neutrons

    International Nuclear Information System (INIS)

    Tanner, J.E.; Piper, R.K.; Leonowich, J.A.; Faust, L.G.

    1992-01-01

    Since the effective dose equivalent, based on the weighted sum of organ dose equivalents, is not a directly measurable quantity, it must be estimated with the assistance of computer modelling techniques and a knowledge of the incident radiation field. Although extreme accuracy is not necessary for radiation protection purposes, a few well chosen measurements are required to confirm the theoretical models. Neutron doses and dose equivalents were measured in a RANDO phantom at specific locations using thermoluminescence dosemeters, etched track dosemeters, and a 1.27 cm (1/2 in) tissue-equivalent proportional counter. The phantom was exposed to a bare and a D 2 O-moderated 252 Cf neutron source at the Pacific Northwest Laboratory's Low Scatter Facility. The Monte Carlo code MCNP with the MIRD-V mathematical phantom was used to model the human body and to calculate the organ doses and dose equivalents. The experimental methods are described and the results of the measurements are compared with the calculations. (author)

  5. Personnel neutron dose assessment upgrade: Volume 1, Personnel neutron dosimetry assessment: [Final report

    International Nuclear Information System (INIS)

    Hadlock, D.E.; Brackenbush, L.W.; Griffith, R.V.; Hankins, D.E.; Parkhurst, M.A.; Stroud, C.M.; Faust, L.G.; Vallario, E.J.

    1988-07-01

    This report provides guidance on the characteristics, use, and calibration criteria for personnel neutron dosimeters. The report is applicable for neutrons with energies ranging from thermal to less than 20 MeV. Background for general neutron dosimetry requirements is provided, as is relevant federal regulations and other standards. The characteristics of personnel neutron dosimeters are discussed, with particular attention paid to passive neutron dosimetry systems. Two of the systems discussed are used at DOE and DOE-contractor facilities (nuclear track emulsion and thermoluminescent-albedo) and another (the combination TLD/TED) was recently developed. Topics discussed in the field applications of these dosimeters include their theory of operation, their processing, readout, and interpretation, and their advantages and disadvantages for field use. The procedures required for occupational neutron dosimetry are discussed, including radiation monitoring and the wearing of dosimeters, their exchange periods, dose equivalent evaluations, and the documenting of neutron exposures. The coverage of dosimeter testing, maintenance, and calibration includes guidance on the selection of calibration sources, the effects of irradiation geometries, lower limits of detectability, fading, frequency of calibration, spectrometry, and quality control. 49 refs., 6 figs., 8 tabs

  6. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1983-01-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone tissue-equivalent (TE) solutions, mineral oil, and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU Report No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. The OARs measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. Therefore, neutron beam CADDs and OARs may be measured in either TE solution (USA practice) or water (European practice), and having determined the respective scaling lengths, all measurements may be scaled from one medium to any other. It is recommended that for general treatment planning purposes, scaling be made to TE muscle with a density of 1.04 g cm -3 , since this value represents muscle and other soft tissues better than TE solution of density 1.07 g cm -3 . For such a transformation, relative measurements made in water are found to require very small corrections. Hence, it is further recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. Finally, a table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  7. Life shortening, tumor induction, and tissue dose for fission-neutron and gamma-ray irradiations

    International Nuclear Information System (INIS)

    Grahn, D.; Duggal, K.; Lombard, L.S.

    1985-01-01

    The primary focus of this program is to obtain information on the late effects of whole body exposure to low doses of a high linear-energy-transfer (LET) and a low-LET ionizing radiation in experimental animals to provide guidance for the prediction of radiation hazards to man. The information obtained takes the form of dose-response curves for life shortening and for the induction of numerous specific types of tumors. The animals are irradiated with fission neutrons from the Janus reactor and with 60 Co gamma rays, delivered as single, weekly, or duration-of-life exposures covering the range of doses and dose rates. 6 refs

  8. Experimental investigations of the neutron contamination in high-energy photon fields at medical linear accelerators

    International Nuclear Information System (INIS)

    Brunckhorst, Elin

    2009-01-01

    The scope of this thesis was to develop a device for the detection of the photoneutron dose inside the high-energy photon field. The photoneutron contamination of a Siemens PRIMUS linear accelerator was investigated in detail in its 15 MV photon mode. The experimental examinations were performed with three ionisation chambers (a tissue equivalent chamber, a magnesium chamber and a 10 B-coated magnesium chamber) and two types of thermoluminescence detectors (enriched with 6 Li and 7 Li, respectively). The detectors have different sensitivities to photons and neutrons and their combination allows the dose separation in a mixed neutron/photon field. The application of the ionisation chamber system, as well as the present TLD system for photoneutron detection in high-energy photon beams is a new approach. The TLD neutron sensitivity was found to be too low for a measurement inside the open photon field and the further investigation focused on the ionisation chambers. The three ionisation chambers were calibrated at different photon and neutron sources and a the borated magnesium chamber showed a very high response to thermal neutrons. For a cross check of the calibration, the three chambers were also used for dose separation of a boron neutron capture therapy beam where the exact determination of the thermal neutron dose is essential. Very accurate results were achieved for the thermal neutron dose component. At the linear accelerator the chamber system was reduced to a paired chamber system utilising the two magnesium chambers, since the fast neutron component was to small to be separated. The neutron calibration of the three chambers could not be applied, instead a conversion of measured thermal neutron signal by the borated chamber to Monte Carlo simulated total neutron dose was performed. Measurements for open fields in solid water and liquid water were performed with the paired chamber system. In larger depths the neutron dose could be determined with an

  9. Experimental investigations of the neutron contamination in high-energy photon fields at medical linear accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Brunckhorst, Elin

    2009-02-26

    The scope of this thesis was to develop a device for the detection of the photoneutron dose inside the high-energy photon field. The photoneutron contamination of a Siemens PRIMUS linear accelerator was investigated in detail in its 15 MV photon mode. The experimental examinations were performed with three ionisation chambers (a tissue equivalent chamber, a magnesium chamber and a {sup 10}B-coated magnesium chamber) and two types of thermoluminescence detectors (enriched with {sup 6}Li and {sup 7}Li, respectively). The detectors have different sensitivities to photons and neutrons and their combination allows the dose separation in a mixed neutron/photon field. The application of the ionisation chamber system, as well as the present TLD system for photoneutron detection in high-energy photon beams is a new approach. The TLD neutron sensitivity was found to be too low for a measurement inside the open photon field and the further investigation focused on the ionisation chambers. The three ionisation chambers were calibrated at different photon and neutron sources and a the borated magnesium chamber showed a very high response to thermal neutrons. For a cross check of the calibration, the three chambers were also used for dose separation of a boron neutron capture therapy beam where the exact determination of the thermal neutron dose is essential. Very accurate results were achieved for the thermal neutron dose component. At the linear accelerator the chamber system was reduced to a paired chamber system utilising the two magnesium chambers, since the fast neutron component was to small to be separated. The neutron calibration of the three chambers could not be applied, instead a conversion of measured thermal neutron signal by the borated chamber to Monte Carlo simulated total neutron dose was performed. Measurements for open fields in solid water and liquid water were performed with the paired chamber system. In larger depths the neutron dose could be determined

  10. Dose modification factors in boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Allen, B.J. (Australian Nuclear Science and Technology Organization (ANSTO), Menai (Australia))

    1993-01-01

    The effective treatment depth and therapeutic ratio in boron neutron capture therapy (BNCT) depend on a number of macroscopic dose factors such as boron concentrations in the tumor, normal tissue and blood. However, the role of various microscopic dose modification factors can be of critical importance in the evaluation of normal tissue tolerance levels. An understanding of these factors is valuable in designing BNCT experiments and the selection of appropriate boron compounds. These factors are defined in this paper and applied to the case of brain tumors with particular attention to capillary endothelial cells and oligodendrocytes. (orig.).

  11. Skin Dose Equivalent Measurement from Neutron-Deficient Isotopes

    International Nuclear Information System (INIS)

    Hsu, Hsiao-Hua; Costigan, Steve A.; Romero, Leonard L.; Whicker, Jeffrey J.

    1997-12-01

    Neutron-deficient-isotopes decay via positron emission and/or electron capture often followed by x-ray, gamma-ray, and 0.511 MeV photons from positron annihilation. For cases of significant area and/or personnel contamination with these isotopes, determination of skin dose equivalent (SDE) is required by 10CFR835. For assessment of SDE, we evaluated the MICROSPEC-2(TM) system manufactured by Bubble Technology Industries of Canada which uses three different probes for dose measurement. We used two probes: (1) the X-probe which measures lower energy (4 - 120 keV) photon energy distributions and determines deep dose equivalent, SDE and dose equivalent to eyes, and (2) the B-probe which measures electron (positron) energy distributions, and determines skin dose equivalent. Also, the measured photon and beta spectra can be used to identify radioactive isotopes in the contaminated area. Measurements with several neutron-deficient sources showed that this system provided reasonably accurate SDE rate measurements when compared with calculated benchmark SDE rates with an average percent difference of 40%. Variations were expected because of differences between the assumed geometries used by MlCROSPEC-2 and the calculations when compared to the measurement conditions

  12. 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

  13. Gamma dosimetry of high doses

    International Nuclear Information System (INIS)

    Martinez C, T.; Galvan G, A.; Canizal, G.

    1991-01-01

    The gamma dosimetry of high doses is problematic in almost all the classic dosemeters either based on the thermoluminescence, electric, chemical properties, etc., because they are saturated to very high dose and they are no longer useful. This work carries out an investigation in the interval of high doses. The solid system of heptahydrate ferrous sulfate, can be used as solid dosemeter of routine for high doses of radiation. The proposed method is simple, cheap and it doesn't require sophisticated spectrophotometers or spectrometers but expensive and not common in some laboratories

  14. Some thoughts on tolerance, dose, and fractionation in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Gahbauer, R.; Goodman, J.; Blue, T.

    1988-01-01

    Unique to boron neutron capture therapy, the tolerance very strongly depends on the boron concentration in normal brain, skin and blood. If one first considers the ideal situation of a 2 KeV beam and a compound clearing from normal tissues and blood, the tolerance dose to epithermal beams relates to the maximum tolerated capture gamma dose and capture high LET dose, H (n,gamma)D and N(n,p) 14 C. The authors can relate this gamma and high LET dose to known clinical experience. Assuming gamma and high LET dose ratios as given by Fairchild and Bond, one may first choose a clearly safe high LET whole brain dose and calculate the unavoidably resulting gamma dose. To a first approximation 500 cGy of high LET dose results in 3,000 cGy gamma dose. One can speculate that this approximates the tolerance of whole brain to the 2 KeV beam with no contributing boron dose if the radiation is fractionated. It would clearly be beyond tolerance in a single fraction where most therapists would be uncomfortable to deliver even one third of the above doses

  15. Low level neutron monitoring using high pressure 3He detectors

    International Nuclear Information System (INIS)

    Pszona, S.

    1995-01-01

    Three detectors, two spherical proportional counters and an ionisation chamber, all filled with 3 He to pressures of 160 kPa, 325 kPa and 1 MPa respectively have been experimentally studied with respect to their use for low level neutron monitoring. The ambient dose equivalent responses and the energy resolutions of these detectors have been determined. It is shown that spectral analysis of the signals from these detectors not only gives high sensitivity with regard to ambient dose equivalent but also improves the quality of the measurements. A special instrumentation for low level neutron monitoring is described in which a quality control method has been implemented. (Author)

  16. Evaluation of mixed energy neutron doses using TLD NG-67 type

    International Nuclear Information System (INIS)

    Akhadi, Mukhlis; Thoyib Thamrin, M; Usmiyati Dewi, K.

    2000-01-01

    A research has been carried out to develop dose evaluation method of mixed neutron source with its neutron doses can be classified to two groups, I.e neutron doses with energy ≥ 0.5 eV and thermal neutron doses with energy less than 0.5 e V consist of epithermal and fast neutron, but in this research they were classified as fast neutron. Development of this dose evaluation method was carried out by sensitivity (S) intercomparison of TLD-600 to fast neutron, mixed energy neutron of nuclear rectors, and thermal neutron. From the experiment it was obtained that the value of Sfast : Sreactor : Sthermal = 0.005 : 0.010 : 1. Calibration factor (CF) of TLD is defined as 1/S. from the sensitivity data it can be obtained that the value of Cffast : Cfreactor : Cfthermal = 200 :100 : 1. The value of Cfreactor can be applied for mixed energy neutron doses evaluation of TLD-600. Key word : dosemeter, neutron dose, calibration factor, fast neutron, thermal neutron, nuclear reactor

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  18. Effects of secondary interactions on the dose calculation in treatments with Boron Neutron Capture Therapy (BNCT)

    International Nuclear Information System (INIS)

    Monteiro, E.

    2004-01-01

    The aimed of this work consists of evaluating the influence of the secondary contributions of dose (thermal neutrons dose, epithermal neutrons dose, fast neutrons dose and photon dose) in treatment planning with BNCT. MCNP4B Code was used to calculate RBE-Gy doses through the irradiation of the modified Snyder head head phantom.A reduction of the therapeutical gain of monoenergetic neutron beans was observed in non invasive treatments, provoked for the predominance of the fast neutron dose component in the skin, showing that the secondary contributions of dose can contribute more in the direction to raise the dose in the fabric healthy that in the tumor, thus reducing the treatment efficiency. (author)

  19. Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

    International Nuclear Information System (INIS)

    Marinkovic, P.

    1991-01-01

    Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)

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

    International Nuclear Information System (INIS)

    Rimpler, A.; Kneschke, H.

    1985-01-01

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

  1. Development of an anthropomorfic simulator for simulation and measurements of neutron dose and flux the facility for BNCT studies

    International Nuclear Information System (INIS)

    Muniz, Rafael Oliveira Rondon

    2010-01-01

    IPEN facility for researches in BNCT (Boron Neutron Capture Therapy) uses IEA-R1 reactor's irradiation channel number 3, where there is a mixed radiation field - neutrons and gamma. The researches in progress require the radiation fields, in the position of the irradiation of sample, to have in its composition maximized thermal neutrons component and minimized, fast and epithermal neutron flux and gamma radiation. This work was developed with the objective of evaluating whether the present radiation field in the facility is suitable for BNCT researches. In order to achieve this objective, a methodology for the dosimetry of thermal neutrons and gamma radiation in mixed fields of high doses, which was not available in IPEN, was implemented in the Center of Nuclear Engineering of IPEN, by using thermoluminescent dosimeters - TLDs 400, 600 and 700. For the measurements of thermal and epithermal neutron flux, activation detectors of gold were used applying the cadmium ratio technique. A cylindrical phantom composed by acrylic discs was developed and tested in the facility and the DOT 3.5. computational code was used in order to obtain theoretical values of neutron flux and the dose along phantom. In the position corresponding to about half the length of the cylinder of the phantom, the following values were obtained: thermal neutron flux (2,52 ± 0,06).10 8 n/cm 2 s, epithermal neutron flux (6,17 ± 0,26).10 7 .10 6 n/cm 2 s, absorbed dose due to thermal neutrons (4,2 ± 1,8)Gy and (10,1 ± 1,3)Gy due to gamma radiation. The obtained values show that the fluxes of thermal and epithermal neutrons flux are appropriate for studies in BNCT, however, the dose due to gamma radiation is high, indicating that the facility should be improved. (author)

  2. Temperature and neutron dose rate measurements at a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Krause, F.

    1982-01-01

    Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

  3. High intensity TOF spectrometer for cold neutrons

    International Nuclear Information System (INIS)

    Maayouf, R.M.; Abd El-Kawy, A.; Habib, N.; Adib, M.; Hamouda, I.

    1984-01-01

    This work presents a neutron time-of-flight (TOF) spectrometer developed specially for total neutron cross-section measurements at neutron energies below 5 MeV and sample's temperature varying from the liquid nitrogen one and up to 500 0 K. The spectrometer is equipped by remote control unit, designed especially, in order to move the sample in and out of the beam during the experimental measurements. The spectrometer has proved to be useful for transmission measurements at neutron energies below 5 MeV. It has a reasonable energy resolution (4.4%) and high effect to background ratio (11.1) at 5 MeV

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

    Science.gov (United States)

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

    2018-02-01

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

  5. Three-dimensional neutron dose distribution in the environment around a 1-GeV electron synchrotron facility at INS

    International Nuclear Information System (INIS)

    Uwamino, Y.; Nakamura, T.

    1987-01-01

    The three-dimensional (surface and altitude) skyshine neutron-dose-equivalent distribution around the 1-GeV electron synchrotron (ES) of the Institute for Nuclear Study, University of Tokyo, was measured with a high-sensitivity dose-equivalent counter. The neutron spectrum in the environment was also measured with a multimoderator spectrometer incorporating a 3 He counter. The dose-equivalent distribution and the leakage neutron spectrum at the surface of the ES building were measured with a Studsvik 2202D counter and the multimoderator spectrometer, including an indium activation detector. Skyshine neutron transport calculations, beginning with the photoneutron spectrum and yielding the dose-equivalent distribution in the environment, were performed with the DOT3.5 code and two Monte Carlo codes, MMCR-2 and MMCR-3, using the DLC-87/HILO group cross sections. The calculated neutron spectra at the top surface of the concrete ceiling and at a point 111 m from the ES agreed well with the measured results, and the calculated three-dimensional dose-equivalent distribution also agreed. The dose value increased linearly with altitude, and the slope was estimated for neutron-producing facilities. (author)

  6. Transport of accelerator produced high energy neutrons though concrete

    International Nuclear Information System (INIS)

    Prabhakar Rao, G.; Sarkar, P.K.

    1996-01-01

    Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs

  7. High efficiency focus neutron generator

    Science.gov (United States)

    Sadeghi, H.; Amrollahi, R.; Zare, M.; Fazelpour, S.

    2017-12-01

    In the present paper, the new idea to increase the neutron yield of plasma focus devices is investigated and the results are presented. Based on many studies, more than 90% of neutrons in plasma focus devices were produced by beam target interactions and only 10% of them were due to thermonuclear reactions. While propounding the new idea, the number of collisions between deuteron ions and deuterium gas atoms were increased remarkably well. The COMSOL Multiphysics 5.2 was used to study the given idea in the known 28 plasma focus devices. In this circumstance, the neutron yield of this system was also obtained and reported. Finally, it was found that in the ENEA device with 1 Hz working frequency, 1.1 × 109 and 1.1 × 1011 neutrons per second were produced by D-D and D-T reactions, respectively. In addition, in the NX2 device with 16 Hz working frequency, 1.34 × 1010 and 1.34 × 1012 neutrons per second were produced by D-D and D-T reactions, respectively. The results show that with regards to the sizes and energy of these devices, they can be used as the efficient neutron generators.

  8. Response to annealing and reirradiation of AISI 304L stainless steel following initial high-dose neutron irradiation in EBR-II

    International Nuclear Information System (INIS)

    Porter, D.L.; McVay, G.L.; Walters, L.C.

    1980-01-01

    The object of this study was to measure the stability of irradiation-induced microstructure upon annealing and, by selectively annealing out some of these features and reirradiating the material, it was expected that information could be gained concerning the role of microstructural changes in the void swelling process. Transmission electron microscopic examinations of isochronally annealed (200 to 1050 0 C) AISI 304L stainless steel, which had been irradiated at approximately 415 0 C to a fast (E > 0.1 MeV) neutron fluence of approximately 5.1 x 10 26 n/m 2 , verified that the two-stage hardness recovery with temperatures was related to a low temperature annealing of dislocation structures and a higher temperature annealing of voids and solute redistribution

  9. Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations.

    Science.gov (United States)

    Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

    2014-01-01

    High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.

  10. Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-03-15

    From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

  11. Assessment of doses due to secondary neutrons received by patient treated by proton therapy

    International Nuclear Information System (INIS)

    Sayah, R.; Martinetti, F.; Donadille, L.; Clairand, I.; Delacroix, S.; De Oliveira, A.; Herault, J.

    2010-01-01

    Proton therapy is a specific technique of radiotherapy which aims at destroying cancerous cells by irradiating them with a proton beam. Nuclear reactions in the device and in the patient himself induce secondary radiations involving mainly neutrons which contribute to an additional dose for the patient. The author reports a study aimed at the assessment of these doses due to secondary neutrons in the case of ophthalmological and intra-cranial treatments. He presents a Monte Carlo simulation of the room and of the apparatus, reports the experimental validation of the model (dose deposited by protons in a water phantom, ambient dose equivalent due to neutrons in the treatment room, absorbed dose due to secondary particles in an anthropomorphic phantom), and the assessment with a mathematical phantom of doses dues to secondary neutrons received by organs during an ophthalmological treatment. He finally evokes current works of calculation of doses due to secondary neutrons in the case of intra-cranial treatments

  12. The radiobiology of boron neutron capture therapy: Are ''photon-equivalent'' doses really photon-equivalent?

    International Nuclear Information System (INIS)

    Coderre, J.A.; Diaz, A.Z.; Ma, R.

    2001-01-01

    Boron neutron capture therapy (BNCT) produces a mixture of radiation dose components. The high-linear energy transfer (LET) particles are more damaging in tissue than equal doses of low-LET radiation. Each of the high-LET components can multiplied by an experimentally determined factor to adjust for the increased biological effectiveness and the resulting sum expressed in photon-equivalent units (Gy-Eq). BNCT doses in photon-equivalent units are based on a number of assumptions. It may be possible to test the validity of these assumptions and the accuracy of the calculated BNCT doses by 1) comparing the effects of BNCT in other animal or biological models where the effects of photon radiation are known, or 2) if there are endpoints reached in the BNCT dose escalation clinical trials that can be related to the known response to photons of the tissue in question. The calculated Gy-Eq BNCT doses delivered to dogs and to humans with BPA and the epithermal neutron beam of the Brookhaven Medical Research Reactor were compared to expected responses to photon irradiation. The data indicate that Gy-Eq doses in brain may be underestimated. Doses to skin are consistent with the expected response to photons. Gy-Eq doses to tumor are significantly overestimated. A model system of cells in culture irradiated at various depths in a lucite phantom using the epithermal beam is under development. Preliminary data indicate that this approach can be used to detect differences in the relative biological effectiveness of the beam. The rat 9L gliosarcoma cell survival data was converted to photon-equivalent doses using the same factors assumed in the clinical studies. The results superimposed on the survival curve derived from irradiation with Cs-137 photons indicating the potential utility of this model system. (author)

  13. SHINE-III. Simple code for skyshine dose calculation up to 3 GeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tsukiyama, Toshihisa; Tayama, Ryuichi; Handa, Hiroyuki [Hitachi Engineering Co. Ltd., Ibaraki (Japan)] [and others

    2000-03-01

    Skyshine dose at site boundary is considered as one of the most fundamental issues to get approval of constructing nuclear installations. Skyshine conical beam response functions (CBRF) for high energy neutrons up to 3 GeV are obtained using NMTC-JAERI and MCNP code. This CBRF is fitted to the four parameters equation. Simple code named SHINE-III using this equation with updated data is developed. (author)

  14. Characteristics of neutron irradiation facility and dose estimation method for neutron capture therapy at Kyoto University research reactor institute

    International Nuclear Information System (INIS)

    Kobayashi, T.; Sakurai, Y.; Kanda, K.

    2001-01-01

    The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)

  15. Out‐of‐field doses and neutron dose equivalents for electron beams from modern Varian and Elekta linear accelerators

    Science.gov (United States)

    Cardenas, Carlos E.; Nitsch, Paige L.; Kudchadker, Rajat J.; Howell, Rebecca M.

    2016-01-01

    Out‐of‐field doses from radiotherapy can cause harmful side effects or eventually lead to secondary cancers. Scattered doses outside the applicator field, neutron source strength values, and neutron dose equivalents have not been broadly investigated for high‐energy electron beams. To better understand the extent of these exposures, we measured out‐of‐field dose characteristics of electron applicators for high‐energy electron beams on two Varian 21iXs, a Varian TrueBeam, and an Elekta Versa HD operating at various energy levels. Out‐of‐field dose profiles and percent depth‐dose curves were measured in a Wellhofer water phantom using a Farmer ion chamber. Neutron dose was assessed using a combination of moderator buckets and gold activation foils placed on the treatment couch at various locations in the patient plane on both the Varian 21iX and Elekta Versa HD linear accelerators. Our findings showed that out‐of‐field electron doses were highest for the highest electron energies. These doses typically decreased with increasing distance from the field edge but showed substantial increases over some distance ranges. The Elekta linear accelerator had higher electron out‐of‐field doses than the Varian units examined, and the Elekta dose profiles exhibited a second dose peak about 20 to 30 cm from central‐axis, which was found to be higher than typical out‐of‐field doses from photon beams. Electron doses decreased sharply with depth before becoming nearly constant; the dose was found to decrease to a depth of approximately E(MeV)/4 in cm. With respect to neutron dosimetry, Q values and neutron dose equivalents increased with electron beam energy. Neutron contamination from electron beams was found to be much lower than that from photon beams. Even though the neutron dose equivalent for electron beams represented a small portion of neutron doses observed under photon beams, neutron doses from electron beams may need to be considered for

  16. Boron neutron capture irradiation of the rat spinal cord: effects of variable doses of borocaptate sodium

    International Nuclear Information System (INIS)

    Morris, Gerard M.; Coderre, Jeffrey A.; Hopewell, John W.; Micca, Peggy L.; Fisher, Craig

    1996-01-01

    The Fischer 344 rat spinal cord model has been used to evaluate the response of the central nervous system to boron neutron capture irradiation with variable doses of the neutron capture agent, borocaptate sodium (BSH). Three doses of BSH, 190, 140 and 80 mg/kg body weight, administered by i.p. injection, were used to establish the time course of 10 B accumulation in and removal from the blood. After administration of the two lower doses of BSH, blood 10 B levels peaked at 0.5 h after injection, with no significant (P > 0.1) change at 1 h after injection. Beyond this time point, levels of 10 B in the blood began to decrease after a dose of 80 mg/kg BSH, but remained constant until 3 h after administration after the two higher doses of BSH. Myelopathy developed after latent intervals of 20.4 ± 0.1, 20.8 ± 1.4, 15.0 ± 0.8, 15.4 ± 0.4 and 15.6 ± 0.4 weeks, following irradiation with thermal neutrons in combination with BSH at doses of 20, 40, 80, 140 and 190 mg/kg body weight, respectively. The radiation-induced lesion in the spinal cord was white matter necrosis. ED 50 values for myelopathy were calculated from probit-fitted dose-effect curves. Expressed as total physical absorbed doses, these values were 20.7 ± 1.9, 24.9 ± 1.2, 27.2 ± 0.9, 28.4 ± 0.6 and 32.4 ± 1.9 Gy after irradiation with thermal neutrons in the presence of 20, 40, 80, 140 and 190 mg/kg body weight of BSH, respectively. The compound biological effectiveness (CBE) factor values, estimated from this data, were in the range 0.49-0.55. There was no significant (P >0.1) variation in the CBE factor for BSH as a function of increasing 10 B concentration in the blood. It was concluded that there was no significant synergistic interaction between the low and high linear energy transfer (LET) components of the boron neutron capture (BNC) radiation field

  17. Techniques in high pressure neutron scattering

    CERN Document Server

    Klotz, Stefan

    2013-01-01

    Drawing on the author's practical work from the last 20 years, Techniques in High Pressure Neutron Scattering is one of the first books to gather recent methods that allow neutron scattering well beyond 10 GPa. The author shows how neutron scattering has to be adapted to the pressure range and type of measurement.Suitable for both newcomers and experienced high pressure scientists and engineers, the book describes various solutions spanning two to three orders of magnitude in pressure that have emerged in the past three decades. Many engineering concepts are illustrated through examples of rea

  18. High resolution neutron spectroscopy for helium isotopes

    International Nuclear Information System (INIS)

    Abdel-Wahab, M.S.; Klages, H.O.; Schmalz, G.; Haesner, B.H.; Kecskemeti, J.; Schwarz, P.; Wilczynski, J.

    1992-01-01

    A high resolution fast neutron time-of-flight spectrometer is described, neutron time-of-flight spectra are taken using a specially designed TDC in connection to an on-line computer. The high time-of-flight resolution of 5 ps/m enabled the study of the total cross section of 4 He for neutrons near the 3/2 + resonance in the 5 He nucleus. The resonance parameters were determined by a single level Breit-Winger fit to the data. (orig.)

  19. Optimization of a neutron ambient dose equivalent rate meter

    International Nuclear Information System (INIS)

    Burgkhardt, B.; Fieg, G.; Piesch, E.; Klett, A.; Maushart, R.

    1994-01-01

    Co-operating in a technology transfer project, the Karlsruhe Nuclear Research Center and EG and G Berthold have developed a neutron equivalent doserate probe with high sensitivity and an energy dependent detection efficiency in accordance with the ICRP60 requirements. The special features of this probe were realized, on the one hand, by optimizing the moderator and absorber geometry through simulation calculations with the neutron transport code MCNP, and, on the other hand, by using a newly designed 3 He-methane proportional counter tube. The measurements were not yet completed when this paper went to press, however, it is to be expected that the response sensitivity will be more than 3 counts/nSv. (orig.) [de

  20. Analysis of neutron dose rates on RGTT200K core using MCNP5

    International Nuclear Information System (INIS)

    Suwoto; Zuhair

    2016-01-01

    The conceptual design of RGTT200K (High Temperature Gas-cooled Reactor of 200 MWth Cogeneration) is the non-annular cylindrical reactor core with TRISO kernel coated fuel particles in the form of balls called pebble and cooled by helium gas. The RGTT200K reactor core design adopts high temperature gas cooled reactor (HTGR) technology with inherent passive safety. The RGTT200K spherical fuel called pebble fuel containing thousand of TRISO-coated fuel particles of uranium oxide (UO 2 ) 10 % enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner pyrolytic carbon layer (IPyC, Inner Pyrolytic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer portion (OPyC, Outer Pyrolytic Carbon). Modeling and analysis of preliminary calculation of neutron dose rate on normal operating temperature (T kernel =1200K) and accident temperature (T kernel =1800K) of the RGTT200K core were performed using Monte Carlo MCNP5v1.2 code. The continuous energy nuclear data cross-sections was taken from ENDF/B-VII, JENDL-4 and JEFF-3.1 nuclear data files . Double heterogeneity model in TRISO-coated fuel particles kernel and the pebble of RGTT200K core. By utilizing EGS99304 code, the 640 amount of energy group structures (SAND-II neutron group structures) is used in the neutron fluxes and spectrum calculation in RGTT200K reactor. The RGTT200K reactor core is divided into 25 zones (5 zones in radial and 10 zones in axial directions), while the modeling of radiation and biological shielding reactor RGTT200K are used to determine of preliminary neutron dose rate emitted by the neutron source with tally cards are available in the MCNP5v1.2 code. The calculation result analyses of the neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection, ICRP. The preliminary calculations result show that the neutrons dose rate using ICRP-74 conversion factor for

  1. Neutron and gamma dose and spectra measurements on the Little Boy replica

    International Nuclear Information System (INIS)

    Hoots, S.; Wadsworth, D.

    1984-01-01

    The radiation-measurement team of the Weapons Engineering Division at Lawrence Livermore National Laboratory (LLNL) measured neutron and gamma dose and spectra on the Little Boy replica at Los Alamos National Laboratory (LANL) in April 1983. This assembly is a replica of the gun-type atomic bomb exploded over Hiroshima in 1945. These measurements support the National Academy of Sciences Program to reassess the radiation doses due to atomic bomb explosions in Japan. Specifically, the following types of information were important: neutron spectra as a function of geometry, gamma to neutron dose ratios out to 1.5 km, and neutron attenuation in the atmosphere. We measured neutron and gamma dose/fission from close-in to a kilometer out, and neutron and gamma spectra at 90 and 30 0 close-in. This paper describes these measurements and the results. 12 references, 13 figures, 5 tables

  2. Preliminary characterization of the passive neutron dose equivalent monitor with TLDs

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, Norio; Kanai, Katsuta; Momose, Takumaro; Hayashi, Naomi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan); Chen Erhu [Beijing Institute of Nuclear Engineering, Beijing (China)

    2001-02-01

    The passive neutron dose equivalent monitor with TLDs is composed of a cubic polyethylene moderator and TLDs at the center of moderator. This monitor was originally designed for measurements of neutron doses over long-term period of time around the nuclear facilities. In this study, the energy response of this monitor was calculated by Monte Carlo methods and experimentally obtained under {sup 241}Am-Be, {sup 252}Cf and moderated {sup 252}Cf neutron irradiation. Additionally, the responses of two types of conventional neutron dose equivalent meters (rem counters) were also investigated as comparison. The authors concluded that this passive neutron monitor with TLDs had a good energy response similar to conventional rem counters and could evaluate neutron doses within 10% of accuracy to the moderated fission spectra. (author)

  3. Neutron dose equivalent next to the target shield of a neutron therapy facility using an LET counter

    International Nuclear Information System (INIS)

    Stinchcomb, T.G.; Kuchnir, F.T.

    1981-01-01

    The use of a spherical tissue-equivalent proportional counter for measurements of the lineal energy (y) and derivations of the linear energy transfer (LET) for fast neutrons has the advantage of giving distributions of dose and dose equivalent as functions of either LET or y. A measurement next to the target shielding of the neutron therapy facility at the University of Chicago Hospitals and Clinics (UCHC) is described, and the data processing is outlined. The distributions are presented and compared to those from measurements in the neutron beam. The average quality factors are presented

  4. Measurement of stray neutron doses inside the treatment room from a proton pencil beam scanning system

    Czech Academy of Sciences Publication Activity Database

    Mojzeszek, N.; Farah, J.; Klodowska, M.; Ploc, Ondřej; Stolarczyk, L.; Waligorski, M. P. R.; Olko, P.

    2017-01-01

    Roč. 34, č. 2 (2017), s. 80-84 ISSN 1120-1797 Institutional support: RVO:61389005 Keywords : secondary neutrons * proton therapy * pencil beam scanning systtems * out-of-field doses * stray neutron doses * TEPC Subject RIV: FP - Other Medical Disciplines OBOR OECD: Radiology, nuclear medicine and medical imaging Impact factor: 1.990, year: 2016

  5. Genetic effects induced by neutrons in Drosophila melanogaster I. Determination of absorbed dose

    International Nuclear Information System (INIS)

    Delfin, A.; Paredes, L.C.; Zambrano, F.; Guzman-Rincon, J.; Urena-Nunez, F.

    2001-01-01

    A method to obtain the absorbed dose in Drosophila melanogaster irradiated in the thermal column facility of the Triga Mark III Reactor has been developed. The method is based on the measurements of neutron activation of gold foils produced by neutron capture to obtain the neutron fluxes. These fluxes, combined with the calculations of kinetic energy released per unit mass, enables one to obtain the absorbed doses in Drosophila melanogaster

  6. Neutron relative biological effectiveness for solid cancer incidence in the Japanese A-bomb survivors: an analysis considering the degree of independent effects from γ-ray and neutron absorbed doses with hierarchical partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Linda [Federal Office for Radiation Protection, Department Radiation Protection and Health, Oberschleissheim (Germany); University of Manchester, The Faculty of Medical and Human Sciences, Manchester (United Kingdom)

    2013-03-15

    It has generally been assumed that the neutron and γ-ray absorbed doses in the data from the life span study (LSS) of the Japanese A-bomb survivors are too highly correlated for an independent separation of the all solid cancer risks due to neutrons and due to γ-rays. However, with the release of the most recent data for all solid cancer incidence and the increased statistical power over previous datasets, it is instructive to consider alternatives to the usual approaches. Simple excess relative risk (ERR) models for radiation-induced solid cancer incidence fitted to the LSS epidemiological data have been applied with neutron and γ-ray absorbed doses as separate explanatory covariables. A simple evaluation of the degree of independent effects from γ-ray and neutron absorbed doses on the all solid cancer risk with the hierarchical partitioning (HP) technique is presented here. The degree of multi-collinearity between the γ-ray and neutron absorbed doses has also been considered. The results show that, whereas the partial correlation between the neutron and γ-ray colon absorbed doses may be considered to be high at 0.74, this value is just below the level beyond which remedial action, such as adding the doses together, is usually recommended. The resulting variance inflation factor is 2.2. Applying HP indicates that just under half of the drop in deviance resulting from adding the γ-ray and neutron absorbed doses to the baseline risk model comes from the joint effects of the neutrons and γ-rays - leaving a substantial proportion of this deviance drop accounted for by individual effects of the neutrons and γ-rays. The average ERR/Gy γ-ray absorbed dose and the ERR/Gy neutron absorbed dose that have been obtained here directly for the first time, agree well with previous indirect estimates. The average relative biological effectiveness (RBE) of neutrons relative to γ-rays, calculated directly from fit parameters to the all solid cancer ERR model with both

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  9. 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)

  10. Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

    Energy Technology Data Exchange (ETDEWEB)

    Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

    2005-11-15

    A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

  11. Implementation of an Analytical Model for Leakage Neutron Equivalent Dose in a Proton Radiotherapy Planning System

    Energy Technology Data Exchange (ETDEWEB)

    Eley, John [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Newhauser, Wayne, E-mail: newhauser@lsu.edu [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Homann, Kenneth; Howell, Rebecca [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Schneider, Christopher [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Durante, Marco; Bert, Christoph [GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany)

    2015-03-11

    Equivalent dose from neutrons produced during proton radiotherapy increases the predicted risk of radiogenic late effects. However, out-of-field neutron dose is not taken into account by commercial proton radiotherapy treatment planning systems. The purpose of this study was to demonstrate the feasibility of implementing an analytical model to calculate leakage neutron equivalent dose in a treatment planning system. Passive scattering proton treatment plans were created for a water phantom and for a patient. For both the phantom and patient, the neutron equivalent doses were small but non-negligible and extended far beyond the therapeutic field. The time required for neutron equivalent dose calculation was 1.6 times longer than that required for proton dose calculation, with a total calculation time of less than 1 h on one processor for both treatment plans. Our results demonstrate that it is feasible to predict neutron equivalent dose distributions using an analytical dose algorithm for individual patients with irregular surfaces and internal tissue heterogeneities. Eventually, personalized estimates of neutron equivalent dose to organs far from the treatment field may guide clinicians to create treatment plans that reduce the risk of late effects.

  12. SU-E-T-566: Neutron Dose Cloud Map for Compact ProteusONE Proton Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Syh, J; Patel, B; Syh, J; Rosen, L; Wu, H [Willis-Knighton Medical Center, Shreveport, LA (United States)

    2015-06-15

    Purpose: To establish the base line of neutron cloud during patient treatment in our new compact Proteus One proton pencil beam scanning (PBS) system with various beam delivery gantry angles, with or without range shifter (RS) at different body sites. Pencil beam scanning is an emerging treatment technique, for the concerns of neutron exposure, this study is to evaluate the neutron dose equivalent per given delivered dose under various treatment conditions at our proton therapy center. Methods: A wide energy neutron dose equivalent detector (SWENDI-II, Thermo Scientific, MA) was used for neutron dose measurements. It was conducted in the proton therapy vault during beam was on. The measurement location was specifically marked in order to obtain the equivalent dose of neutron activities (H). The distances of 100, 150 and 200 cm at various locations are from the patient isocenter. The neutron dose was measured of proton energy layers, # of spots, maximal energy range, modulation width, field radius, gantry angle, snout position and delivered dose in CGE. The neutron dose cloud is reproducible and is useful for the future reference. Results: When distance increased the neutron equivalent dose (H) reading did not decrease rapidly with changes of proton energy range, modulation width or spot layers. For cranial cases, the average mSv/CGE was about 0.02 versus 0.032 for pelvis cases. RS will induce higher H to be 0.10 mSv/CGE in average. Conclusion: From this study, neutron per dose ratio (mSv/CGE) slightly depends upon various treatment parameters for pencil beams. For similar treatment conditions, our measurement demonstrates this value for pencil beam scanning beam has lowest than uniform scanning or passive scattering beam with a factor of 5. This factor will be monitored continuously for other upcoming treatment parameters in our facility.

  13. Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

    Science.gov (United States)

    Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

    2017-04-01

    The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

  14. Neutron Scattering and High Magnetic Fields

    Energy Technology Data Exchange (ETDEWEB)

    Winn, Barry L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Stone, Matthew B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-11-01

    The workshop “Neutron Scattering and High Magnetic Fields” was held September 4-5, 2014 at the Oak Ridge National Laboratory (ORNL). The workshop was held in response to a recent report by the National Research Council of the National Academy of Sciences entitled “High Magnetic Field Science and Its Application in the United States: Current Status and Future Directions.”1 This report highlights the fact that neutron scattering measurements carried out in high magnetic fields provide important opportunities for new science. The workshop explored the range of the scientific discoveries that could be enabled with neutron scattering measurements at high fields (25 Tesla or larger), the various technologies that might be utilized to build specialized instruments and sample environment equipment to enable this research at ORNL, and possible routes to funding and constructing these facilities and portable high field sample environments.

  15. High-speed motion neutron radiography

    International Nuclear Information System (INIS)

    Bossi, R.H.; Barton, J.P.; Robinson, A.H.

    1982-01-01

    A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames. Synchronization has provided high-speed motion neutron radiographs for evaluation of the firing cycles of 7.62-mm munition rounds within a thick steel rifle barrel. The system has also been used to demonstrate its ability to produce neutron radiographic movies of two-phase flow. The equipment includes a TRIGA reactor capable of pulsing to a peak power of 3000 MW, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16-mm high-speed movie camera. The peak neutron flux incident at the object position is about 4 X 10 11 n/cm 2 X s with a pulse, full-width at half-maximum, of 9 ms. Modulation transfer function techniques have been used to assist optimization of the system performance. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on information availability

  16. Calculation of fast neutron dose in plastic-coated optical fibers

    International Nuclear Information System (INIS)

    Siebert, B.R.L.; Henschel, H.

    1998-01-01

    The dose of fast neutrons in optical fibers with hydrogen-containing coating materials is considerably increased by energetic recoil protons. Their contribution to the dose in a SiO 2 fiber core is calculated by the Monte Carlo method for different fiber geometries and a fiber optic cable. With 14 MeV neutrons the dose in a single fiber is increased by about 21%, whereas in fiber bundles the dose increase can reach about 170%. Maximum dose enhancement in fiber bundles (about 610%) occurs at neutron energies around 5.5 MeV. The dose increase caused by 14 MeV neutrons in the fiber of a typical laboratory cable is about 124%

  17. Development of slow neutron dose meter by track counting

    International Nuclear Information System (INIS)

    Nozaki, Tetsuya; Takeuchi, Hiroshi; Hasnel, S.; Honda, Teruyuki; Harasawa, Susumu.

    1993-01-01

    Prototypes of a plate type and two types of semispherical dosemeter were manufactured and their performances were tested. Polyallyl diglycol carbonate (PADC) was employed as neutron detector and covered with natural LiF and 6 Li-enriched LiF ceramics. They were irradiated in the TRIGA II Reactor of the Rikkyo University, and the sensitivity characteristics for incident angles and neutron energies were analyzed. It is concluded that it may be possible to manufacture small sized neutron dosemeter of flat response to wide energy range from thermal neutrons to epithermal neutrons with sufficient sensitivity for personal monitoring, using LiF ceramic as neutron filters and neutron converters. However the following issues are to be solved: the optimization of thickness of the LiF filter, the effects of albedo of the human body, and the applicability to the intermediate neutrons. (A.Y.)

  18. Reduction of the secondary neutron dose in passively scattered proton radiotherapy, using an optimized pre-collimator/collimator

    International Nuclear Information System (INIS)

    Brenner, David J; Elliston, Carl D; Hall, Eric J; Paganetti, Harald

    2009-01-01

    Proton radiotherapy represents a potential major advance in cancer therapy. Most current proton beams are spread out to cover the tumor using passive scattering and collimation, resulting in an extra whole-body high-energy neutron dose, primarily from proton interactions with the final collimator. There is considerable uncertainty as to the carcinogenic potential of low doses of high-energy neutrons, and thus we investigate whether this neutron dose can be significantly reduced without major modifications to passively scattered proton beam lines. Our goal is to optimize the design features of a patient-specific collimator or pre-collimator/collimator assembly. There are a number of often contradictory design features, in terms of geometry and material, involved in an optimal design. For example, plastic or hybrid plastic/metal collimators have a number of advantages. We quantify these design issues, and investigate the practical balances that can be achieved to significantly reduce the neutron dose without major alterations to the beamline design or function. Given that the majority of proton therapy treatments, at least for the next few years, will use passive scattering techniques, reducing the associated neutron-related risks by simple modifications of the collimator assembly design is a desirable goal.

  19. Distributions of neutron and gamma doses in phantom under a mixed field

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.

    1982-06-01

    A calculation program, based on Monte Carlo method, allowed to estimate the absorbed doses relatives to the reactor primary radiation, in a water cubic phantom and in cylindrical phantoms modelized from tissue compositions. This calculation is a theoretical approach of gamma and neutron dose gradient study in an animal phantom. PIN junction dosimetric characteristics have been studied experimentally. Air and water phantom radiation doses measured by PIN junction and lithium 7 fluoride, in reactor field have been compared to doses given by dosimetry classical techniques as tissue equivalent plastic and aluminium ionization chambers. Dosimeter responses have been employed to evaluate neutron and gamma doses in plastinaut (tissue equivalent plastic) and animal (piglet). Dose repartition in the piglet bone medulla has been also determined. This work has been completed by comparisons with Doerschell, Dousset and Brown results and by neutron dose calculations; the dose distribution related to lineic energy transfer in Auxier phantom has been also calculated [fr

  20. Peripheral photon and neutron doses from prostate cancer external beam irradiation.

    Science.gov (United States)

    Bezak, Eva; Takam, Rundgham; Marcu, Loredana G

    2015-12-01

    Peripheral photon and neutron doses from external beam radiotherapy (EBRT) are associated with increased risk of carcinogenesis in the out-of-field organs; thus, dose estimations of secondary radiation are imperative. Peripheral photon and neutron doses from EBRT of prostate carcinoma were measured in Rando phantom. (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P glass-rod thermoluminescence dosemeters (TLDs) were inserted in slices of a Rando phantom followed by exposure to 80 Gy with 18-MV photon four-field 3D-CRT technique. The TLDs were calibrated using 6- and 18-MV X-ray beam. Neutron dose equivalents measured with CR-39 etch-track detectors were used to derive readout-to-neutron dose conversion factor for (6)LiF:Mg,Cu,P TLDs. Average neutron dose equivalents per 1 Gy of isocentre dose were 3.8±0.9 mSv Gy(-1) for thyroid and 7.0±5.4 mSv Gy(-1) for colon. For photons, the average dose equivalents per 1 Gy of isocentre dose were 0.2±0.1 mSv Gy(-1) for thyroid and 8.1±9.7 mSv Gy(-1) for colon. Paired (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P TLDs can be used to measure photon and neutron doses simultaneously. Organs in close proximity to target received larger doses from photons than those from neutrons whereas distally located organs received higher neutron versus photon dose. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  1. Very High Energy Neutron Scattering from Hydrogen

    International Nuclear Information System (INIS)

    Cowley, R A; Stock, C; Bennington, S M; Taylor, J; Gidopoulos, N I

    2010-01-01

    The neutron scattering from hydrogen in polythene has been measured with the direct time-of flight spectrometer, MARI, at the ISIS facility of the Rutherford Appleton Laboratory with incident neutron energies between 0.5 eV and 600 eV. The results of experiments using the spectrometer, VESUVIO, have given intensities from hydrogen containing materials that were about 60% of the intensity expected from hydrogen. Since VESUVIO is the only instrument in the world that routinely operates with incident neutron energies in the eV range we have chosen to measure the scattering from hydrogen at high incident neutron energies with a different type of instrument. The MARI, direct time-of-flight, instrument was chosen for the experiment and we have studied the scattering for several different incident neutron energies. We have learnt how to subtract the gamma ray background, how to calibrate the incident energy and how to convert the spectra to an energy plot . The intensity of the hydrogen scattering was independent of the scattering angle for scattering angles from about 5 degrees up to 70 degrees for at least 3 different incident neutron energies between 20 eV and 100 eV. When the data was put on an absolute scale, by measuring the scattering from 5 metal foils with known thicknesses under the same conditions we found that the absolute intensity of the scattering from the hydrogen was in agreement with that expected to an accuracy of ± 5.0% over a wide range of wave-vector transfers between 1 and 250 A -1 . These measurements show that it is possible to measure the neutron scattering with incident neutron energies up to at least 100 eV with a direct geometry time-of-flight spectrometer and that the results are in agreement with conventional scattering theory.

  2. NEUTRON GENERATOR FACILITY AT SFU: GEANT4 DOSE RATE PREDICTION AND VERIFICATION.

    Science.gov (United States)

    Williams, J; Chester, A; Domingo, T; Rizwan, U; Starosta, K; Voss, P

    2016-11-01

    Detailed dose rate maps for a neutron generator facility at Simon Fraser University were produced via the GEANT4 Monte Carlo framework. Predicted neutron dose rates throughout the facility were compared with radiation survey measurements made during the facility commissioning process. When accounting for thermal neutrons, the prediction and measurement agree within a factor of 2 or better in most survey locations, and within 10 % inside the vault housing the neutron generator. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  3. High precision neutron polarization for PERC

    International Nuclear Information System (INIS)

    Klauser, C.

    2013-01-01

    The decay of the free neutron into a proton, an electron and an anti-electron neutrino offers a simple system to study the semi-leptonic weak decay. High precision measurements of angular correlation coefficients of this decay provide the opportunity to test the standard model on the low energy frontier. The Proton Electron Radiation Channel PERC is part of a new generation of expriments pushing the accuracy of such an angular correlation coefficient measurement towards 10 -4 . Past experiments have been limited to an accuracy of 10 -3 with uncertainties on the neutron polarization as one of the leading systematic errors. This thesis focuses on the development of a stable, highly precise neutron polarization for a large, divergent cold neutron beam. A diagnostic tool that provides polarization higher than 99.99 % and analyzes with an accuracy of 10 -4 , the Opaque Test Bench, is presented and validated. It consists of two highly opaque polarized helium cells. The Opaque Test Bench reveals depolarizing effects in polarizing supermirrors commonly used for polarization in neutron decay experiments. These effects are investigated in detail. They are due to imperfect lateral magnetization in supermirror layers and can be minimized by significantly increased magnetizing fields and low incidence angle and supermirror factor m. A subsequent test in the crossed (X-SM) geometry demonstrated polarizations up to 99.97% from supermirrors only, improving neutron polarization with supermirrors by an order of magnitude. The thesis also discusses other neutron optical components of the PERC beamline: Monte-Carlo simulations of the beamline under consideration of the primary guide are carried out. In addition, calculation shows that PERC would statistically profit from an installation at the European Spallation source. Furthermore, beamline components were tested. A radio-frequency spin flipper was confirmed to work with an efficiency higher than 0.9999. (author) [de

  4. Accuracy of neutron dose evaluation in the area monitoring for LHD experiments

    CERN Document Server

    Yamanishi, H; Uda, T; Tanahashi, S; Saitou, M; Handa, H

    2000-01-01

    The error in the evaluation of neutron dose during calculation of the neutron field around the large helical device (LHD) in D-D operation is discussed. The expected neutron dose at each monitoring point was derived from the dose conversion factor and neutron fluence data, which was calculated with the radiation transport code DOT-3.5. In contrast, the detected dose at the neutron counter was obtained from the fluence data and the detector response given by calculation with MCNP-4b. The neutron counter used in these calculations consisted of a helium-3 proportional counter with a cylindrical polyethylene moderator. According to the results of the calculations, the ratio of the detected dose to the expected dose was found to lie in the range 1.0-3.0 on the outdoor monitoring points. Since the response of a single neutron counter may lead to inconsistencies in the dose conversion factor, we attempted to minimize these inconsistencies by using a pair of counters with moderators of different thickness. The ratio ...

  5. Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, Jerome M.

    1999-12-14

    A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only.

  6. Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

    International Nuclear Information System (INIS)

    Verbeke, Jerome M.

    1999-01-01

    A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only

  7. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  8. The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

    International Nuclear Information System (INIS)

    Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

    2014-01-01

    Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

  9. Tensile property changes of metals and irradiated to low doses with fission, fusion and spallation neutrons

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

    1992-01-01

    The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructures and mechanical properties of metals. Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36-55 C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90 C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa

  10. Development of neutron dosimeter using CR-39 for measurement of ambient dose equivalent

    International Nuclear Information System (INIS)

    Maki, Daisuke; Shinozaki, Wakako; Ohguchi, Hiroyuki; Yamamoto, Takayoshi; Nakamura, Takayoshi

    2010-01-01

    A CR-39 has good advantages such as cumulative type dosimeter, small fading effect and gamma-ray insensitive. Therefore, we developed the wide energy-range environmental neutron dosimeter using eight CR-39s for area monitoring in this study. This dosimeter is made of octagonal columnar polyethylene block which height is 60 mm and bottom side is 25 mm. The dosimeter contains two types of CR-39s for fast neutron detection and slow neutron detection. Four CR-39s for fast neutron detection are used for detection of recoil protons produced by H (n, p) reactions. Four CR-39s for slow neutron detection are used with boron nitride converter to detect alpha-rays produced by 10 B (n, α) 7 Li reactions. Ambient dose equivalent is obtained by adding the number of etch-pits observed in four CR-39s for fast neutron detection to the number of etch-pits observed in four CR-39s for slow neutron detection with appropriate constants respectively. Dosimeters were irradiated with some energetic neutrons and evaluated results of ambient dose equivalent were compared with results from neutron transport calculations. Energy response of dosimeter shows good agreement with neutron fluence to ambient dose equivalent conversion coefficients. Directional dependence of dosimeter is at the same level as the rem-counter. (author)

  11. Phantom experiment of depth-dose distributions for gadolinium neutron capture therapy

    International Nuclear Information System (INIS)

    Matsumoto, T.; Kato, K.; Sakuma, Y.; Tsuruno, A.; Matsubayashi, M.

    1993-01-01

    Depth-dose distributions in a tumor simulated phantom were measured for thermal neutron flux, capture gamma-ray and internal conversion electron dose rates for gadolinium neutron capture therapy. The results show that (i) a significant dose enhancement can be achieved in the tumor by capture gamma-rays and internal conversion electrons but the dose is mainly due to capture gamma-rays from the Gd(n, γ) reactions, therefore, is not selective at the cellular level, (ii) the dose distribution was a function of strongly interrelated parameters such as gadolinium concentrations, tumor site and neutron beam size (collimator aperture size), and (iii) the Gd-NCT by thermal neutrons appears to be a potential for treatment of superficial tumor. (author)

  12. Improving the neutron-to-photon discrimination capability of detectors used for neutron dosimetry in high energy photon beam radiotherapy

    International Nuclear Information System (INIS)

    Irazola, L.; Terrón, J.A.; Bedogni, R; Pola, A.; Lorenzoli, M.; Sánchez-Nieto, B.; Gómez, F.; Sánchez-Doblado, F.

    2016-01-01

    The increasing interest of the medical community to radioinduced second malignancies due to photoneutrons in patients undergoing high-energy radiotherapy, has stimulated in recent years the study of peripheral doses, including the development of some dedicated active detectors. Although these devices are designed to respond to neutrons only, their parasitic photon response is usually not identically zero and anisotropic. The impact of these facts on measurement accuracy can be important, especially in points close to the photon field-edge. A simple method to estimate the photon contribution to detector readings is to cover it with a thermal neutron absorber with reduced secondary photon emission, such as a borated rubber. This technique was applied to the TNRD (Thermal Neutron Rate Detector), recently validated for thermal neutron measurements in high-energy photon radiotherapy. The positive results, together with the accessibility of the method, encourage its application to other detectors and different clinical scenarios. - Highlights: • Neutron-to-photon discrimination of a thermal neutron detector used in radiotherapy. • Photon and anisotropic response study with distance and beam incidence of thermal neutron detector. • Borated rubber for estimating photon contribution in any thermal neutron detector.

  13. Measurement of neutron dose equivalent outside and inside of the treatment vault of GRID therapy

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xudong; Charlton, Michael A.; Esquivel, Carlos; Eng, Tony Y.; Li, Ying; Papanikolaou, Nikos [University of Texas Health Science Center, San Antonio, Texas 78229 (United States)

    2013-09-15

    Purpose: To evaluate the neutron and photon dose equivalent rates at the treatment vault entrance (H{sub n,D} and H{sub G}), and to study the secondary radiation to the patient in GRID therapy. The radiation activation on the grid was studied.Methods: A Varian Clinac 23EX accelerator was working at 18 MV mode with a grid manufactured by .decimal, Inc. The H{sub n,D} and H{sub G} were measured using an Andersson–Braun neutron REM meter, and a Geiger Müller counter. The radiation activation on the grid was measured after the irradiation with an ion chamber γ-ray survey meter. The secondary radiation dose equivalent to patient was evaluated by etched track detectors and OSL detectors on a RANDO{sup ®} phantom.Results: Within the measurement uncertainty, there is no significant difference between the H{sub n,D} and H{sub G} with and without a grid. However, the neutron dose equivalent to the patient with the grid is, on average, 35.3% lower than that without the grid when using the same field size and the same amount of monitor unit. The photon dose equivalent to the patient with the grid is, on average, 44.9% lower. The measured average half-life of the radiation activation in the grid is 12.0 (±0.9) min. The activation can be categorized into a fast decay component and a slow decay component with half-lives of 3.4 (±1.6) min and 15.3 (±4.0) min, respectively. There was no detectable radioactive contamination found on the surface of the grid through a wipe test.Conclusions: This work indicates that there is no significant change of the H{sub n,D} and H{sub G} in GRID therapy, compared with a conventional external beam therapy. However, the neutron and scattered photon dose equivalent to the patient decrease dramatically with the grid and can be clinical irrelevant. Meanwhile, the users of a grid should be aware of the possible high dose to the radiation worker from the radiation activation on the surface of the grid. A delay in handling the grid after the beam

  14. Optimized dose distribution of a high dose rate vaginal cylinder

    International Nuclear Information System (INIS)

    Li Zuofeng; Liu, Chihray; Palta, Jatinder R.

    1998-01-01

    Purpose: To present a comparison of optimized dose distributions for a set of high-dose-rate (HDR) vaginal cylinders calculated by a commercial treatment-planning system with benchmark calculations using Monte-Carlo-calculated dosimetry data. Methods and Materials: Optimized dose distributions using both an isotropic and an anisotropic dose calculation model were obtained for a set of HDR vaginal cylinders. Mathematical optimization techniques available in the computer treatment-planning system were used to calculate dwell times and positions. These dose distributions were compared with benchmark calculations with TG43 formalism and using Monte-Carlo-calculated data. The same dwell times and positions were used for a quantitative comparison of dose calculated with three dose models. Results: The isotropic dose calculation model can result in discrepancies as high as 50%. The anisotropic dose calculation model compared better with benchmark calculations. The differences were more significant at the apex of the vaginal cylinder, which is typically used as the prescription point. Conclusion: Dose calculation models available in a computer treatment-planning system must be evaluated carefully to ensure their correct application. It should also be noted that when optimized dose distribution at a distance from the cylinder surface is calculated using an accurate dose calculation model, the vaginal mucosa dose becomes significantly higher, and therefore should be carefully monitored

  15. Investigation of dose distribution in mixed neutron-gamma field of boron neutron capture therapy using N isopropylacrylamide gel

    Energy Technology Data Exchange (ETDEWEB)

    Bavarmegin, Elham; Sadremomtaz, Alireza [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Khalafi, Hossein; Kasesaz, Yaser [Dept. of Physics, University of Guilan, Rasht (Iran, Islamic Republic of); Khajeali, Azim [Medical Education Research Center, Tabriz (Iran, Islamic Republic of)

    2017-02-15

    Gel dosimeters have unique advantages in comparison with other dosimeters. Until now, these gels have been used in different radiotherapy techniques as a reliable dosimetric tool. Because dose distribution measurement is an important factor for appropriate treatment planning in different radiotherapy techniques, in this study, we evaluated the ability of the N-isopropylacrylamide (NIPAM) polymer gel to record the dose distribution resulting from the mixed neutron-gamma field of boron neutron capture therapy (BNCT). In this regard, a head phantom containing NIPAM gel was irradiated using the Tehran Research Reactor BNCT beam line, and then by a magnetic resonance scanner. Eventually, the R2 maps were obtained in different slices of the phantom by analyzing T2-weighted images. The results show that NIPAM gel has a suitable potential for recording three-dimensional dose distribution in mixed neutron-gamma field dosimetry.

  16. Dose planning with comparison to in vivo dosimetry for epithermal neutron irradiation of the dog brain

    International Nuclear Information System (INIS)

    Seppaelae, Tiina; Auterinen, Iiro; Aschan, Carita; Seren, Tom; Benczik, Judit; Snellman, Marjatta; Huiskamp, Rene; Ramadan, Usama Abo; Kankaanranta, Leena; Joensuu, Heikki; Savolainen, Sauli

    2002-01-01

    Boron neutron capture therapy (BNCT) is an experimental type of radiotherapy, presently being used to treat glioblastoma and melanoma. To improve patient safety and to determine the radiobiological characteristics of the epithermal neutron beam of Finnish BNCT facility (FiR 1) dose-response studies were carried on the brain of dogs before starting the clinical trials. A dose planning procedure was developed and uncertainties of the epithermal neutron-induced doses were estimated. The accuracy of the method of computing physical doses was assessed by comparing with in vivo dosimetry. Individual radiation dose plans were computed using magnetic resonance images of the heads of 15 Beagle dogs and the computational model of the FiR 1 epithermal neutron beam. For in vivo dosimetry, the thermal neutron fluences were measured using Mn activation foils and the gamma-ray doses with MCP-7s type thermoluminescent detectors placed both on the skin surface of the head and in the oral cavity. The degree of uncertainty of the reference doses at the thermal neutron maximum was estimated using a dose-planning program. The estimated uncertainty (±1 standard deviation) in the total physical reference dose was ±8.9%. The calculated and the measured dose values agreed within the uncertainties at the point of beam entry. The conclusion is that the dose delivery to the tissue can be verified in a practical and reliable fashion by placing an activation dosimeter and a TL detector at the beam entry point on the skin surface with homogeneous tissues below. However, the point doses cannot be calculated correctly in the inhomogeneous area near air cavities of the head model with this type of dose-planning program. This calls for attention in dose planning in human clinical trials in the corresponding areas

  17. Dose distribution and clinical response of glioblastoma treated with boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Matsuda, M. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan)], E-mail: mhide-m@gk9.so-net.ne.jp; Yamamoto, T. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan); Kumada, H. [Japan Atomic Energy Agency, Shirakatashirane 2-4, Tokai (Japan); Nakai, K.; Shirakawa, M.; Tsurubuchi, T.; Matsumura, A. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan)

    2009-07-15

    The dose distribution and failure pattern after treatment with the external beam boron neutron capture therapy (BNCT) protocol were retrospectively analyzed. BSH (5 g/body) and BPA (250 mg/kg) based BNCT was performed in eight patients with newly diagnosed glioblastoma. The gross tumor volume (GTV) and clinical target volume (CTV)-1 were defined as the residual gadolinium-enhancing volume. CTV-2 and CTV-3 were defined as GTV plus a margin of 2 and 3 cm, respectively. As additional photon irradiation, a total X-ray dose of 30 Gy was given to the T2 high intensity area on MRI. Five of the eight patients were alive at analysis for a mean follow-up time of 20.3 months. The post-operative median survival time of the eight patients was 27.9 months (95% CI=21.0-34.8). The minimum tumor dose of GTV, CTV-2, and CTV-3 averaged 29.8{+-}9.9, 15.1{+-}5.4, and 12.4{+-}2.9 Gy, respectively. The minimum tumor non-boron dose of GTV, CTV-2, and CTV-3 averaged 2.0{+-}0.5, 1.3{+-}0.3, and 1.1{+-}0.2 Gy, respectively. The maximum normal brain dose, skin dose, and average brain dose were 11.4{+-}1.5, 9.6{+-}1.4, and 3.1{+-}0.4 Gy, respectively. The mean minimum dose at the failure site in cases of in-field recurrence (IR) and out-field recurrence (OR) was 26.3{+-}16.7 and 14.9 GyEq, respectively. The calculated doses at the failure site were at least equal to the tumor control doses which were previously reported. We speculate that the failure pattern was related to an inadequate distribution of boron-10. Further improvement of the microdistribution of boron compounds is expected, and may improve the tumor control by BNCT.

  18. Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma-ray mixed fields

    International Nuclear Information System (INIS)

    Ogawa, Yoshihiro

    1985-01-01

    In order to consider the problems concerned with personnel dosimetry using film badges and TLDs, an intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in five different neutron and gamma-ray mixed fields at KUR and UTR-KINKI from the practical point of view. For the estimation of thermal neutron dose equivalent, it may be concluded that each personnel dosimeter has good performances in the precision, that is, the standard deviations in the measured values by individual dosimeter were within 24 %, and the dose equivalent responses to thermal neutron were almost independent on cadmium ratio and gamma-ray contamination. However, the relative thermal neutron dose equivalent of individual dosimeter normalized to the ICRP recommended value varied considerably and a difference of about 4 times was observed among the dosimeters. From the results obtained, it is suggested that the standardization of calibration factors and procedures is required from the practical point of radiation protection and safety. (author)

  19. High energy radiation from neutron stars

    International Nuclear Information System (INIS)

    Ruderman, M.

    1985-04-01

    Topics covered include young rapidly spinning pulsars; static gaps in outer magnetospheres; dynamic gaps in pulsar outer magnetospheres; pulse structure of energetic radiation sustained by outer gap pair production; outer gap radiation, Crab pulsar; outer gap radiation, the Vela pulsar; radioemission; and high energy radiation during the accretion spin-up of older neutron stars. 26 refs., 10 figs

  20. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)

    2017-07-11

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  1. Portable instrument for measuring neutron energy spectra and neutron dose in a mixed n-γ field

    International Nuclear Information System (INIS)

    Daniels, C. J.; Silberberg, J. L.

    1980-01-01

    A portable high-speed neutron spectrometer consists of an organic scintillator, a true zero-crossing pulse shape discriminator, a 1 MHZ conversion-rate multichannel analyzer, an 8-bit microcomputer, and appropriate displays. The device can be used to measure neutron energy spectra and kerma rate in intense n- gamma radiation fields in which the neutron energy is from 5 to 15 MEV

  2. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

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

    1988-11-01

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  3. Radiation dose distribution monitoring at neutron radiography facility area, Nuclear Energy Unit, Malaysia

    International Nuclear Information System (INIS)

    Abdul Razak Daud

    1995-01-01

    One experiment was carried out to get the distribution of radiation doses at the neutron radiography facilities, Nuclear Energy Unit, Malaysia. The analysis was done to evaluate the safety level of the area. The analysis was used in neutron radiography work

  4. Monte Carlo calculation of ''skyshine'' neutron dose from ALS [Advanced Light Source

    International Nuclear Information System (INIS)

    Moin-Vasiri, M.

    1990-06-01

    This report discusses the following topics on ''skyshine'' neutron dose from ALS: Sources of radiation; ALS modeling for skyshine calculations; MORSE Monte-Carlo; Implementation of MORSE; Results of skyshine calculations from storage ring; and Comparison of MORSE shielding calculations

  5. Occupational doses due to photoneutrons in medical linear accelerators rooms; Doses ocupacionais devido a neutrons em salas de aceleradores lineares de uso medico

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Alessandro Facure Neves de Salles

    2006-04-15

    Medical linear accelerators, with maximum photon energies above 10 MeV, are becoming of common use in Brazil. Although desirable in the therapeutic point of view, the increase in photon energies causes the generation of undesired neutrons, which are produced through nuclear reactions between photons and the high Z target nuclei of the materials that constitute the accelerator head. In this work, MCNP simulation was undertaken to examine the neutron equivalent doses around the accelerators head and at the entrance of medical linear accelerators treatment rooms, some of them licensed in Brazil by the National Regulatory Agency (CNEN). The simulated neutron dose equivalents varied between 2 e 26 {mu} Sv/Gy{sub RX}, and the results were compared with calculations performed with the use of some semi-empirical equations found in literature. It was found that the semi-empirical equations underestimate the simulated neutron doses in the majority of the cases, if compared to the simulated values, suggesting that these equations must be revised, due to the increasing number of high energy machines in the country. (author)

  6. Dose field research of analysis room for in-hospital neutron irradiator

    International Nuclear Information System (INIS)

    Zhang Zizhu; Song Mingzhe; Li Wei; Chen Jun; Yang Yong; Li Yiguo

    2012-01-01

    Neutron equivalent dose rate and y ray dose rate inside the analysis room of the in-hospital neutron irradiator (IHNI) and outdoor were measured. The results show that γ ray dose rate inside the analysis room exceeds calculation value many times and γ/ ray dose rate outdoor is higher than supervision region dose limit of 7.5 μSv/h. According to the measurement results and the Monte Carlo simulation, the following shielding plan was adopted. Lead shielding with thickness of 16 cm was installed on the wall, which faces the neutron beam, to shield γ ray, and lithium polyethylene plate with thickness of l cm was installed on all the wall (not including ceiling and floor) to shield scattering neutron. After shielding transformation, the highest γ ray dose rate point inside the analysis room decreased 277 times, the neutron equivalent dose rate decreased 5.8 times, and the outdoor γ/ray dose rate decreased nearly 90 times. (authors)

  7. Dose Measurements of Bremsstrahlung-Produced Neutrons at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Job, P.K.; Pisharody, M.; Semones, E.

    1998-01-01

    Bremsstrahlung is generated in the storage rings of the synchrotron radiation facilities by the radiative interaction of the circulating particle beam with both the residual gas molecules and storage ring components. These bremsstrahlung photons, having an energy range of zero to the maximum energy of the particle beam, interact with beamline components like beam stops and collimators generating photoneutrons of varying energies. There are three main processes by which photoneutrons may be produced by the high energy bremsstrahlung photons: giant nuclear dipole resonance and decay (10 MeV γ γ γ > 140 MeV). The giant resonance neutrons are emitted almost isotropically and have an average energy of about 2 MeV. High energy neutrons (E > 10 MeV) emitted from the quasi-deuteron decay and intranuclear cascade are peaked in the forward direction. At the Advanced Photon Source (APS), where bremsstrahlung energy can be as high as 7 GeV, production of photoneutrons in varying yields is possible from all of the above three processes. The bremsstrahlung produced along a typical 15.38-m straight path of the insertion device (ID) beamline of the APS has been measured and analyzed in previous studies. High-Z materials constituting the beamline components, such as collimators and beam stops, can produce photoneutrons upon interaction with these bremsstrahlung photons. The 1/E nature of the bremsstrahlung spectrum and the fact that the photoneutron production cross section is comparatively larger in the energy region 10 MeV γ 3 detector, as well as a very sensitive pressurized 3 He detector, is used for neutron dose measurements. The dose equivalent rates, normalized to bremsstrahlung power, beam current, and storage ring vacuum, are measured for various targets. This report details the experimental setup,

  8. A high-sensitivity neutron counter and waste-drum counting with the high-sensitivity neutron instrument

    International Nuclear Information System (INIS)

    Hankins, D.E.; Thorngate, J.H.

    1993-04-01

    At Lawrence Livermore National Laboratory (LLNL), a highly sensitive neutron counter was developed that can detect and accurately measure the neutrons from small quantities of plutonium or from other low-level neutron sources. This neutron counter was originally designed to survey waste containers leaving the Plutonium Facility. However, it has proven to be useful in other research applications requiring a high-sensitivity neutron instrument

  9. 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)

  10. Soil biological shield exposed to high energy neutrons; Zemlja kao bioloski stit od neutrona visokih energija

    Energy Technology Data Exchange (ETDEWEB)

    Simovic, R; Marinkovic, N [Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1993-04-15

    Shielding efficiency of soil biological shield exposed to high energy neutrons was investigated. Dose rate equivalents for neutrons, secondary gamma and gamma radiation were computed on the surface of soil slabs having different thicknesses. Yields of primary and secondary nuclear radiation in the total dose were evaluated. Influence of the incident neutron spectrum, water content and chemical composition of the material on its shielding efficiency was examined. It was found that the soil density and the water content determine the quality of biological shield, the influence of other factors being less important. Comparison of shielding efficiencies for soil with sand, brick and ordinary concrete shields was done.

  11. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

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

    1993-01-01

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  12. A high-resolution neutron spectra unfolding method using the Genetic Algorithm technique

    CERN Document Server

    Mukherjee, B

    2002-01-01

    The Bonner sphere spectrometers (BSS) are commonly used to determine the neutron spectra within various nuclear facilities. Sophisticated mathematical tools are used to unfold the neutron energy distribution from the output data of the BSS. This paper highlights a novel high-resolution neutron spectra-unfolding method using the Genetic Algorithm (GA) technique. The GA imitates the biological evolution process prevailing in the nature to solve complex optimisation problems. The GA method was utilised to evaluate the neutron energy distribution, average energy, fluence and equivalent dose rates at important work places of a DIDO class research reactor and a high-energy superconducting heavy ion cyclotron. The spectrometer was calibrated with a sup 2 sup 4 sup 1 Am/Be (alpha,n) neutron standard source. The results of the GA method agreed satisfactorily with the results obtained by using the well-known BUNKI neutron spectra unfolding code.

  13. High pressure neutron powder diffraction at LANSCE

    International Nuclear Information System (INIS)

    Von Dreele, R.B.

    1994-01-01

    By making use of the recently developed ''Paris-Edinburgh'' high pressure cell, the author has successfully performed neutron powder experiments to 10GPa at ambient temperature. Results for the structural compression of the high Tc 1223-Hg superconductor to 9.2 GPa, the compression and possible hydrogen bond formation in brucite, Mg(OD) 2 , to 9.3 GPa, and the molecular reorientation in nitromethane to 5.5 GPa will be presented

  14. 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)

  15. Measurement of neutron and gamma absorbed doses in phantoms exposed to mixed fields

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.; Lemaire, G.; Maas, J.

    1985-01-01

    In order to study the dosimetric characteristics of PIN junctions, the absorbed doses measured by junctions and FLi7 in air and water phantoms were compared with the doses measured by classical neutron dosimetry in mixed fields. The validity of the experimental responses of PIN junctions being thus checked and established, neutron and gamma dose distributions in tissue equivalent plastic phantoms (plastinaut) and mammals (piglets) were evaluated as well as the absorbed dose distributions in the pig bone-marrow producing areas. By using correlatively a Monte-Carlo calculation method and applying some simplifying assumptions, the absorbed doses were derived from the spectrum of SILENE's neutrons at various depths inside a cubic water phantom and the results were compared with some from the literature [fr

  16. Neutron dose rate at the SwissFEL injector test facility: first measurements

    International Nuclear Information System (INIS)

    Hohmann, E.; Frey, N.; Fuchs, A.; Harm, C.; Hoedlmoser, H.; Luescher, R.; Mayer, S.; Morath, O.; Philipp, R.; Rehmann, A.; Schietinger, T.

    2014-01-01

    At the Paul Scherrer Institute, the new SwissFEL Free Electron Laser facility is currently in the design phase. It is foreseen to accelerate electrons up to a maximum energy of 7 GeV with a pulsed time structure. An injector test facility is operated at a maximum energy of 300 MeV and serves as the principal test and demonstration plant for the SwissFEL project. Secondary radiation is created in unavoidable interactions of the primary beam with beamline components. The resulting ambient dose-equivalent rate due to neutrons was measured along the beamline with different commercially available survey instruments. The present study compares the readings of these neutron detectors (one of them is specifically designed for measurements in pulsed fields). The experiments were carried out in both, a normal and a diagnostic mode of operation of the injector. Measurements were taken at the SwissFEL injector test facility using three different types of commercially available survey instruments for normal and diagnostic mode of operation at different positions inside the accelerator vault. During normal operation, the doses indicated by the different instruments agree within the measurement uncertainty except for the beam dump region. There, due to its limited energy range and high sensitivity, the LB6411 shows significantly lower dose values than the other instruments. The photon background in the vault associated with each pulse causes the scintillator used by the LB6419 to saturate. As a result, only the channel using the delayed 12 C(n,p)12-reaction could be used during the measurements. The highest doses per pulse were measured next to the beam dump and the bunch compressor. For the optimisation of the accelerator, luminescent screens can be inserted into the beam path causing a dose distributed over several metres depending on the screen type. The dose arise to 40 % from neutrons with energies of >20 MeV. Although the charge of each pulse were reduced to decrease

  17. Neutron activation analysis of high purity substances

    International Nuclear Information System (INIS)

    Gil'bert, Eh.N.

    1987-01-01

    Peculiarities of neutron-activation analysis (NAA) of high purity substances are considered. Simultaneous determination of a wide series of elements, high sensitivity (the lower bound of determined contents 10 -9 -10 -10 %), high selectivity and accuracy (Sr=0.10-0.15, and may be decreased up to 0.001), possibility of analysis of the samples from several micrograms to hundreds of grams, simplicity of calibration may be thought NAA advantages. Questions of accounting of NAA systematic errors associated with the neutron flux screening by the analysed matrix and with production of radionuclides of determined elements from accompanying elements according to concurrent nuclear reactions, as well as accounting of errors due to self-absorption of recorded radiation by compact samples, are considered

  18. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

    1993-01-01

    Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

  19. Response functions for computing absorbed dose to skeletal tissues from neutron irradiation

    Science.gov (United States)

    Bahadori, Amir A.; Johnson, Perry; Jokisch, Derek W.; Eckerman, Keith F.; Bolch, Wesley E.

    2011-11-01

    Spongiosa in the adult human skeleton consists of three tissues—active marrow (AM), inactive marrow (IM) and trabecularized mineral bone (TB). AM is considered to be the target tissue for assessment of both long-term leukemia risk and acute marrow toxicity following radiation exposure. The total shallow marrow (TM50), defined as all tissues lying within the first 50 µm of the bone surfaces, is considered to be the radiation target tissue of relevance for radiogenic bone cancer induction. For irradiation by sources external to the body, kerma to homogeneous spongiosa has been used as a surrogate for absorbed dose to both of these tissues, as direct dose calculations are not possible using computational phantoms with homogenized spongiosa. Recent micro-CT imaging of a 40 year old male cadaver has allowed for the accurate modeling of the fine microscopic structure of spongiosa in many regions of the adult skeleton (Hough et al 2011 Phys. Med. Biol. 56 2309-46). This microstructure, along with associated masses and tissue compositions, was used to compute specific absorbed fraction (SAF) values for protons originating in axial and appendicular bone sites (Jokisch et al 2011 Phys. Med. Biol. 56 6857-72). These proton SAFs, bone masses, tissue compositions and proton production cross sections, were subsequently used to construct neutron dose-response functions (DRFs) for both AM and TM50 targets in each bone of the reference adult male. Kerma conditions were assumed for other resultant charged particles. For comparison, AM, TM50 and spongiosa kerma coefficients were also calculated. At low incident neutron energies, AM kerma coefficients for neutrons correlate well with values of the AM DRF, while total marrow (TM) kerma coefficients correlate well with values of the TM50 DRF. At high incident neutron energies, all kerma coefficients and DRFs tend to converge as charged-particle equilibrium is established across the bone site. In the range of 10 eV to 100 Me

  20. Neutron dose rate for {sup 252} Cf AT source in medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Paredes, L.; Balcazar, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [UAM-I, 09340 Mexico D.F. (Mexico); Francois, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)

    2006-07-01

    The AAPM TG-43 modified protocol was used for the calculation of the neutron dose rate of {sup 252}Cf sources for two tissue substitute materials, five normal tissues and six tumours. The {sup 252}Cf AT source model was simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) neutron air kerma strength conversion factor, b) dose rate constant, c) radial dose function, d) geometry factor, e) anisotropy function and f) neutron dose rate. The calculated dose rate in water at 1 cm and 90 degrees from the source long axis, using the Watt fission spectrum, was D{sub n}(r{sub 0}, {theta}{sub 0})= 1.9160 cGy/h-{mu}g. When this value is compared with Rivard et al. calculation using MCNP4B code, 1.8730 cGy/h-{mu}g, a difference of 2.30% is obtained. The results for the reference neutron dose rate in other media show how small variations in the elemental composition between the tissues and malignant tumours, produce variations in the neutron dose rate up to 12.25%. (Author)

  1. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Directory of Open Access Journals (Sweden)

    Ersez Tunay

    2017-01-01

    Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  2. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Science.gov (United States)

    Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

    2017-09-01

    The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  3. 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.)

  4. Application of Whole Body Counter to Neutron Dose Assessment in Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, O.; Tsujimura, N.; Takasaki, K.; Momose, T.; Maruo, Y. [Japan Nuclear Cycle Development Institute, Tokai (Japan)

    2001-09-15

    Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of {sup 24}Na in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of {sup 24}Na is approximately 50Bq for 10minute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 [(Bq{sup 24}Na/g{sup 23}Na)/mGy].

  5. The alanine detector in BNCT dosimetry: dose response in thermal and epithermal neutron fields.

    Science.gov (United States)

    Schmitz, T; Bassler, N; Blaickner, M; Ziegner, M; Hsiao, M C; Liu, Y H; Koivunoro, H; Auterinen, I; Serén, T; Kotiluoto, P; Palmans, H; Sharpe, P; Langguth, P; Hampel, G

    2015-01-01

    The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model. The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. The alanine detector can be used without

  6. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.

    1979-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other trace elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  7. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.; Bowman, W.W.; Zeh, C.W.

    1980-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program, which is sponsored and funded by the United States Department of Energy, Grand Junction Office. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  8. Method for measuring and evaluation dose equivalent rate from fast neutrons in mixed gamma-neutron fields around particles accelerators

    International Nuclear Information System (INIS)

    Cruceru, I.; Sandu, M.; Cruceru, M.

    1994-01-01

    A method for measuring and evaluation of doses and dose equivalent rate in mixed gamma- neutron fields is discussed in this paper. The method is basedon a double detector system consist of an ionization chamber with components made from a plastic scintillator, coupled to on photomultiplier. Generally the radiation fields around accelerators are complex, often consisting of many different ionizing radiations extending over a broad range of energies. This method solve two major difficulties: determination of response functions of radiation detectors; interpretation of measurement and determination of accuracy. The discrimination gamma-fast neutrons is assured directly without a pulse shape discrimination circuit. The method is applied to mixed fields in which particle energies are situated in the energy range under 20 MeV and an izotropic emision (Φ=10 4 -10 11 n.s -1 ). The dose equivalent rates explored is 0.01mSV--0.1SV

  9. A Novel Detector for High Neutron Flux Measurements

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. SU-E-T-568: Neutron Dose Survey of a Compact Single Room Proton Machine

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y; Prusator, M; Islam, M; Johnson, D; Ahmad, S [University of Oklahoma Health Sciences Center, Oklahoma City, OK (United States)

    2015-06-15

    Purpose: To ensure acceptable radiation limits are maintained for those working at and near the machine during its operation, a comprehensive radiation survey was performed prior to the clinical release of Mevion S250 compact proton machine at Stephenson Oklahoma Cancer Center. Methods: The Mevion S250 proton therapy system consists of the following: a superconducting cyclotron to accelerate the proton particles, a passive double scattering system for beam shaping, and paired orthogonal x-ray imaging systems for patient setup and verification via a 6D robotic couch. All equipment is housed within a single vault of compact design. Two beam delivery applicators are available for patient treatment, offering field sizes of as great as 14 cm and 25 cm in diameter, respectively. Typical clinical dose rates are between 1 and 2 Gy/min with a fixed beam energy of 250 MeV. The large applicator (25 cm in diameter) was used in conjunction with a custom cut brass aperture to create a 20 cm x 20 cm field size at beam isocenter. A 30 cm − 30 cm − 35 cm high density plastic phantom was placed in the beam path to mimic the conditions creating patient scatter. Measurements integrated-ambient-neutron-dose-equivalence were made with a SWENDII detector. Gantry angles of 0, 90 and 180 degrees, with a maximum dose rate of 150 MU/min (for large applicator) and beam configuration of option 1 (range 25 cm and 20 cm modulation), were selected as testing conditions. At each point of interest, the highest reading was recorded at 30 cm from the barrier surface. Results: The highest neutron dose was estimated to be 0.085 mSv/year at the console area. Conclusion: All controlled areas are under 5 mSv/year and the uncontrolled areas are under 1 mSv/year. The radiation protection provided by the proton vault is of sufficient quality.

  11. Estimate of neutron secondary doses received by patients in proton therapy: cases of ophthalmologic treatments

    International Nuclear Information System (INIS)

    Martinetti, F.

    2009-12-01

    This research thesis aims at assessing doses due to secondary neutrons and received by the organs of a patient which are located outside of the treatment field. The study focused on ophthalmological treatments performed at the Orsay proton therapy centre. A 75 eV beam line model has first been developed with the MCNPX Monte Carlo code. Several experimental validations of this model have been performed: proton dose distribution in a water phantom, ambient equivalent dose due to secondary neutrons and neutron spectra in the treatment room, and doses deposited by secondary neutrons in an anthropomorphous phantom. Simulations and measurements are in correct agreement. Then, a numeric assessment of secondary doses received by the patient's organs has been performed by using a MIRD-type mathematical phantom. These doses have been computed for several organs: the non-treated eye, the brain, the thyroid, and other parts of the body situated either in the front part of the body (the one directly exposed to neutrons generated in the treatment line) or deeper and further from the treatment field

  12. Monte Carlo simulation of secondary neutron dose for scanning proton therapy using FLUKA.

    Directory of Open Access Journals (Sweden)

    Chaeyeong Lee

    Full Text Available Proton therapy is a rapidly progressing field for cancer treatment. Globally, many proton therapy facilities are being commissioned or under construction. Secondary neutrons are an important issue during the commissioning process of a proton therapy facility. The purpose of this study is to model and validate scanning nozzles of proton therapy at Samsung Medical Center (SMC by Monte Carlo simulation for beam commissioning. After the commissioning, a secondary neutron ambient dose from proton scanning nozzle (Gantry 1 was simulated and measured. This simulation was performed to evaluate beam properties such as percent depth dose curve, Bragg peak, and distal fall-off, so that they could be verified with measured data. Using the validated beam nozzle, the secondary neutron ambient dose was simulated and then compared with the measured ambient dose from Gantry 1. We calculated secondary neutron dose at several different points. We demonstrated the validity modeling a proton scanning nozzle system to evaluate various parameters using FLUKA. The measured secondary neutron ambient dose showed a similar tendency with the simulation result. This work will increase the knowledge necessary for the development of radiation safety technology in medical particle accelerators.

  13. Determination of the dose equivalents due to neutrons produced during therapeutic irradiations with a Varian CLINAC 2500

    International Nuclear Information System (INIS)

    Carrillo, Ricardo E.

    1991-01-01

    This experiment it was designed to quantify that so important it is the dose equivalent deposited by the neutron flow that is generated by photonuclear reactions during therapeutic irradiations with X rays of produced high-energy for an accelerator Varian CLINAC 2500. This accelerator type is routinely used in the Department of Radiotherapy of the Hospital of the University of Wisconsin, E.U. The equivalent dose was measured in diverse towns of the room of irradiations using the activation of thin sheets of gold put in the center of plastic recipients full with water. In general, the recipients were 1 m or more than the floor and at distances still bigger than the walls. The irradiations were made using photons with the highest energy that you can select with this team - 24 MeV. The due equivalent dose to neutrons taken place here by the energy photons used they were measured and reported. (author)

  14. Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls

    International Nuclear Information System (INIS)

    Kylloenen, J.; Lindborg, L.

    2005-01-01

    Full text: The determination of both the low- and high-LET components of ambient dose equivalent in mixed fields is possible with microdosimetric methods. With the multiple-event microdosimetric variance covariance method the sum of those components are directly obtained also in pulsed beams. However, if the value of each dose component is needed a more extended analysis is required. The use of a graphite walled proportional detector in combination with a tissue-equivalent proportional counter in combination with the variance covariance method was here investigated. MCNP simulations were carried out for relevant energies to investigate the photon and neutron responses of the two detectors. The combined graphite and TEPC system, the Sievert instrument, was used for measurements at IRSN, Cadarache, in the workplace calibration fields of CANEL+, SIGMA, a Cf-252 and a moderated Cf(D 2 O,Cd) radiation field. The response of the instrument in various monoenergetic neutron fields is also known from measurements at PTB. The instrument took part in the measurement campaigns in workplace fields in the nuclear industry organized within the EVIDOS contract. The results are analyzed and the method of using a graphite detector compared with alternative methods of analysis is discussed. (author)

  15. Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Krstic, D.; Markovic, V.M.; Jovanovic, Z.; Milenkovic, B.; Nikezic, D.; Atanackovic, J.

    2014-01-01

    Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. The difference in evaluated dose in cancer and normal lung tissue suggests that BNCT could be applied for the treatment of cancers. The difference in exposure of cancer and healthy tissue can be observed, so the healthy tissue can be spared from damage. An absorbed dose ratio of metastatic tissue-to-the healthy tissue was ∼5. Absorbed dose to all other organs was low when compared with the lung dose. Absorbed dose depth distribution shows that BNC therapy can be very useful in the treatments for tumour. The ratio of the tumour absorbed dose and irradiated healthy tissue absorbed dose was also ∼5. It was seen that an elliptical neutron field was better irradiation choice. (authors)

  16. Pharmacokinetically guided dosing of (high-dose) chemotherapeutic agents

    NARCIS (Netherlands)

    Attema-de Jonge, M.E. (Milly Ellen)

    2004-01-01

    Due to variation in drug distribution, metabolism and elimination processes between patients, systemic exposure to chemotherapeutic agents may be highly variable from patient to patient after administration of similar doses. This pharmacokinetic variability may explain in part the large variability

  17. Evaluation of neutron doses beyond of primary shielding of rooms housing clinical linear accelerators

    International Nuclear Information System (INIS)

    Rezende, Gabriel Fonseca da Silva

    2011-01-01

    The growing need to build radiotherapy rooms in places with lack of available space leads to the necessity of unconventional solutions for the shielding projects. In most cases, adding metals to the primary barriers is the best way to shield the rooms properly. However, when photons with energies equal to or great than 10 MeV interact with nuclei of materials with high atomic number, neutrons are ejected and can result in a problem of radioprotection both inside and outside the room. Currently, the only empirical formula existing in the literature to assess the dose equivalent due to neutrons beyond the laminated barriers works only under very specific conditions, and a validation of this formula had not yet been done. In this work, the Monte Carlo code MCNPX was used to verify the validity of the above formula for cases of primary barriers containing lead or iron sheets in rooms that house linear accelerators with 10, 15 and 18 MV. Moreover, such a code was used to evaluate the coefficient of neutron production and tenth-value layer for neutrons in concrete, both parameters that directly influence the equation studied. The study results showed that over 90% of the values compared between the formula and the simulations present discrepancies above 100%, which led to conclude that the formula from the literature produces values that do not match the reality. In addition, there were inconsistencies in the parameters that make up the formula, leading to a need to review this formula in order to build a new model that will better represent the real case. (author)

  18. Antiproton Radiotherapy Peripheral Dose from Secondary Neutrons produced in the Annihilation of Antiprotons in the Target

    CERN Document Server

    Fahimian, Benjamin P; Keyes, Roy; Bassler, Niels; Iwamoto, Keisuke S; Zankl, Maria; Holzscheiter, Michael H

    2009-01-01

    The AD-4/ACE collaboration studies the biological effects of antiprotons with respect to a possible use of antiprotons in cancer therapy. In vitro experiments performed by the collaboration have shown an enhanced biological effectiveness for antiprotons relative to protons. One concern is the normal tissue dose resulting from secondary neutrons produced in the annihilation of antiprotons on the nucleons of the target atoms. Here we present the first organ specific Monte Carlo calculations of normal tissue equivalent neutron dose in antiproton therapy through the use of a segmented CT-based human phantom. The MCNPX Monte Carlo code was employed to quantify the peripheral dose for a cylindrical spread out Bragg peak representing a treatment volume of 1 cm diameter and 1 cm length in the frontal lobe of a segmented whole-body phantom of a 38 year old male. The secondary neutron organ dose was tallied as a function of energy and organ.

  19. Neutron activation analysis of high purity tellurium

    International Nuclear Information System (INIS)

    Gil'bert, Eh.N.; Verevkin, G.V.; Obrazovskij, E.G.; Shatskaya, S.S.

    1980-01-01

    A scheme of neutron activation analysis of high purity tellurium is developed. Weighed amount of Te (0.5 g) is irradiated for 20-40 hr in the flux of 2x10 13 neutron/(cm 2 xs). After decomposition of the sample impurities of gold and palladium are determined by the extraction with organic sulphides. Tellurium separation from the remaining impurities is carried out by the extraction with monothiobenzoic acid from weakly acidic HCl solutions in the presence of iodide-ions, suppressing silver extraction. Remaining impurity elements in the refined product are determined γ-spectrometrically. The method allows to determine 34 impurities with determination limits 10 -6 -10 -11 g

  20. Neutron production and dose rate in the IFMIF/EVEDA LIPAc injector beam commissioning

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Narita, Takahiro; Usami, Hiroki; Takahashi, Hiroki; Ochiai, Kentaro; Shinto, Katsuhiro; Kasugai, Atsushi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Okumura, Yoshikazu [IFMIF/EVEDA Project Team, Rokkasho-mura, Kamikita-gun, Aomori (Japan)

    2016-11-01

    Highlights: • A dedicated neutron production yield monitoring system for LIPAc has been developed. • The biological dose rate during operation of the LIPAc injector was analyzed. • The neutron streaming effect due to penetrations in the shielding wall was investigated. - Abstract: The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in Rokkasho, Japan, and the deuteron beam commissioning of the injector began in July 2015. Due to the huge beam current of 125 mA, a large amount of d-D neutrons are produced in the commissioning. The neutron streaming effect through pipe penetrations and underground pits may dominate the radiation dose at the outside of the accelerator vault during the injector operation. In the present study the effective dose rate expected during the injector commissioning was analyzed by a Monte Carlo calculation and compared with the measured value. For the comparison it is necessary to know the total neutron production yield in the accelerator vault, thus a dedicated neutron production yield monitoring system was developed. The yield obtained was smaller than that previously reported in a literature by a factor of a few and seems to depend on some beam conditions. From the comparison it was proved that the calculation always provides a conservative estimate and the dose rates in places where occupational works can always access and the controlled area boundary are expected to be far less than the legal criteria throughout the injector commissioning.

  1. A new online detector for estimation of peripheral neutron equivalent dose in organ

    Energy Technology Data Exchange (ETDEWEB)

    Irazola, L., E-mail: leticia@us.es; Sanchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009, Spain and Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Lorenzoli, M.; Pola, A. [Departimento di Ingegneria Nuclear, Politecnico di Milano, Milano 20133 (Italy); Bedogni, R. [Laboratori Nazionali di Frascati, Istituto Nazionale di Fisica Nucleare (INFN), Frascati Roma 00044 (Italy); Terrón, J. A. [Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Sanchez-Nieto, B. [Instituto de Física, Pontificia Universidad Católica de Chile, Santiago 4880 (Chile); Expósito, M. R. [Departamento de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Lagares, J. I.; Sansaloni, F. [Centro de Investigaciones Energéticas y Medioambientales y Tecnológicas (CIEMAT), Madrid 28040 (Spain)

    2014-11-01

    Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

  2. A new online detector for estimation of peripheral neutron equivalent dose in organ

    International Nuclear Information System (INIS)

    Irazola, L.; Sanchez-Doblado, F.; Lorenzoli, M.; Pola, A.; Bedogni, R.; Terrón, J. A.; Sanchez-Nieto, B.; Expósito, M. R.; Lagares, J. I.; Sansaloni, F.

    2014-01-01

    Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

  3. Fiber optics in high dose radiation fields

    International Nuclear Information System (INIS)

    Partin, J.K.

    1985-01-01

    A review of the behavior of state-of-the-art optical fiber waveguides in high dose (greater than or equal to 10 5 rad), steady state radiation fields is presented. The influence on radiation-induced transmission loss due to experimental parameters such as dose rate, total dose, irradiation history, temperature, wavelength, and light intensity, for future work in high dose environments are given

  4. High Intensity, Pulsed, D-D Neutron Generator

    International Nuclear Information System (INIS)

    Williams, D.L.; Vainionpaa, J.H.; Jones, G.; Piestrup, M.A.; Gary, C.K.; Harris, J.L.; Fuller, M.J.; Cremer, J.T.; Ludewigt, Bernhard A.; Kwan, J.W.; Reijonen, J.; Leung, K.-N.; Gough, R.A.

    2008-01-01

    Single ion-beam RF-plasma neutron generators are presented as a laboratory source of intense neutrons. The continuous and pulsed operations of such a neutron generator using the deuterium-deuterium fusion reaction are reported. The neutron beam can be pulsed by switching the RF plasma and/or a gate electrode. These generators are actively vacuum pumped so that a continuous supply of deuterium gas is present for the production of ions and neutrons. This contributes to the generator's long life. These single-beam generators are capable of producing up to 1E10 n/s. Previously, Adelphi and LBNL have demonstrated these generators applications in fast neutron radiography, Prompt Gamma Neutron Activation Analysis (PGNAA) and Neutron Activation Analysis (NAA). Together with an inexpensive compact moderator, these high-output neutron generators extend useful applications to home laboratory operations.

  5. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

  6. The relationship between dose rate and transformation induction in C3H/10T1/2 cells by TRIGA reactor fission neutrons at 0.3 Gy

    International Nuclear Information System (INIS)

    Balcer-Kubiczek, E.K.; Harrison, G.H.

    1989-01-01

    The authors present their own and other data showing dose-effect relations for cell survival and the induction of transformations in C3H/IOT 1/2 cells in exponential or stationary cultures after a range of high dose-rate irradiations with X-rays or AFRRI neutrons. (UK)

  7. Use of a high repetition rate neutron generator for in vivo body composition measurements via neutron inelastic scattering

    International Nuclear Information System (INIS)

    Kehayias, J.J.; Ellis, K.J.; Cohn, S.H.; Weinlein, J.H.

    1986-01-01

    A small D-T neutron generator with a high pulse rate is used for the in vivo measurement of body carbon, oxygen and hydrogen. The core of the neutron generator is a 13 cm-long Zetatron tube pulsed at a rate of 10 kHz delivering 10 3 to 10 4 neutrons per pulse. A target-current feedback system regulates the source of the accelerator to assure constant neutron output. Carbon is measured by detecting the 4.44 MeV γ-rays from inelastic scattering. The short half-life of the 4.44 MeV state of carbon requires detection of the γ-rays during the 10 μs neutron pulse. Generators with low pulsing rate were found inappropriate for carbon measurements because of their low duty-cycle (high neutron output during the pulse). In vivo measurements were performed with normal volunteers using a scanning bed facility for a dose less than 25 mrem. This technique offers medical as well as general bulk analysis applications. 8 refs., 5 figs

  8. Monte carlo calculation of energy-dependent response of high-sensitive neutron monitor, HISENS

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji; Ebisawa, Tohru; Kobayashi, Keiji; Koide, Hiroaki; Seo, Takeshi; Kawano, Shinji

    1988-01-01

    A highly sensitive neutron monitor system, HISENS, has been developed to measure leakage neutrons from nuclear facilities. The counter system of HISENS contains a detector bank which consists of ten cylindrical proportional counters filled with 10 atm 3 He gas and a paraffin moderator mounted in an aluminum case. The size of the detector bank is 56 cm high, 66 cm wide and 10 cm thick. It is revealed by a calibration experiment using an 241 Am-Be neutron source that the sensitivity of HISENS is about 2000 times as large as that of a typical commercial rem-counter. Since HISENS is designed to have a high sensitivity in a wide range of neutron energy, the shape of its energy dependent response curve cannot be matched to that of the dose equivalent conversion factor. To estimate dose equivalent values from neutron counts by HISENS, it is necessary to know the energy and angular characteristics of both HISENS and the neutron field. The area of one side of the detector bank is 3700 cm 2 and the detection efficiency in the constant region of the response curve is about 30 %. Thus, the sensitivity of HISENS for this energy range is 740 cps/(n/cm 2 /sec). This value indicates the extremely high sensitivity of HISENS as compared with exsisting highly sensitive neutron monitors. (Nogami, K.)

  9. Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

    Science.gov (United States)

    Pang, Bo; Becker, Frank

    2017-04-28

    Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  10. Neutron-photon mixed field dosimetry by TLD-700 glow curve analysis and its implementation in dose monitoring for Boron Neutron Capture Therapy (BNCT) treatments

    Energy Technology Data Exchange (ETDEWEB)

    Boggio, E. F.; Longhino, J. M. [Centro Atomico Bariloche, Departamento de Fisica de Reactores y Radiaciones / CNEA, Av. E. Bustillo Km 9.5, R8402AGP San Carlos de Bariloche (Argentina); Andres, P. A., E-mail: efboggio@cab.cnea.gov.ar [Centro Atomico Bariloche, Division Proteccion Radiologica / CNEA, Av. E. Bustillo Km 9.5, R8402AGP San Carlos de Bariloche (Argentina)

    2015-10-15

    BNCT is a cancerous cells selective, non-conventional radiotherapy modality to treat malignant tumors such as glioblastoma, melanoma and recurrent head and neck cancer. It consists of a two-step procedure: first, the patient is injected with a tumor localizing drug containing a non-radioactive isotope (Boron-10) with high slow neutron capture cross-section. In a second step, the patient is irradiated with neutrons, which are absorbed by the Boron-10 agent with the subsequently nuclear reaction B- 10(n,a)Li-7, thereby resulting in dose at cellular level due to the high-Let particles. The neutron fields suitable for BNCT are characterized by high neutron fluxes and low gamma dose. Determination of each component is not an easy task, especially when the volume of measurement is quite small or inaccessible for a miniature ionization chamber, for example. A method of measuring the photon and slow neutron dose(mainly by N-14 and B-10) from the glow curve (GC) analysis of a single {sup 7}LiF thermoluminescence detector is evaluated. This method was suggested by the group headed by Dr. Grazia Gambarini. The dosemeters used were TLD-600 ({sup 6}LiF:Mg,Ti with 95.6% {sup 6}Li) and TLD-700 ({sup 7}LiF:Mg,Ti with 99.9% {sup 7}LiF) from Harshaw. Photon dose measurement using the GC analysis method with TLD-700 in mixed fields requires the relation of the two main peaks of a TLD-600 GC shape obtained from an exposition to the same neutron field, and a photon calibrated GC with TLD-700. The requirements for slow neutron dose measurements are similar. In order to properly apply the GC analysis method at the Ra-6 Research Reactor BNCT facility, measurements were carried out in a standard water phantom, fully characterized on the BNCT beam by conventional techniques (activation detectors and paired ionization chambers technique). Next, the method was implemented in whole body dose monitoring of a patient undergoing a BNCT treatment, using a Bo MAb (Bottle Manikin Absorption) phantom

  11. Dose-equivalent response CR-39 track detector for personnel neutron dosimetry

    International Nuclear Information System (INIS)

    Oda, K.; Ito, M.; Yoneda, H.; Miyake, H.; Yamamoto, J.; Tsuruta, T.

    1991-01-01

    A dose-equivalent response detector based on CR-39 has been designed to be applied for personnel neutron dosimetry. The intrinsic detection efficiency of bare CR-39 was first evaluated from irradiation experiments with monoenergetic neutrons and theoretical calculations. In the second step, the radiator effect was investigated for the purpose of sensitization to fast neutrons. A two-layer radiator consisting of deuterized dotriacontane (C 32 D 66 ) and polyethylene (CH 2 ) was designed. Finally, we made the CR-39 detector sensitive to thermal neutrons by doping with orthocarbone (B 10 H 12 C 2 ), and also estimated the contribution of albedo neutrons. It was found that the new detector - boron-doped CR-39 with the two-layer radiator - would have a flat response with an error of about 70% in a wide energy region, ranging from thermal to 15 MeV. (orig.)

  12. Endorectal high dose rate brachytherapy quality assurance

    International Nuclear Information System (INIS)

    Devic, S.; Vuong, T.; Evans, M.; Podgorsak, E.

    2008-01-01

    We describe our quality assurance method for preoperative high dose rate (HDR) brachytherapy of endorectal tumours. Reproduction of the treatment planning dose distribution on a daily basis is crucial for treatment success. Due to the cylindrical symmetry, two types of adjustments are necessary: applicator rotation and dose distribution shift along the applicator axis. (author)

  13. Evaluation of energy response of neutron rem monitor applied to high-energy accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakane, Yoshihiro; Harada, Yasunori; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    A neutron rem monitor was newly developed for applying to the high-intensity proton accelerator facility (J-PARC) that is under construction as a joint project between the Japan Atomic Energy Research Institute and the High Energy Accelerator Research Organization. To measure the dose rate accurately for wide energy range of neutrons from thermal to high-energy region, the neutron rem monitor was fabricated by adding a lead breeder layer to a conventional neutron rem monitor. The energy response of the monitor was evaluated by using neutron transport calculations for the energy range from thermal to 150 MeV. For verifying the results, the response was measured at neutron fields for the energy range from thermal to 65 MeV. The comparisons between the energy response and dose conversion coefficients show that the newly developed neutron rem monitor has a good performance in energy response up to 150 MeV, suggesting that the present study offered prospects of a practical fabrication of the rem monitor applicable to the high intensity proton accelerator facility. (author)

  14. Orion, a high efficiency 4π neutron detector

    International Nuclear Information System (INIS)

    Crema, E.; Piasecki, E.; Wang, X.M.; Doubre, H.; Galin, J.; Guerreau, D.; Pouthas, J.; Saint-Laurent, F.

    1990-01-01

    In intermediate energy heavy ion collisions the multiplicity of emitted neutrons is strongly connected to energy dissipation and to impact parameter. We present the 4π detector ORION, a high efficiency liquid scintillator detector which permits to get information on the multiplicity of neutrons measured event-wise and on the spatial distribution of these neutrons [fr

  15. ACDOS2: a code for neutron-induced activities and dose rates

    International Nuclear Information System (INIS)

    Ruby, L.; Keney, G.S.; Lagache, J.C.

    1981-10-01

    In order to anticipate problems from the radioactivation of neutral beam sources as a result of testing, a code has been developed which calculates both the radioactivities produced and the dose rates resulting therefrom. The code ACDOS2 requires neutron source strength and spectral distribution as input, or alternately, the source strength can be calculated internally from an input of neutral beam source parameters. A variety of simple geometries can be specified, and up to 12 times of interest following the shutdown of the neutron source. Radiation attenuating and daughter radioactivities are treated accurately. ACDOS2 is also of use for neutron-induced radioactivation problems involving accelerators, fusion reactors, or fission reactors

  16. Monte Carlo simulated dose to the human body due to neutrons emitted in laser-fusion

    International Nuclear Information System (INIS)

    Gileadi, A.E.; Cohen, M.O.

    1977-01-01

    Considering a point neutron source located at a given distance from the human body, modeled by a 'standard reference man' phantom, neutron doses to the whole body, as well as to selected organs thereof, are determined, using the SAM-CE system, a Monte Carlo computer code, written in Fortran and designed to solve time, space and energy dependent neutron and gamma ray transport equations in complex three-dimensional geometrice. Collision density, energy deposition and dose are treated in the SAM-CE system as flux functionals. A special feature of SAM-CE is its use of the 'Combinatorial Geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All neutron and gamma ray cross section data, as well as gamma ray production data, are derived from the ENDF libraries. Both resolved and unresolved resonance parameters from ENDF neutron data files are treated automatically and extremely precise and detailed descriptions of cross section behavior is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux-averaged cross sections based on assumed flux distributions which may or may not be appropriate. The 'standard reference man', a heterogeneous phantom, uses simple geometric forms to approximate the shape and dimensions of the human body. Materials composition of each subregion representing a certain 'organ' is given. Typical values of neutron doses to the whole body and to selected 'organs' of interest are presented

  17. American National Standard: neutron and gamma-ray flux-to-dose rate factors

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This Standard presents data recommended for computing biological dose rates due to neutron and gamma-ray radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are given; the energy range for the gamma-ray conversion factors is 0.01 to 15 MeV. Specifically, this Standard is intended for use by shield designers to calculate wholebody dose rates to radiation workers and the general public. Establishing dose-rate limits is outside the scope of this Standard. Use of this Standard in cases where the dose equivalents are far in excess of occupational exposure guidelines is not recommended

  18. Dose equivalent response of personal neutron dosemeters as a function of angle

    International Nuclear Information System (INIS)

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

    1997-01-01

    The measured and calculated dose equivalent response as a function of angle has been examined for an albedo-type thermoluminescence dosemeter (TLD) that was exposed to unmoderated and D 2 O-moderated 252 Cf neutron sources while mounted on a 40 x 40 15 cm 3 polymethylmethacrylate phantom. The dosemeter used in this study is similar to many neutron personal dosemeters currently in use. The detailed construction of the dosemeter was modelled, and the dose equivalent response was calculated, using the MCNP code. Good agreement was found between the measured and calculated values of the relative dose equivalent angular response for the TLD albedo dosemeter. The relative dose equivalent angular response was also compared with the values of directional and personal dose equivalent as a function of angle published by Siebert and Schuhmacher. (author)

  19. High resolution Neutron and Synchrotron Powder Diffraction

    International Nuclear Information System (INIS)

    Hewat, A.W.

    1986-01-01

    The use of high-resolution powder diffraction has grown rapidly in the past years, with the development of Rietveld (1967) methods of data analysis and new high-resolution diffractometers and multidetectors. The number of publications in this area has increased from a handful per year until 1973 to 150 per year in 1984, with a ten-year total of over 1000. These papers cover a wide area of solid state-chemistry, physics and materials science, and have been grouped under 20 subject headings, ranging from catalysts to zeolites, and from battery electrode materials to pre-stressed superconducting wires. In 1985 two new high-resolution diffractometers are being commissioned, one at the SNS laboratory near Oxford, and one at the ILL in Grenoble. In different ways these machines represent perhaps the ultimate that can be achieved with neutrons and will permit refinement of complex structures with about 250 parameters and unit cell volumes of about 2500 Angstrom/sp3/. The new European Synchotron Facility will complement the Grenoble neutron diffractometers, and extend the role of high-resolution powder diffraction to the direct solution of crystal structures, pioneered in Sweden

  20. Initial measurement of site boundary neutron dose and comparison with calculations

    International Nuclear Information System (INIS)

    Degtyarenko, P.; Dotson, D.; May, R.; Schwahn, S.; Stapleton, G.

    1996-01-01

    For most accelerators adequate side shielding can be provided at minimal cost to meet the most aggressive radiation protection regulations and, further, the likely requirement to increase shielding thickness still more at a later date can be done usually by heaping more earth or applying local shielding at minimal expense and inconvenience. This moderately happy state of affairs does not unfortunately hold true with roof shielding. The cost of roof shielding is largely predicated on the roof span and the necessary structural engineering requirements for its support. These measures can be extremely expensive and where one is dealing with the rather extensive unsupported spans typical of experimental halls devoted to experiments with high energy electron beams; it is necessary to specify the roof thickness as carefully as possible with the constant concern that adding more earth later is not likely to be possible without rebuilding the hall. Because of the nature of roof skyshine, and for most high energy accelerator facilities neutron skyshine, the effect of the radiation is likely to extend to the facility fence-line where one is concerned about the exposure of the general population. Very properly the dose limit for the general population is set at a rather low value (1 mSv y -1 ) and in order for the Jefferson Lab (JLab) to ensure strict compliance with this limit they have a design goal for the fence line of 0.1 mSv y -1 . However, because natural neutron backgrounds are low (30--40 microSv y -1 ) and the methods of detection and measurement permit rejection of background interference from photons, they can measure the JLab produced neutron radiation with good sensitivity and precision

  1. Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source

    International Nuclear Information System (INIS)

    Li, Gang; Xu, Jiayun; Zhang, Jie

    2015-01-01

    Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3 He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3 He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. - Highlights: • The scope of the affected area for

  2. Swelling of spinel after low-dose neutron irradiation

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Clinard, F.W. Jr.; Itoh, N.; Greenwood, L.R.

    1986-01-01

    Swelling was determined in samples of single-crystal MgAl 2 O 4 spinel, irradiated to doses as high as 8 x 10 22 n/m 2 (E > 0.1 MeV) at approx. =50 0 C in the Omega West Reactor. Swelling effectively saturated at approx. =2 x 10 22 n/m 2 which corresponds to a damage level of only approx. =2 x 10 -3 dpa. In addition subsequent measurements after irradiation have revealed that the samples continued swelling for several weeks. These results imply that irradiation defects begin to interact by recombination and aggregation at low damage levels in this material at 50 0 C and perhaps continue to cluster at room temperature after irradiation. Rate equations have been employed to determine defect concentrations at saturation. Results to date show that the observed swelling is consistent with the number of surviving defects if swelling per Frenkel defect pair is taken to be one atomic volume

  3. Californium-252 neutron activation analysis of high-level processed nuclear tank waste

    International Nuclear Information System (INIS)

    Troyer, G.L.; Purcell, M.A.

    2000-01-01

    The basis for production assessment of the vitrification of Hanford nuclear fuel reprocessing wastes will be high-precision measurements of the elemental sodium content. However, the chemical analysis of both radioactive and nonradioactive components in nuclear waste can be challenged by high radiation dose rates. The dose rates compromise many analytical techniques as well as pose personnel dosimetry risks. In many cases, reduction of dose rates through dilution compromises the precision and sensitivity for certain key components. The use of neutron activation analysis (NAA) provides a method of analysis that avoids the need for dilutions or extensive sample preparation. These waste materials also contain trace quantities of fissionable isotopes, which, through neutron activation, can be estimated by delayed neutron counting of fissioned fragments

  4. Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

    International Nuclear Information System (INIS)

    Singh, B.N.; Horsewell, A.; Toft, P.; Edwards, D.J.

    1995-01-01

    Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5x10 21 to 1.5x10 24 n/m 2 (E>1 MeV) with a flux of 2.5x10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of similar 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ∼ =0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources. ((orig.))

  5. Progress in high-dose radiation dosimetry

    International Nuclear Information System (INIS)

    Ettinger, K.V.; Nam, J.W.; McLaughlin, W.L.; Chadwick, K.H.

    1981-01-01

    The last decade has witnessed a deluge of new high-dose dosimetry techniques and expanded applications of methods developed earlier. Many of the principal systems are calibrated by means of calorimetry, although production of heat is not always the final radiation effect of interest. Reference systems also include a number of chemical dose meters: ferrous sulphate, ferrous-cupric sulphate, and ceric sulphate acidic aqueous solutions. Requirements for stable and reliable transfer dose meters have led to further developments of several important high-dose systems: amino acids and saccharides analysed by ESR or lyoluminescence, thermoluminescent materials, radiochromic dyes and plastics, ceric-cerous solutions analysed by potentiometry, and ethanol-chlorobenzene solutions analysed by high-frequency oscillometry. A number of other prospective dose meters are also treated in this review. In addition, an IAEA programme of high-dose standardization and intercomparison for industrial radiation processing is described. (author)

  6. Transit dose calculations around straight cathethers in high dose ...

    African Journals Online (AJOL)

    A fast and non-destructive technique based on thermal neutron moderation has been used for determining the total hydrogen content in two types of red palm oil (dzomi and amidze) and palm kernel oil produced by traditio-nal methods in Ghana. An equipment consisting of an 241Am-Be neutron source and 3He neutron ...

  7. Dose distributions in thorax inhomogeneity for fast neutron beam from NIRS cyclotron

    International Nuclear Information System (INIS)

    Kutsutani-Nakamura, Yuzuru; Furukawa, Shigeo; Iinuma, T.A.; Kawashima, Katsuhiro; Hoshino, Kazuo; Hiraoka, Takeshi; Maruyama, Takashi; Sakashita, Kunio; Tsunemoto, Hiroshi

    1990-01-01

    The power law tissue-air ratio (TAR) method developed by Batho appears to be practical use for inhomogeneity corrections to the dose calculated in a layered media for photon beam therapy. The validity was examined in applying the modified power law TAR and the isodose shift methods to the dose calculation in thorax tissue inhomogeneity containing the boundary region for fast neutron beam. The neutron beam is produced by bombarding a thick beryllium target with 30 MeV deuterons. Lung phantom was made of granulated tissue equivalent plastic, which resulted in density of 0.30 and 0.60 g/cm 3 . Depth dose distributions for neutron beam were measured in thorax phantom by an air-filled cylindrical ionization chamber with TE plastic wall. The power law TAR method considering TAR of zero depth at boundary was compared with the measured data and a good result was obtained that the calculated dose was within ±3 % against the measured. But the isodose shift method is not so good for dose calculation in thorax tissue inhomogeneity using fast neutron beam. (author)

  8. 6LiF sandwich type detectors for low dose individual monitoring in mixed neutron-photon fields

    International Nuclear Information System (INIS)

    Olko, P.; Budzanowski, M.; Bilski, P.; Burgkhardt, B.; Piesch, E.

    1994-01-01

    ICRP Publication 60 recommends the reduction of the annual dose limit for occupational exposure from 50 to 20 mSv and a doubling of the quality factor for medium energy neutrons. If occupational doses are evaluated every month (which is obligatory e.g. in Germany and in Poland), the individual neutron dosemeter will have to measure neutron doses in the range of 100 μSv. No commercially available, automatic individual dosimetry monitoring system exists that fulfils this requirement. Some of the parameters which influence the evaluation of the neutron dose from readings of TL dosemeters have been studied in order to decrease the variance of the measured neutron signal. In mixed neutron-photon fields, clear separation of the neutron component from the total reading depends also on the uncertainty of the gamma dose measurements. While the thermal albedo neutrons are absorbed mostly at the surface of the 6 LiF detector, the reduction of the detector thickness results in a decrease of its photon sensitivity, while its neutron sensitivity is almost principally maintained. As a consequence, the uncertainty of gamma dose contributes with lower weight to the variance of the evaluated neutron signal. First tests of an optimised 200 μm thick sandwich detector and 0.9 mm thick standard LiF chips were made at low neutron and photon dose ranges using different readers, in order to determine the uncertainty versus dose for different neutron-photon combinations. The conditions under which the new sandwich type detectors may improve albedo neutron dosimetry are demonstrated. (Author)

  9. Novel Concrete Chemistry Achieved with Low Dose Gamma Radiation Curing and Resistance to Neutron Activation

    Science.gov (United States)

    Burnham, Steven Robert

    As much as 50% of ageing-related problems with concrete structures can be attributed to con-struction deficiencies at the time of placement. The most influential time affecting longevity of concrete structures is the curing phase, or commonly the initial 28 days following its placement. A novel advanced atomistic analysis of novel concrete chemistry is presented in this dissertation with the objective to improve concrete structural properties and its longevity. Based on experiments and computational models, this novel concrete chemistry is discussed in two cases: (a) concrete chemistry changes when exposed to low-dose gamma radiation in its early curing stage, thus improving its strength in a shorter period of time then curing for the conventional 28 days; (b) concrete chemistry is controlled by its atomistic components to assure strength is not reduced but that its activation due to long-term exposure to neutron flux in nuclear power plants is negligible. High dose gamma radiation is well documented as a degradation mechanism that decreases concrete's compressive strength; however, the effects of low-dose gamma radiation on the initial curing phase of concrete, having never been studied before, proved its compressive strength increases. Using a 137 Cs source, concrete samples were subjected to gamma radiation during the initial curing phase for seven, 14, and 28 days. The compressive strength after seven days is improved for gamma cured concrete by 24% and after 14 days by 76%. Concrete shows no improvement in compressive strength after 28 days of exposure to gamma radiation, showing that there is a threshold effect. Scanning Electron Microscopy is used to examine the microstructure of low-dose gamma radiation where no damage to its microstructure is found, showing no difference between gamma cured and conventionally cured concrete. Molecular dynamics modeling based on the MOPAC package is used to study how gamma radiation during the curing stage improves

  10. Measurementof photo-neutron dose from an 18-MV medical linac using a foil activation method in view of radiation protection of patients

    International Nuclear Information System (INIS)

    Yuecel, Haluk; Kolbasi, Asuman; Yueksel, Alptug Oezer; Cobanbas, Ibrahim; Kaya, Vildan

    2016-01-01

    High-energy linear accelerators are increasingly used in the medical field. However, the unwanted photo-neutrons can also be contributed to the dose delivered to the patients during their treatments. In this study, neutron fluxes were measured in a solid water phantom placed at the isocenter 1-m distance from the head of an 18-MV linac using the foil activation method. The produced activities were measured with a calibrated well-type Ge detector. From the measured fluxes, the total neutron fluence was found to be (1.17 ± 0.06) X 10 7 n/cm 2 per Gy at the phantom surface in a 20 X 20 cm 2 X-ray field size. The maximum photo-neutron dose was measured to be 0.67 ± 0.04 mSv/Gy at d max = 5 cm depth in the phantom at isocenter. The present results are compared with those obtained for different field sizes of 10 X 10cm 2 , 15 X 15cm 2 , and 20 X 20cm 2 from 10-, 15-, and 18-MV linacs. Additionally, ambient neutron dose equivalents were determined at different locations in the room and they were found to be negligibly low. The results indicate that the photo-neutron dose at the patient position is not a negligible fraction of the therapeutic photon dose. Thus, there is a need for reduction of the contaminated neutron dose by taking some additional measures, for instance, neutron absorbing-protective materials might be used as aprons during the treatment

  11. Field calibration of a TLD albedo dosemeter in the high-energy neutron field of CERF

    International Nuclear Information System (INIS)

    Haninger, T.; Kleinau, P.; Haninger, S.

    2017-01-01

    The new albedo dosemeter-type AWST-TL-GD 04 has been calibrated in the CERF neutron field (CERN-EU high-energy Reference Field). This type of albedo dosemeter is based on thermoluminescent detectors (TLDs) and used by the individual monitoring service of the Helmholtz Zentrum Muenchen (AWST) since 2015 for monitoring persons, who are exposed occupationally against photon and neutron radiation. The motivation for this experiment was to gain a field specific neutron correction factor N n for workplaces at high-energy particle accelerators. N n is a dimensionless factor relative to a basic detector calibration with 137 Cs and is used to calculate the personal neutron dose in terms of H p (10) from the neutron albedo signal. The results show that the sensitivity of the albedo dosemeter for this specific neutron field is not significantly lower as for fast neutrons of a radionuclide source like 252 Cf. The neutron correction factor varies between 0.73 and 1.16 with a midrange value of 0.94. The albedo dosemeter is therefore appropriate to monitor persons, which are exposed at high-energy particle accelerators. (authors)

  12. Polarized neutron reflectometry in high magnetic fields

    International Nuclear Information System (INIS)

    Fritzsche, H.

    2005-01-01

    A simple method is described to maintain the polarization of a neutron beam on its way through the large magnetic stray fields produced by a vertical field of a cryomagnet with a split-coil geometry. The two key issues are the proper shielding of the neutron spin flippers and an additional radial field component in order to guide the neutron spin through the region of the null point (i.e., point of reversal for the vertical field component). Calculations of the neutron's spin rotation as well as polarized neutron reflectometry experiments on an ErFe 2 /DyFe 2 multilayer show the perfect performance of the used setup. The recently commissioned cryomagnet M5 with a maximum vertical field of up to 7.2 T in asymmetric mode for polarized neutrons and 9 T in symmetric mode for unpolarized neutrons was used on the C5 spectrometer in reflectometry mode, at the NRU reactor in Chalk River, Canada

  13. Dose and dose rate extrapolation factors for malignant and non-malignant health endpoints after exposure to gamma and neutron radiation

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Van; Little, Mark P. [National Cancer Institute, Radiation Epidemiology Branch, Rockville, MD (United States)

    2017-11-15

    non-malignant, show downward curvature in the dose response, and for most endpoints this is statistically significant (p < 0.05). Associated with this, the low-dose extrapolation factor associated with neutron exposure is generally statistically significantly less than 1 for most malignant and non-malignant endpoints, with central estimates mostly in the range 0.1-0.9. In contrast to the situation at higher dose rates, there are statistically non-significant decreases of risk per unit dose at gamma dose rates of less than or equal to 5 mGy/h for most malignant endpoints, and generally non-significant increases in risk per unit dose at gamma dose rates ≤5 mGy/h for most non-malignant endpoints. Associated with this, the dose-rate extrapolation factor, the ratio of high dose-rate to low dose-rate (≤5 mGy/h) gamma dose response slopes, for many tumour sites is in the range 1.2-2.3, albeit not statistically significantly elevated from 1, while for most non-malignant endpoints the gamma dose-rate extrapolation factor is less than 1, with most estimates in the range 0.2-0.8. After neutron exposure there are non-significant indications of lower risk per unit dose at dose rates ≤5 mGy/h compared to higher dose rates for most malignant endpoints, and for all tumours (p = 0.001), and respiratory tumours (p = 0.007) this reduction is conventionally statistically significant; for most non-malignant outcomes risks per unit dose non-significantly increase at lower dose rates. Associated with this, the neutron dose-rate extrapolation factor is less than 1 for most malignant and non-malignant endpoints, in many cases statistically significantly so, with central estimates mostly in the range 0.0-0.2. (orig.)

  14. Design and optimization of a beam shaping assembly for BNCT based on D-T neutron generator and dose evaluation using a simulated head phantom.

    Science.gov (United States)

    Rasouli, Fatemeh S; Masoudi, S Farhad

    2012-12-01

    A feasibility study was conducted to design a beam shaping assembly for BNCT based on D-T neutron generator. The optimization of this configuration has been realized in different steps. This proposed system consists of metallic uranium as neutron multiplier, TiF(3) and Al(2)O(3) as moderators, Pb as reflector, Ni as shield and Li-Poly as collimator to guide neutrons toward the patient position. The in-air parameters recommended by IAEA were assessed for this proposed configuration without using any filters which enables us to have a high epithermal neutron flux at the beam port. Also a simulated Snyder head phantom was used to evaluate dose profiles due to the irradiation of designed beam. The dose evaluation results and depth-dose curves show that the neutron beam designed in this work is effective for deep-seated brain tumor treatments even with D-T neutron generator with a neutron yield of 2.4×10(12) n/s. The Monte Carlo Code MCNP-4C is used in order to perform these calculations. Copyright © 2012 Elsevier Ltd. All rights reserved.

  15. Report on high energy neutron dosimetry workshop

    International Nuclear Information System (INIS)

    Alvar, K.R.; Gavron, A.

    1993-01-01

    The workshop was called to assess the performance of neutron dosimetry per the responses from ten DOE accelerator facilities to an Office of Energy Research questionnaire regarding implementation of a personnel dosimetry requirement in DRAFT DOE 5480.ACC, ''Safety of Accelerator Facilities''. The goals of the workshop were to assess the state of dosimetry at high energy accelerators and if such dosimetry requires improvement, to reach consensus on how to proceed with such improvements. There were 22 attendees, from DOE Programs and contract facilities, DOE, Office of Energy Research (ER), Office of Environmental Safety and Health (EH), Office of Fusion Energy, and the DOE high energy accelerator facilities. A list of attendees and the meeting agenda are attached. Copies of the presentations are also attached

  16. Tolerance of human spinal cord to high-energy p(66)Be(49) neutrons

    International Nuclear Information System (INIS)

    Cohen, L.; Haken, R.K.T.; Mansell, J.A.; Yalavarthi, D.; Hendrickson, F.R.; Awschalom, M.

    1985-01-01

    A total of 76 patients with cancer of the head and neck have been irradiated at the Fermilab Neutron Therapy Facility using high-energy neutrons. Dose, time and cord-length factors were determined for each patient from their individual treatment plans. Cord doses ranged from 5 to 16 Gy in 8 to 24 fractions over 6 to 70 days. The treated lengths were between 5 and 15 cm. No myelopathy was seen during follow-up periods ranging from 2 to 6 years. By comparing these observations with published data, the upper and lower limits for spinal cord tolerance to neutrons can be determined. There is no apparent risk of injury with cord doses under 13 Gy

  17. Evaluation of energy responses for neutron dose-equivalent meters made in Japan

    International Nuclear Information System (INIS)

    Saegusa, J.; Yoshizawa, M.; Tanimura, Y.; Yoshida, M.; Yamano, T.; Nakaoka, H.

    2004-01-01

    Energy responses of three types of Japanese neutron dose-equivalent (DE) meters were evaluated by Monte Carlo simulations and measurements. The energy responses were evaluated for thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and also for neutrons from such radionuclide sources as 252 Cf and 241 Am-Be. The calculated results were corroborated with the measured ones. The angular dependence of the response and the DE response were also evaluated. As a result, reliable energy responses were obtained by careful simulations of the proportional counter, moderator and absorber of the DE meters. Furthermore, the relationship between pressure of counting gas and response of the DE meter was discussed. By using the obtained responses, relations between predicted readings of the DE meters and true DE values were studied for various workplace spectra

  18. Comparison of Out-Of-Field Neutron Equivalent Doses in Scanning Carbon and Proton Therapies for Cranial Fields

    DEFF Research Database (Denmark)

    Athar, B.; Henker, K.; Jäkel, O.

    2010-01-01

    Purpose: The purpose of this analysis is to compare the secondary neutron lateral doses from scanning carbon and proton beam therapies. Method and Materials: We simulated secondary neutron doses for out-of-field organs in an 11-year old male patient. Scanned carbon and proton beams were simulated...

  19. Limiting values for the RBE of fission neutrons at low doses for life shortening in mice

    International Nuclear Information System (INIS)

    Storer, J.B.; Mitchell, T.J.

    1984-01-01

    The authors have analyzed recently published data on the effects of low doses of fission neutrons on the mean survival times of mice. The analysis for single-dose exposures was confined to doses of 20 rad or less, while for fractionated exposures only total doses of 80 rad or less were considered. They fitted the data to the frequently used power function model: life shortening = βD/sup γ/, where D is the radiation dose. They show that, at low doses per fraction, either the effects are not additive or the dose-effect curve for single exposures cannot show a greater negative curvature than about the 0.9 power of dose. Analysis of the data for γ rays showed that an exponent of 1.0 gave an acceptable fit. They conclude that at neutron doses of 20 rad or less the RBE for life shortening is constant and ranges from 13 to 22 depending on mouse strain and sex

  20. Neutron and gamma-ray dose-rates from the Little Boy replica

    International Nuclear Information System (INIS)

    Plassmann, E.A.; Pederson, R.A.

    1984-01-01

    We report dose-rate information obtained at many locations in the near vicinity of, and at distances out to 0.64 km from, the Little Boy replica while it was operated as a critical assembly. The measurements were made with modified conventional dosimetry instruments that used an Anderson-Braun detector for neutrons and a Geiger-Mueller tube for gamma rays with suitable electronic modules to count particle-induced pulses. Thermoluminescent dosimetry methods provide corroborative data. Our analysis gives estimates of both neutron and gamma-ray relaxation lengths in air for comparison with earlier calculations. We also show the neutron-to-gamma-ray dose ratio as a function of distance from the replica. Current experiments and further data analysis will refine these results. 7 references, 8 figures

  1. Effects of split fast neutron doses on the liver cells of albino Swiss mice

    International Nuclear Information System (INIS)

    Abdelmeguid, N.; Ramadan, A.A.; El-Khatib, A.M.

    1990-01-01

    The effect of neutron doses from a compact D-T neutron generator on the liver cells of adult male and female albino Swiss mice was investigated. Fast neutrons (14.5 MeV) were delivered to the whole body in a single dose or in two, four, six or eight equal doses separated by 3-day intervals. The lowest dose, 100 rem, was given over an exposure time of 6 hours and was then steadily raised to 912 rem over an exposure time of 48 hours. During exposure the neutron flux was controlled by the activation foil technique. The animals were killed for testing after each irradiation. Histological examination of the hepatocytes with a light microscope showed marked degenerative changes only after the longer irradiation periods (24, 36 and 48 h). Electron microscopy showed cytological (cytoplasmic and nuclear) changes in the hepatocytes after only 12 hours' irradiation. Densitometric scans of electron micrographs of control and 12 h-irradiated livers indicated that the control hepatocyte interphase nucleus contains approximately 72% heterochromatin, while the irradiated nucleus contains only 64% heterochromatin. (author). 13 figs., 1 tab., 18 refs

  2. ACDOS2: an improved neutron-induced dose rate code

    International Nuclear Information System (INIS)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere

  3. ACDOS2: an improved neutron-induced dose rate code

    Energy Technology Data Exchange (ETDEWEB)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  4. The Clatterbridge high-energy neutron therapy facility: specification and performance

    International Nuclear Information System (INIS)

    Bonnett, D.E.; Blake, S.W.; Shaw, J.E.; Bewley, D.K.

    1988-01-01

    A high energy neutron therapy facility has been installed at the Douglas Cyclotron Centre, Clatterbridge Hospital Merseyside, to extend M.R.C. clinical trials of fast neutrons. The neutron beam is produced by bombarding a beryllium target with 62 MeV protons. The target is isocentrically mounted with potential for 360 0 rotation, with a fully variable collimator, giving a range of rectilinear field sizes from 5 cm x 5 cm to 30 cm x 30 cm. Basic neutron beam data including output, field flatness, penumbra and depth-dose data have been measured. For a 10 cm x 10 cm field, 50% depth dose occurs at 16.2 cm in water and output is 1.63 cGy μ A -1 min -1 at maximum dose depth. Effectiveness of the target shielding and neutron-induced radioactivity in the treatment head were also measured. It is concluded that the equipment meets design specifications and fully satisfies criticisms of earlier neutron therapy equipment. A full radiation survey showed that radiation levels present no significant staff hazard. (UK)

  5. Secondary radiation dose during high-energy total body irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Janiszewska, M.; Raczkowski, M. [Lower Silesian Oncology Center, Medical Physics Department, Wroclaw (Poland); Polaczek-Grelik, K. [University of Silesia, Medical Physics Department, Katowice (Poland); Szafron, B.; Konefal, A.; Zipper, W. [University of Silesia, Department of Nuclear Physics and Its Applications, Katowice (Poland)

    2014-05-15

    The goal of this work was to assess the additional dose from secondary neutrons and γ-rays generated during total body irradiation (TBI) using a medical linac X-ray beam. Nuclear reactions that occur in the accelerator construction during emission of high-energy beams in teleradiotherapy are the source of secondary radiation. Induced activity is dependent on the half-lives of the generated radionuclides, whereas neutron flux accompanies the treatment process only. The TBI procedure using a 18 MV beam (Clinac 2100) was considered. Lateral and anterior-posterior/posterior-anterior fractions were investigated during delivery of 2 Gy of therapeutic dose. Neutron and photon flux densities were measured using neutron activation analysis (NAA) and semiconductor spectrometry. The secondary dose was estimated applying the fluence-to-dose conversion coefficients. The main contribution to the secondary dose is associated with fast neutrons. The main sources of γ-radiation are the following: {sup 56}Mn in the stainless steel and {sup 187}W of the collimation system as well as positron emitters, activated via (n,γ) and (γ,n) processes, respectively. In addition to 12 Gy of therapeutic dose, the patient could receive 57.43 mSv in the studied conditions, including 4.63 μSv from activated radionuclides. Neutron dose is mainly influenced by the time of beam emission. However, it is moderated by long source-surface distances (SSD) and application of plexiglass plates covering the patient body during treatment. Secondary radiation gives the whole body a dose, which should be taken into consideration especially when one fraction of irradiation does not cover the whole body at once. (orig.) [German] Die zusaetzliche Dosis durch sekundaere Neutronen- und γ-Strahlung waehrend der Ganzkoerperbestrahlung mit Roentgenstrahlung aus medizinischen Linearbeschleunigern wurde abgeschaetzt. Bei der Emission hochenergetischer Strahlen zur Teletherapie finden hauptsaechlich im Beschleuniger

  6. High resolution neutron tomography applied to tooth fillings on real teeth by use of neutron lens

    International Nuclear Information System (INIS)

    Masschaele, B.; Cauwels, P.; Mondelaers, W.; Baechler, S.; Jolie, J.; Materna, T.

    2000-01-01

    Today tomography is a well known technique for nondestructive analysis of samples. By taking several X-ray pictures from an object, it is possible to make a 3D reconstruction. The same thing can be done with neutrons. Since very recent it is possible to produce a high-flux neutron beam. By looking at the attenuation of the neutron beam in the sample from different angles, it is possible to make a neutron tomography. The properties of neutrons are so much different from X-rays that a new era in tomography has started. Where X-rays have a hard time penetrating samples containing heavy elements (Pb, Bi, U, Hg, Au), neutrons just seem to walk through. But when the neutrons encounter samples containing light compounds like water, oil, paper, B, Li,... they are easily absorbed. This makes the use of neutrons for imaging complementary to the well known X-ray imaging. The most used tooth filling material nowadays is amalgam. Amalgam is a mixture of different metals, like silver, tin, copper, mercury. Mercury is dangerous for the human body when it enters the blood stream. These fillings are very dense and X-rays have a very hard time penetrating it. Neutrons are the ideal probe for investigation of these high density regions. The result of the tomography reveals information on the long term stability of amalgam fillings and could help the still ongoing debate on the safety of the fillings. (author)

  7. The Efficiency of the BC-720 Scintillator in a High-Energy (20--800 MeV) Accelerator Neutron Field

    Energy Technology Data Exchange (ETDEWEB)

    Miles, Leslie H. [Univ. of Missouri, Columbia, MO (United States)

    2005-12-01

    High-energy neutron doses (>20 MeV) are of little importance to most radiation workers. However, space and flight crews, and people working around medical and scientific accelerators receive over half of their radiation dose from high-energy neutrons. Unfortunately, neutrons are difficult to measure, and no suitable dosimetry has yet been developed to measure this radiation. In this paper, basic high-energy neutron interactions, characteristics of high-energy neutron environments, present neutron dosimetry, and quantities used in neutron dosimetry are discussed before looking into the potential of the BC-720 scintillator to improve dosimetry. This research utilized 800 MeV protons impinging upon the WNR Facility spallation neutron source at Los Alamos National Laboratory. Time-of-flight methods and a U-238 Fission Chamber were used to aid evaluation of the efficiency of the BC-720. Results showed that the efficiency is finite over the 20–650 MeV energy region studied, although it decreases by a factor of ten between 40 and 100 MeV. This limits the use of this dosimeter to measure doses at sitespecific locations. It also encourages modifications to use this dosimeter for any unknown neutron field. As such, this dosimeter has the potential for a small, lightweight, real-time dose measurement, which could impact neutron dosimetry in all high-energy neutron environments.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  9. Neutron/gamma dose separation by the multiple-ion-chamber technique

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1983-01-01

    Many mixed n/γ dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and 60 Co gamma ray irradiation in air. The MORSE-CG coupled n/γ three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references

  10. High-dose irradiation of food

    International Nuclear Information System (INIS)

    Diehl, J.F.

    1999-01-01

    Studies performed on behalf of the International Project on Food Irradiation in the period from 1971 until 1980 resulted in the concluding statement that ''.the irradiation of any food commodity up to an overall average dose of 10 kGy presents no toxicological hazard; hence, toxicological testing of foods so treated is no longer required.'' Since then, licenses for food irradiation have been restricted to this maximum dose in any country applying this technology. Further testing programmes have been carried out investigating the wholesomeness or hazards of high-dose irradiation, but there has been little demand so far by the food industry for licensing of high-dose irradiation, as there is only a small range of products whose irradiation at higher doses offers advantages for given, intended use. These include eg. spices, dried herbs, meat products in flexible pouch packagings for astronauts, or patients with immune deficiencies. (orig./CB) [de

  11. The effect of low-dose neutron irradiation on extracellular matrix

    International Nuclear Information System (INIS)

    Chen Tiehe; Lu Yongjie; Chai Mingsheng; Peng Wulin; Yang Yifang; Pan Yan; Chen Jinguo

    2003-01-01

    Projective: To study the effect of neutron irradiation on extracellular matrix. Methods: 120 male wistar rats were divided into four groups at random, and then exposed to neutron of 252 Cf-source at the doses of 0, 0.29, 0.62 and 1.20 Gy, respectively. After the exposure of 3 days, 1 month and 2 months, the rats were sacrificed and lung tissue specimens stored at -30 degree C. Hyaluronan, laminin, type III procollagen and type IV collagen in the lung tissue were detected by the method of radioimmunoassay. Results: The differences of the levels of hyaluronan in lung tissue among the groups were unsignificant. The levels of laminin in 0.29, 0.62 and 1.20 Gy groups after the 3-day exposure were remarkably different to those of the control group, and unable to recover completely even 2 months after the exposure. The levels of type IV collagen in higher three irradiated groups were all higher, but not significantly. The levels of type III procollagen in the early stage after exposure were higher, and later they lowered. Conclusion: The levels of some components of extracellular matrix in the lung tissue of rat can be changed by low-dose of neutron irradiation, but their variational modes and degrees depend on the dose of neutron irradiation and the length of period after exposure

  12. Concomitant chemoradiotherapy with high dose rate brachytherapy ...

    African Journals Online (AJOL)

    Concomitant chemoradiotherapy with high dose rate brachytherapy as a definitive treatment modality for locally advanced cervical cancer. T Refaat, A Elsaid, N Lotfy, K Kiel, W Small Jr, P Nickers, E Lartigau ...

  13. A sensitivity study on neutron flux variation due to 10B concentration in dose calculation for BNCT

    International Nuclear Information System (INIS)

    Jung, Sang Hoon

    2006-02-01

    The effects of inclusion of 10 B concentration on neutron flux and dose in dose calculation were studied. In order to provide the quantitative effects of inclusion of 10 B concentrations on depressions of neutron and photon flux and dose, the fluxes and doses with voxel head phantoms for various 10 B concentrations homogeneously distributed were calculated by using MCNPX simulations. A lithium target system and beam shaping assembly, which have been developed at the Hanyang University, were used as epithermal neutron beam. The calculation results show that the neutron flux at the center of the head phantom decreases by approximately 5.4% per 10 ppm of 10 B concentration in comparison with the neutron flux in the case of boron-free. It was also observed that the tissue dose at the center of the head phantom is depressed by approximately 4.7% per 10 ppm of the 10 B concentration and the tumor dose by approximately 5.3% per 10 ppm. According to depth of tumors, it was observed that the depressions of the doses in the tumors are ranged in 3.7 ∼ 9.2%. The dose calculations in the case of boron-free show that it is overestimated in comparison with the dose calculations in the cases of the inclusion of 10 B concentrations for the normal tissue and the tumors. Therefore, in dose calculation for BNCT, the depressions of neutron flux and dose should be considered. The results in this study are available to setting up the depression ratios which can be used for converting neutron and gamma fluxes and doses in phantom with boron free into the fluxes and doses in phantom with inclusion of 10 B concentrations in treatment. It is expected that the depression ratios is practicable to dose evaluation for BNCT

  14. Radiation effects of high and low doses

    International Nuclear Information System (INIS)

    El-Naggar, A.M.

    1998-01-01

    The extensive proliferation of the uses and applications of atomic and nuclear energy resulted in possible repercussions on human health. The prominent features of the health hazards that may be incurred after exposure to high and low radiation doses are discussed. The physical and biological factors involved in the sequential development of radiation health effects and the different cellular responses to radiation injury are considered. The main criteria and features of radiation effects of high and low doses are comprehensively outlined

  15. Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit.

    Science.gov (United States)

    El-Jaby, Samy; Richardson, Richard B

    2015-07-01

    Occupational exposures from ionizing radiation are currently regulated for airline travel (Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit. Crown Copyright © 2015. Published by Elsevier Ltd. All rights reserved.

  16. The effect of the neutron spectra unfolding method on the fast neutron dose determination

    International Nuclear Information System (INIS)

    Marinkovic, P.; Zavaljevski, N.

    1992-01-01

    Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (orig.)

  17. The effect of the neutron spectra unfolding method on the fast neutron dose determination

    International Nuclear Information System (INIS)

    Marinkovic, P.; Avdic, S.; Pesic, M.; Zavaljevski, N

    1992-09-01

    Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (author)

  18. High-Energy Neutron Backgrounds for Underground Dark Matter Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yu [Syracuse Univ., NY (United States)

    2016-01-01

    Direct dark matter detection experiments usually have excellent capability to distinguish nuclear recoils, expected interactions with Weakly Interacting Massive Particle (WIMP) dark matter, and electronic recoils, so that they can efficiently reject background events such as gamma-rays and charged particles. However, both WIMPs and neutrons can induce nuclear recoils. Neutrons are then the most crucial background for direct dark matter detection. It is important to understand and account for all sources of neutron backgrounds when claiming a discovery of dark matter detection or reporting limits on the WIMP-nucleon cross section. One type of neutron background that is not well understood is the cosmogenic neutrons from muons interacting with the underground cavern rock and materials surrounding a dark matter detector. The Neutron Multiplicity Meter (NMM) is a water Cherenkov detector capable of measuring the cosmogenic neutron flux at the Soudan Underground Laboratory, which has an overburden of 2090 meters water equivalent. The NMM consists of two 2.2-tonne gadolinium-doped water tanks situated atop a 20-tonne lead target. It detects a high-energy (>~ 50 MeV) neutron via moderation and capture of the multiple secondary neutrons released when the former interacts in the lead target. The multiplicity of secondary neutrons for the high-energy neutron provides a benchmark for comparison to the current Monte Carlo predictions. Combining with the Monte Carlo simulation, the muon-induced high-energy neutron flux above 50 MeV is measured to be (1.3 ± 0.2) ~ 10-9 cm-2s-1, in reasonable agreement with the model prediction. The measured multiplicity spectrum agrees well with that of Monte Carlo simulation for multiplicity below 10, but shows an excess of approximately a factor of three over Monte Carlo prediction for multiplicities ~ 10 - 20. In an effort to reduce neutron backgrounds for the dark matter experiment SuperCDMS SNO- LAB, an active neutron veto was developed

  19. Contribution to the development of a primary standard for high energy neutron beams

    International Nuclear Information System (INIS)

    Mancaux, M.

    1983-12-01

    A tissue equivalent calorimeter, made of Shonka A-150 plastic, has been constructed in order to create a primary standard for high energy neutrons and to establish a calibration procedure for ionization chambers used in neutrontherapy. After a detailed description of the calorimeter and the associated measuring system, the preliminary tests are presented, in particular, the evolution of the response as a function of accumulated dose. The measurements of the total absorbed dose (n + γ) by calorimetry in a neutron beam, in order to determine chambers' calibration factors in terms of absorbed dose to A-150 plastic, have been performed at the Neutrontherapy Unit of the Centre Hospitalier Regional d'Orleans. The uncertainty in the determination of the total absorbed dose to the tissu equivalent material using the new procedure is 3% lower than that obtained with the usual procedure, derived from an exposure calibration [fr

  20. Effective dose evaluation for BNCT treatment in the epithermal neutron beam at THOR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.N. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Division of Health Physics, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Huang, C.K. [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Tsai, W.C. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, Y.H. [Nuclear Science and Technol. Develop. Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Jiang, S.H., E-mail: shjiang@mx.nthu.edu.tw [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

    2011-12-15

    This paper aims to evaluate the effective dose as well as equivalent doses of several organs of an adult hermaphrodite mathematical phantom according to the definition of ICRP Publication 60 for BNCT treatments of brain tumors in the epithermal neutron beam at THOR. The MCNP5 Monte Carlo code was used for the calculation of the average absorbed dose of each organ. The effective doses for a typical brain tumor treatment with a tumor treatment dose of 20 Gy-eq were evaluated to be 0.59 and 0.35 Sv for the LLAT and TOP irradiation geometries, respectively. In addition to the stochastic effect, it was found that it is also likely to produce deterministic effects, such as cataracts and depression of haematopoiesis.

  1. Neutron shielding properties of a borated high-density glass

    Directory of Open Access Journals (Sweden)

    Saeed Aly Abdallah

    2017-01-01

    Full Text Available The neutron shielding properties of a borated high density glass system was characterized experimentally. The total removal macroscopic cross-section of fast neutrons, slow neutrons as well as the linear attenuation coefficient of total gamma rays, primary in addition to secondary, were measured experimentally under good geometric condition to characterize the attenuation properties of (75-x B2O3-1Li2O-5MgO-5ZnO-14Na2O-xBaO glassy system. Slabs of different thicknesses from the investigated glass system were exposed to a collimated beam of neutrons emitted from 252Cf and 241Am-Be neutron sources in order to measure the attenuation properties of fast and slow neutrons as well as total gamma rays. Results confirmed that barium borate glass was suitable for practical use in the field of radiation shielding.

  2. Workplace characterization using spectrometry for reliable neutron dose assessment in mixed neutron photon fields

    International Nuclear Information System (INIS)

    Bolognese-Milsztajn, T.; Asselineau, B.; Itie, C.; Lahaye, T.; Muller, H.

    2002-01-01

    Radiation protection for practices should be based on justification of exposure, optimization of protection and dose limitations. The 96/29 European Union directive clearly requires that operational protection of exposed workers shall be based on prior evaluation to identify the nature and magnitude of the radiological risk to exposed workers and implementation of optimization of radiation protection in all working conditions. Operational protection of exposed workers requires the implementation of measures at workplaces. The ICRP60 publication (1991) defines and recommends the use of the effective dose E for radiological protections purposes. This quantity can not be directly measured; the operational quantities ambient dose equivalent: H and personal dose equivalent: Hp as defined in the ICRU 51 report (1993) are recommended to estimate the effective dose E. These operational quantities are related, through calculated conversion coefficients h Φ , to the measurable function fluence Φ by: H = ∫ h Φ (E) Φ E dE (1a); Hp = ∫ hp Φ (E, Ω) Φ E , Ω dE dΩ (1b). Φ E (Φ E ,Ω) is the distribution of the fluence with respect to energy (energy E and direction Ω); h Φ or hp Φ are the fluence to operational quantity conversion coefficients as defined by the ICRU 57 report (1998)

  3. Improved Dose Targeting for a Clinical Epithermal Neutron Capture Beam Using Optional 6Li Filtration

    International Nuclear Information System (INIS)

    Binns, Peter J.; Riley, Kent J.; Ostrovsky, Yakov; Gao Wei; Albritton, J. Raymond; Kiger, W.S.; Harling, Otto K.

    2007-01-01

    Purpose: The aim of this study was to construct a 6 Li filter and to improve penetration of thermal neutrons produced by the fission converter-based epithermal neutron beam (FCB) for brain irradiation during boron neutron capture therapy (BNCT). Methods and Materials: Design of the 6 Li filter was evaluated using Monte Carlo simulations of the existing beam line and radiation transport through an ellipsoidal water phantom. Changes in beam performance were determined using three figures of merit: (1) advantage depth (AD), the depth at which the total biologically weighted dose to tumor equals the maximum weighted dose to normal tissue; (2) advantage ratio (AR), the ratio of the integral tumor dose to that of normal tissue averaged from the surface to the AD; and (3) advantage depth dose rate (ADDR), the therapeutic dose rate at the AD. Dosimetry performed with the new filter installed provided calibration data for treatment planning. Past treatment plans were recalculated to illustrate the clinical potential of the filter. Results: The 8-mm-thick Li filter is more effective for smaller field sizes, increasing the AD from 9.3 to 9.9 cm, leaving the AR unchanged at 5.7 but decreasing the ADDR from 114 to 55 cGy min -1 for the 12 cm diameter aperture. Using the filter increases the minimum deliverable dose to deep seated tumors by up to 9% for the same maximum dose to normal tissue. Conclusions: Optional 6 Li filtration provides an incremental improvement in clinical beam performance of the FCB that could help to establish a therapeutic window in the future treatment of deep-seated tumors

  4. A Monte Carlo Study on the Effect of Various Neutron Capturers on Dose Distribution in Brachytherapy with 252Cf Source

    Directory of Open Access Journals (Sweden)

    Firoozabadi M. M.

    2017-03-01

    Full Text Available Background: In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials. Objective: The purpose of this study is to evaluate dose distribution in the presence of 10B, 157Gd and 33S neutron capturers and to determine the effect of these materials on dose enhancement rate for 252Cf brachytherapy source. Methods: Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF are determined in the absence and presence of 10B, 157Gd and 33S using Monte Carlo simulation. Results: The difference in the thermal neutron flux rate in the presence of 10B and 157Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for 157Gd and 10B, respectively at about 8 cm distance from the source center. DEF for 33S is equal to one. Conclusion: Results show that the magnitude of dose augmentation in tumors containing 10B and 157Gd in brachytherapy with 252Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. 33S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the 33S concentration in tumor and tumor distance from the source.

  5. Non-dispersive method for measuring longitudinal neutron coherence length using high frequency cold neutron pulser

    International Nuclear Information System (INIS)

    Kawai, T.; Tasaki, S.; Ebisawa, T.; Hino, M.; Yamazaki, D.; Achiwa, N.

    1999-01-01

    Complete text of publication follows. A non-dispersive method is proposed for measuring the longitudinal coherence length of a neutron using a high frequency cold neutron pulser (hf-CNP) placed between two multilayer spin splitters (MSS) which composes the cold neutron spin interferometer. Two spin eigenstates of a neutron polarized x-y plane are split non-dispersively and longitudinally in time by the hf-CNP which could reflect two components alternatively in time. The reduction of the visibility of interference fringes after being superposed by the second MSS is measured as a function of the frequency of the pulser by TOF method. From the zero visibility point obtained by extrapolation one could obtain the longitudinal coherence length of the neutron. (author)

  6. The Primary Origin of Dose Rate Effects on Microstructural Evolution of Austenitic Alloys During Neutron Irradiation

    International Nuclear Information System (INIS)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.

    2002-01-01

    The effect of dose rate on neutron-induced microstructural evolution was experimentally estimated. Solution-annealed austenitic model alloys were irradiated at approximately 400 degrees C with fast neutrons at seven different dose rates that vary more than two orders difference in magnitude, and two different doses were achieved at each dose rate. Both cavity nucleation and growth were found to be enhanced at lower dose rate. The net vacancy flux is calculated from the growth rate of cavities that had already nucleated during the first cycle of irradiation and grown during the second cycle. The net vacancy flux was found to be proportional to (dpa/sec) exp (1/2) up to 28.8 dpa and 8.4 x 10 exp (-7) dpa/sec. This implies that mutual recombination dominates point defect annihilation, in this experiment even though point defect sinks such as cavities and dislocations were well developed. Thus, mutual recombination is thought to be the primary origin of the effect of dose rate on microstructural evolution

  7. Development of highly effective neutron shielding material made of phenol-novolac type epoxy resin

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Soo Haeng; Jeong, Myeong Soo; Hong, Sun Seok; Lee, Won Kyoung; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy; Oh, Seok Jin

    1998-06-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. Especially we developed phenol-novolac type epoxy resin based neutron shielding materials and their characteristics were also evaluated. (author). 22 refs., 11 tabs., 21 figs

  8. A neutron calibration technique for detectors with low neutron/high photon sensitivity

    International Nuclear Information System (INIS)

    Jahr, R.; Guldbakke, S.; Cosack, M.; Dietze, G.; Klein, H.

    1978-03-01

    The neutron response of a detector with low neutron-/high photon sensitivity is given by the difference of two terms: the response to the mixed neutron-photon field, measured directly, and the response to the photons, deduced from additional measurements with a photon spectrometer. The technique is particularly suited for use in connection with targets which consist of a thick backing and thin layer of neutron producing material such as T, D, Li nuclei. Then the photon component of the mixed field is very nearly the same as the pure photon field from a 'phantom target', being identical with the neutron producing target except for the missing neutron producing material. Using this technique in connection with a T target (Ti-T-layer on silver backing) and the corresponding phantom target (Ti-layer on silver backing), a GM counter was calibrated at a neutron energy of 2.5 MeV. Possibilities are discussed to subsequently calibrate the GM counter at other neutron energies without the use of the photon spectrometer. (orig./HP) [de

  9. Development of high efficiency neutron detectors

    International Nuclear Information System (INIS)

    Pickrell, M.M.; Menlove, H.O.

    1993-01-01

    The authors have designed a novel neutron detector system using conventional 3 He detector tubes and composites of polyethylene and graphite. At this time the design consists entirely of MCNP simulations of different detector configurations and materials. These detectors are applicable to low-level passive and active neutron assay systems such as the passive add-a-source and the 252 Cf shuffler. Monte Carlo simulations of these neutron detector designs achieved efficiencies of over 35% for assay chambers that can accommodate 55-gal. drums. Only slight increases in the number of detector tubes and helium pressure are required. The detectors also have reduced die-away times. Potential applications are coincident and multiplicity neutron counting for waste disposal and safeguards. The authors will present the general design philosophy, underlying physics, calculation mechanics, and results

  10. Radiation protection metrology at a high-energy neutron therapy facility

    International Nuclear Information System (INIS)

    Bonnett, D.E.; Sherwin, A.G.; More, B.R.

    1991-01-01

    A radiation protection survey has been carried out at a high-energy neutron therapy facility using a combination of different detectors and counters. Included in the survey were measurements with a tissue equivalent proportional counter (TEPC), a rem meter, a large volume ionisation chamber (LVI) and a Geiger counter. Dose equivalent rates, normalised to a proton beam current of 25 μA, of between 1 μSv.h -1 and 0.7 Sv.h -1 were recorded depending on the location. In general the results confirm the tendency of the rem meter to over-read in fields consisting mainly of low energy neutrons and illustrate the advantages of the diagnostic and gamma discriminating properties of the TEPC. The LVI-Geiger system was found to be the least favourable combination of dosemeters, substantially under-reading and being unable to estimate the neutron dose rate at levels below about 32 μGy.h -1 . (author)

  11. Thermal neutron dose calculations in a brain phantom from 7Li(p,n) reaction based BNCT setup

    International Nuclear Information System (INIS)

    Elshahat, B.A.; Naqvi, A.A.; Maalej, N.; Abdallah, Khalid

    2006-01-01

    Monte Carlo simulations were carried out to calculate neutron dose in a brain phantom from a 7 Li(p,n) reaction based setup utilizing a high density polyethylene moderator with graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal /(fast +thermal) neutron intensity ratio as a function of geometric parameters of the setup. Results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated Peak Therapeutic Ratio for the setup was found to be 2.15. With further improvement in the moderator design and brain phantom irradiation arrangement, the setup capabilities can be improved to reach further deep-seated tumor. (author)

  12. Effect of low level doses of fast neutrons on the activity of the snake venom

    International Nuclear Information System (INIS)

    Hanafy, Magda S.; Amin, Aida M.

    1998-01-01

    In this work, the effect of low level doses of fast neutrons from 252 Cf on snake venom (Cerastes cerastes) was studied through measurements of biophysical and haematological changes. The absorption spectrum (200-700 nm) of haemoglobin (Hb) collected from the blood of rats after 3 and 24 hours post injection with irradiated and non-irradiated snake venom with neutron fluences of 3x10 6 , 2.8x10 7 and 3X10 8 n/cm 2 was measured. The results indicated that injection of animals with either non- irradiated or irradiated venom ( with different neutron fluences) resulted into the decrease of the absorption band intensities of Hb. These changes in the properties of the characteristic band showed to be a marker for irradiated venom and is dose dependent. It was concluded that neutron irradiation of the venom leads to the decrease of its toxicity and, consequently, to the increase of the chance of repair mechanism in livings. Obvious changes of most haematological erythrocytic values of Hb, packed cell volume (PCV), red blood counts (RBC), mean corpuscular volume (MCV), mean corpuscular haemoglobin (MCHb) and mean corpuscular haemoglobin concentration (MCHC) were observed in the blood of the rats injected with non-irradiated venom (as a first group) and those injected with the irradiated venom (as a second group). The microcytic haemolytic anemia was more acute in the first group than in the second one which showed lesser extent. It is concluded from this study that low level doses of fast neutrons could postpone and lower acute haematological action induced by the venom. (authors)

  13. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  14. Dose profile modeling of Idaho National Laboratory's active neutron interrogation laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, D.L. [Idaho National Laboratory, 2525 N. Fremont Avenue, Idaho Falls, ID 83415 (United States)], E-mail: david.chichester@inl.gov; Seabury, E.H.; Zabriskie, J.M.; Wharton, J.; Caffrey, A.J. [Idaho National Laboratory, 2525 N. Fremont Avenue, Idaho Falls, ID 83415 (United States)

    2009-06-15

    A new laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for deuterium-tritium (DT) fusion (14.1 MeV) neutron generators (2x10{sup 8} n/s), deuterium-deuterium (DD) fusion (2.5 MeV) neutron generators (1x10{sup 7} n/s), and {sup 252}Cf spontaneous fission neutron sources (6.96x10{sup 7} n/s, 30 {mu}g). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for {sup 252}Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield walls and entrance mazes and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.

  15. Incidence of leukemia among atomic bomb survivors in relation to neutron and gamma dose, Hiroshima and Nagasaki, 1950-71

    International Nuclear Information System (INIS)

    Ishimaru, Toranosuke; Otake, Masanori; Ichimaru, Michito.

    1978-03-01

    The incidence of leukemia during 1950-71 in the fixed mortality sample of atomic bomb survivors in Hiroshima and Nagasaki has been analyzed as a function of individual gamma and neutron kerma and marrow dose. Two dose response models were tested for each of acute leukemia, chronic granulocytic leukemia, and all types of leukemia, respectively. Each model postulates that leukemia incidence depends upon the sum of the separate risks imposed by the gamma ray and neutron doses; in Model I both are assumed to be directly proportional to the respective doses, while Model II assumes that while the risk from neutrons is directly proportional to the dose, the risk from gamma rays is proportional to dose-squared. Weighted regression analyses were performed for each model. When the two models were fitted to the data for all types of leukemia, the estimated regression coefficients corresponding to the neutron and gamma ray doses both differed significantly from zero, for each model. However, when analysis was restricted to acute leukemia, both the neutron and gamma ray coefficients were significant only for Model II, and with respect to chronic granulocytic leukemia, only the coefficient of the neutron dose was significant, using either Model I or Model II. It appeared that the responses of the two leukemia types differed by type of radiation. If the chronic granulocytic and acute leukemias are considered together, the Model II appears to fit the data slightly better than Model I, but neither models is rejected by the data. (author)

  16. Position sensitive detection of neutrons in high radiation background field.

    Science.gov (United States)

    Vavrik, D; Jakubek, J; Pospisil, S; Vacik, J

    2014-01-01

    We present the development of a high-resolution position sensitive device for detection of slow neutrons in the environment of extremely high γ and e(-) radiation background. We make use of a planar silicon pixelated (pixel size: 55 × 55 μm(2)) spectroscopic Timepix detector adapted for neutron detection utilizing very thin (10)B converter placed onto detector surface. We demonstrate that electromagnetic radiation background can be discriminated from the neutron signal utilizing the fact that each particle type produces characteristic ionization tracks in the pixelated detector. Particular tracks can be distinguished by their 2D shape (in the detector plane) and spectroscopic response using single event analysis. A Cd sheet served as thermal neutron stopper as well as intensive source of gamma rays and energetic electrons. Highly efficient discrimination was successful even at very low neutron to electromagnetic background ratio about 10(-4).

  17. Ambient neutron dose equivalent outside concrete vault rooms for 15 and 18 MV radiotherapy accelerators

    International Nuclear Information System (INIS)

    Martinez-ovalle, S. A.; Barquero, R.; Gomez-ros, J. M.; Lallena, A. M.

    2012-01-01

    In this work, the ambient dose equivalent, H*(10), due to neutrons outside three bunkers that house a 15- and a 18-MV Varian Clinac 2100C/D and a 15-MV Elekta Inor clinical linacs, has been calculated. The Monte Carlo code MCNPX (v. 2.5) has been used to simulate the neutron production and transport. The complete geometries including linacs and full installations have been built up according to the specifications of the manufacturers and the planes provided by the corresponding medical physical services of the hospitals where the three linacs operate. Two of these installations, those lodging the Varian linacs, have an entrance door to the bunker while the other one does not, although it has a maze with two bends. Various treatment orientations were simulated in order to establish plausible annual equivalent doses. Specifically anterior-posterior, posterior-anterior, left lateral, right lateral orientations and an additional one with the gantry rotated 30 deg. have been studied. Significant dose rates have been found only behind the walls and the door of the bunker, near the entrance and the console, with a maximum of 12 μSv h -1 . Dose rates per year have been calculated assuming a conservative workload for the three facilities. The higher dose rates in the corresponding control areas were 799 μSv y -1 , in the case of the facility which operates the 15-MV Clinac, 159 μSv y -1 , for that with the 15-MV Elekta, and 21 μSv y -1 for the facility housing the 18-MV Varian. A comparison with measurements performed in similar installations has been carried out and a reasonable agreement has been found. The results obtained indicate that the neutron contamination does not increase the doses above the legal limits and does not produce a significant enhancement of the dose equivalent calculated. When doses are below the detection limits provided by the measuring devices available today, MCNPX simulation provides an useful method to evaluate neutron dose equivalents

  18. High-spin structure of neutron-rich Dy isotopes

    Indian Academy of Sciences (India)

    Neutron-rich Dy isotopes; high-spin states; g-factors; cranked HFB theory. ... for 164Dy marking a clear separation in the behaviour as a function of neutron ... cipal x-axis as the cranking axis) in this mass region we have planned to make a sys-.

  19. A high pressure sample facility for neutron scattering

    International Nuclear Information System (INIS)

    Carlile, C.J.; Glossop, B.H.

    1981-06-01

    Commissioning tests involving deformation studies and tests to destruction as well as neutron diffraction measurements of a standard sample have been carried out on the SERC high pressure sample facility for neutron scattering studies. A detailed description of the pressurising equipment is given. (author)

  20. Survival of tumor bearing mice by sequencing of low dose rate (LDR) neutron and photon radiation

    International Nuclear Information System (INIS)

    Onomura, C.I.; Feola, J.M.; Maruyama, Y.

    1984-01-01

    Cf-252 neutron radiation (NT) has been shown to be effective therapy for bulky, hypoxic human tumor and to produce consistent rapid clearance and 5 year cures. NT has been found to be more or less effective depending upon the schedule in which it is used and upon mixing with photon radiation. In an effort to study this scheduling and photon effect, LSA tumor was irradiated in vivo in a hypoxic, advanced state, in different schedules in combination of NT with Co-60 photons. The LSA lymphoma of C57BL/ym mice represents an accurate system to assess dose-response of tumor cells in vivo. Mean survival time was used as endpoint. A high RBE for LDR Cf-252 NT was observed with a RBE(n) of -- 5.0. The effect was not greatly sensitive to sequence in which photons were used. Comparison studies were also tested relative to LDR Cs-137 photon radiation. The results support the high efficacy of LDR NT for destruction of hypoxic tumor in vivo

  1. Angular dependence of dose equivalent response of an albedo neutron dosimeter

    International Nuclear Information System (INIS)

    Torres, B.A.; Boswell, E.; Schwartz, R.B.

    1994-01-01

    The ANSI provides procedures for testing the performance of dosimetry services. Although neutron dose equivalent angular response studies are not now mandated, future standards may well require that such studies be performed. Current studies with an albedo dosimeter will yield information regarding the angular dependence of dose equivalent response for this type of personnel dosimeter. Preliminary data for bare 252 Cf fluences show a marked decrease in dosimeter reading with increasing angle. The response decreased by an approximate factor of four. For the horizontal orientation, the same response was noted from both positive and negative angles. However, for the vertical orientation, the response was unexplainably assymetric. We are also examining the response of the personnel badge in moderated 252 Cf fluences. Responses from the moderated and unmoderated 252 Cf fields and theoretical calculations of the neutron angular response will be compared. This information will assist in building a data base for future comparisons of neutron angular responses with other neutron albedo dosimeters and phantoms

  2. Ionization chamber for high dose measurements

    International Nuclear Information System (INIS)

    Rodrigues Junior, Ary de Araujo

    2005-01-01

    Industrial gamma irradiators facilities are designed for processing large amounts of products, which are exposed to large doses of gamma radiation. The irradiation, in industrial scale, is usually carried out in a dynamic form, where the products go through a 60 Co gamma source with activity of TBq to P Bq (k Ci to MCi). The dose is estimated as being directly proportional to the time that the products spend to go through the source. However, in some situations, mainly for research purposes or for validation of customer process following the ISO 11137 requirements, it is required to irradiate small samples in a static position with fractional deliver doses. The samples are put inside the irradiation room at a fixed distance from the source and the dose is usually determined using dosimeters. The dose is only known after the irradiation, by reading the dosimeter. Nevertheless, in the industrial irradiators, usually different kinds of products with different densities go through between the source and the static position samples. So, the dose rate varies in function of the product density. A suitable methodology would be to monitor the samples dose in real time, measuring the dose on line with a radiation detector, which would improve the dose accuracy and avoid the overdose. A cylindrical ionization chamber of 0.9 cm 3 has been developed for high-doses real-time monitoring, during the sample irradiation at a static position in a 60 Co gamma industrial plant. Nitrogen and argon gas at pressure of 10 exp 5 Pa (1 bar) was utilized to fill the ionization chamber, for which an appropriate configuration was determined to be used as a detector for high-dose measurements. To transmit the signal generated in the ionization chamber to the associated electronic and processing unit, a 20 m mineral insulated cable was welded to the ionization chamber. The signal to noise ratio produced by the detector was about 100. The dosimeter system was tested at a category I gamma

  3. Applicability of thermoluminescent dosimeters in X-ray organ dose determination and in the dosimetry of systemic and boron neutron capture radiotherapy

    International Nuclear Information System (INIS)

    Aschan, C.

    1999-01-01

    The main detectors used for clinical dosimetry are ionisation chambers and semiconductors. Thermoluminescent (TL) dosimeters are also of interest because of their following advantages: (i) wide useful dose range, (ii) small physical size, (iii) no need for high voltage or cables, i.e. stand alone character, and (iv) tissue equivalence (LiF) for most radiation types. TL detectors can particularly be used for the absorbed dose measurements performed with the aim to investigate cases where dose prediction is difficult and not as part of a routine verification procedure. In this thesis, the applicability of TL detectors was studied in different clinical applications. Particularly, the major phenomena (e.g. energy dependence, sensitivity to high LET radiation, reproducibility) affecting on the precision and accuracy of TL detectors in the dose estimations were considered in this work. In organ dose determinations of diagnostic X-ray examinations, the TL detectors were found to be accurate within 5% (1 S.D.). For in viva studies using internal irradiation source, i.e. for systemic radiation therapy, a method for determining the absorbed doses to organs was introduced. The TL method developed was found to be able to estimate the absorbed doses to those critical organs near the body surface within 50%. In the mixed neutron-gamma field of boron neutron capture therapy (BNCT), TL detectors were used for gamma dose and neutron fluence measurements. They were found able to measure the neutron dose component with the accuracy of 16%, and therefore to be a useful addition to the activation foils in BNCT neutron dosimetry. The absorbed gamma doses can be measured with TL detectors within 20% in the mixed neutron-gamma field, which enables in viva measurements at BNCT beams with approximately the same accuracy. In this study, the uncertainties of TL dosimeters were found to be high but not essentially greater than those in other measurement techniques used for clinical dosimetry

  4. Application of laboratory sourceless object counting for the estimation of the neutron dose

    International Nuclear Information System (INIS)

    Cheng Jie; Ning Jing; Zhang Xiaomin; Qu Decheng; Xie Xiangdong; Nan Hongjie

    2011-01-01

    Objective: To estimate the neutron dose using 24 Na energy spectrum analysis method. Methods: Genius-2000 GeomComposer software package was used to calibrate the efficiency of the detector. Results: The detection efficiency of the detector toward γ photon with an energy of 1.368 MeV was quickly found to be 4.05271×10 -3 while the error of the software was 4.0% . The estimated dose value of the neutron irradiation samples was between 1.94 Gy and 2.82 Gy, with an arithmetic mean value of 2.38 Gy. The uncertainty of the dosimetry was about 20.07% . Conclusion: The application of efficiency calibration without a radioactive source of the energy spectrum analysis of the 24 Na contained in human blood with accelerate the estimation process. (authors)

  5. Stimulation growth effect of Eriocheir sinensis treated with low-dose neutron

    International Nuclear Information System (INIS)

    Luo Keyong; Liu Chunquan; Xu Lixin; Peng Zhangji

    2006-01-01

    This paper was dealt with the relationship between biochemical indexes and different growth stages of Eriocheir sinensis megalopa treated with Low-dose Neutron at 55.24 to 73.66 mGy. It showed that some biochemical component indexes were increased, such as-SH group in protain (between 23.40% to 69.59%), albumen (between 4.99% to 22.6%) and Hyp compared with CK. However, free radical level (between 7.67% to 32.68%) and AKP were decreased. The carapace color was turned into darker than that of CK; Antibacterial immunity of younger crab during the growing stage was increased, the body size of treated Eriocheir sinensis megalopa became uniform and early sexual maturity was inhibited in a certain degree with a low dose neutron treatment. (authors)

  6. Measurement of two-dimensional thermal neutron flux in a water phantom and evaluation of dose distribution characteristics

    International Nuclear Information System (INIS)

    Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Horiguchi, Yoji

    2001-03-01

    To evaluate nitrogen dose, boron dose and gamma-ray dose occurred by neutron capture reaction of the hydrogen at the medical irradiation, two-dimensional distribution of the thermal neutron flux is very important because these doses are proportional to the thermal neutron distribution. This report describes the measurement of the two-dimensional thermal neutron distribution in a head water phantom by neutron beams of the JRR-4 and evaluation of the dose distribution characteristic. Thermal neutron flux in the phantom was measured by gold wire placed in the spokewise of every 30 degrees in order to avoid the interaction. Distribution of the thermal neutron flux was also calculated using two-dimensional Lagrange's interpolation program (radius, angle direction) developed this time. As a result of the analysis, it was confirmed to become distorted distribution which has annular peak at outside of the void, though improved dose profile of the deep direction was confirmed in the case which the radiation field in the phantom contains void. (author)

  7. Spectra and neutron dose of an 18 MV Linac using two geometric models of the head

    International Nuclear Information System (INIS)

    Barrera, M. T.; Pino, F.; Barros, H.; Sajo-Bohus, L.; Davila, J.; Salcedo, E.; Vega C, H. R.; Benites R, J. L.

    2015-10-01

    Full text: Using the Monte Carlo method, by MCNP5 code, simulations were performed with different source terms and 2 geometric models of the head to obtain spectra in energy, flow and doses of photo-neutrons at different positions on the stretcher and in the radiotherapy room. The simplest model was a spherical shell of tungsten; the second was the complete model of a heterogeneous head of an accelerator Varian ix. In both models Tosi function was used as a source term. In addition, for the second model Sheikh-Bagheri distribution was used for photons and photo-neutrons were generated. Also in both models the radiotherapy room of Gurve group of the Teaching Medical Center La Trinidad was included, which is equipped with an accelerator Varian Clinic 2100. In this Center passive detectors PADC (Cr-39) were irradiated with neutron converters, with 18 MeV photons radiation. The measured neutron flow was compared with that obtained with Monte Carlo calculations. The Monte Carlo flows are similar to those measured at the isocenter. The simplest model underestimates the neutron flow compared with the calculated flows with the heterogeneous model of the head. (Author)

  8. A study of the responses of neutron dose equivalent survey meters with computer codes

    International Nuclear Information System (INIS)

    Sartori, D.E.; Beer, G.P. de

    1983-01-01

    The ANISN and DOT discrete-ordinates radiation transport codes for one and two dimensions have been proved as effective and simple techniques to study the response of dose equivalent neutron detectors. Comparisons between results of an experimental calibration of the Harwell 95/0075 survey meter and calculated results rendered satisfactory agreement, considering the different techniques and sources of error involved. Possible improvements in the methods and designs and causes of error are discussed. (author)

  9. Intracranial meningiomas after high-dose irradiation

    International Nuclear Information System (INIS)

    Soffer, D.; Gomori, J.M.; Siegal, T.; Shalit, M.N.

    1989-01-01

    Three patients who presented with intracranial meningiomas 12, 15, and 20 years, respectively, after therapeutic high-dose irradiation of a primary brain tumor are described. Analysis of these cases and similar documented cases suggests that meningiomas after high-dose irradiation constitute a recognizable entity. Patients with such tumors received radiation therapy at a young age (mean age, 9.4 years). After a latent period of 2 to 47 years (mean, 19.8 years) they developed meningiomas at the site of irradiation, at a much younger age than patients with ''spontaneous'' meningiomas. Similar to the situation with meningiomas after low-dose irradiation, a relatively high proportion of meningiomas induced by high-dose irradiation tend to be malignant and biologically aggressive. A very young age at the time of irradiation seems to predispose to the induction of malignant meningiomas, rather than benign tumors. These unusual features provide indirect evidence that high-dose radiation may play a role in the pathogenesis of meningiomas.41 references

  10. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    International Nuclear Information System (INIS)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R.

    2008-01-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  11. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    Energy Technology Data Exchange (ETDEWEB)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R. [McMaster Univ., Medical Physics and Applied Radiation Sciences Unit, Hamilton, Ontario (Canada)

    2008-07-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  12. Development of a phoswich detector for neutron dose rate measurements in the Earth's atmosphere

    International Nuclear Information System (INIS)

    Doensdorf, Esther Miriam

    2014-01-01

    The Earth is constantly exposed to a stream of energetic particles from outer space. Through the interaction of this radiation with the Earth's magnetosphere and atmosphere a complex radiation field is formed which varies with the location inside the Earth's atmosphere. This radiation field consists of charged and uncharged particles leading to the constant exposure of human beings to radiation. As this ionizing radiation can be harmful for humans, it is necessary to perform dose rate measurements in different altitudes in the Earth's atmosphere. Due to their higher biological effectiveness the exposure to neutrons is more harmful than the exposure to γ-rays and charged particles, which is why the determination of neutron dose rates is the focus of this work. In this work the prototype of a Phoswich detector called PING (Phoswich Instrument for Neutrons and Gammas) is developed to determine dose rates caused by neutrons in the Earth's atmosphere and to distinguish these from γ-rays. The instrument is composed of two different scintillators optically coupled to each other and read out by one common photomultiplier tube. The scintillator package consists of an inner plastic scintillator made of the material BC-412 and a surrounding anti-coincidence made of sodium doped caesium iodide (CsI(Na)). In this work the instrument is calibrated, tested and flown and a procedure for a pulse shape analysis for this instrument is developed. With this analysis it is possible to distinguish pulses from the plastic scintillator and pulses from the CsI(Na). The pulses from the plastic scintillator are mainly due to the interaction of neutrons but there is an energy-dependent contribution of γ-rays to these events. Measurements performed on board an airplane show that the dose rates measured with the developed detector are in the same order of magnitude as results of other instruments. During measurements on board stratospheric balloons the altitude dependence of count rates and

  13. 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)

  14. Correlation of clinical outcome to the estimated radiation dose from Boron Neutron Capture Therapy (BNCT)

    Energy Technology Data Exchange (ETDEWEB)

    Chadha, M. [Beth Israel Medical Center, NY (United States). Dept. of Radiation Oncology; Coderre, J.A.; Chanana, A.D. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1996-12-31

    A phase I/II trial delivering a single fraction of BNCT using p-Boronophenylalanine-Fructose and epithermal neutrons at the the Brookhaven Medical Research Reactor was initiated in September 1994. The primary endpiont of the study was to evaluate the feasibility and safety of a given BNCT dose. The clinical outcome of the disease was a secondary endpoint of the study. The objective of this paper is to evaluate the correlation of the clinical outcome of patients to the estimated radiation dose from BNCT.

  15. Correlation of clinical outcome to the estimated radiation dose from Boron Neutron Capture Therapy (BNCT)

    International Nuclear Information System (INIS)

    Chadha, M.

    1996-01-01

    A phase I/II trial delivering a single fraction of BNCT using p-Boronophenylalanine-Fructose and epithermal neutrons at the the Brookhaven Medical Research Reactor was initiated in September 1994. The primary endpiont of the study was to evaluate the feasibility and safety of a given BNCT dose. The clinical outcome of the disease was a secondary endpoint of the study. The objective of this paper is to evaluate the correlation of the clinical outcome of patients to the estimated radiation dose from BNCT

  16. The application of computer and automatic technology in dose measurement of neutron radiation

    International Nuclear Information System (INIS)

    Zhou Yu; Li Chenglin; Luo Yisheng; Guo Yong; Chen Di; Xiaojiang

    1999-01-01

    Generally the dose measurement of neutron radiation requires three electrometers, two bias, three workers in the same time. To improve the accuracy and efficiency of measurement, a Model 6517A electrometer that accommodate Model 6521 scanner cards and a portable computer are used to make up of a automatic measurement system. Corresponding software is developed and used to control it. Because of the application of computer and automatic technology, this system can not only measure dose rate automatically, but also make data's calculating, saving, querying, printing and comparing ease

  17. High Brightness Neutron Source for Radiography. Final report

    International Nuclear Information System (INIS)

    Cremer, J.T.; Piestrup, Melvin A.; Gary, Charles K.; Harris, Jack L.; Williams, David J.; Jones, Glenn E.; Vainionpaa, J.H.; Fuller, Michael J.; Rothbart, George H.; Kwan, J.W.; Ludewigt, B.A.; Gough, R.A.; Reijonen, Jani; Leung, Ka-Ngo

    2008-01-01

    This research and development program was designed to improve nondestructive evaluation of large mechanical objects by providing both fast and thermal neutron sources for radiography. Neutron radiography permits inspection inside objects that x-rays cannot penetrate and permits imaging of corrosion and cracks in low-density materials. Discovering of fatigue cracks and corrosion in piping without the necessity of insulation removal is possible. Neutron radiography sources can provide for the nondestructive testing interests of commercial and military aircraft, public utilities and petrochemical organizations. Three neutron prototype neutron generators were designed and fabricated based on original research done at the Lawrence Berkeley National Laboratory (LBNL). The research and development of these generators was successfully continued by LBNL and Adelphi Technology Inc. under this STTR. The original design goals of high neutron yield and generator robustness have been achieved, using new technology developed under this grant. In one prototype generator, the fast neutron yield and brightness was roughly 10 times larger than previously marketed neutron generators using the same deuterium-deuterium reaction. In another generator, we integrate a moderator with a fast neutron source, resulting in a high brightness thermal neutron generator. The moderator acts as both conventional moderator and mechanical and electrical support structure for the generator and effectively mimics a nuclear reactor. In addition to the new prototype generators, an entirely new plasma ion source for neutron production was developed. First developed by LBNL, this source uses a spiral antenna to more efficiently couple the RF radiation into the plasma, reducing the required gas pressure so that the generator head can be completely sealed, permitting the possible use of tritium gas. This also permits the generator to use the deuterium-tritium reaction to produce 14-MeV neutrons with increases

  18. The Biological Effect of Fast Neutrons and High-Energy Protons

    International Nuclear Information System (INIS)

    Moskalev, Ju.I.; Petrovich, I.K.; Strel'cova, V.N.

    1964-01-01

    The paper gives the results of comparative experiments on the effects of fast neutrons and high-energy protons (500 MeV) on life expectancy, peripheral blood, incidence and rate of appearance of tumours in the rat as a function of administered dose and time of observation. The neutron experiment was performed on 573 and the proton experiment on 490 white rats. The animals irradiated with fast neutrons were given doses between 8.5 and 510 rad, and those irradiated with protons received doses between 28 and 1008 rad. The effective doses for the acute, sub-acute and chronic forms of sickness were established for fast neutrons and for protons. LD 50/30 for neutrons was 408 and for protons 600 rad, and the corresponding LD 50 / 120 values were 380 and 600 rad. The conditions governing rat mortality were analysed both in the early and the later stages of the experiment. It is shown that the average life expectancy of rats irradiated with fast neutrons does not depend on sex. The shape of the dose-effect curve for the various peripheral-blood indexes is strongly dependent not only on the radiosensitivity of the blood cells in question but also on the time of observation. It may change greatly in time for one and the same index. A considerable time after irradiation with either fast neutrons or protons, benign and malignant tumours appear in different tissues of the rats, including the haemopoeitic tissues, mammary glands, pituitary, uterus, ovaries, prostate gland, testicles, liver, kidneys, lungs, gastro-intestinal tract, subcutaneous tissue, lymph nodes, urinary bladder, etc. The over-all incidence of tumours and the number of cases of multi centred neoplasms in females are two to three times higher than in males. The minimum tumour dose for the mammary glands with neutron irradiation is apparently rather less than 42.5 rad. The maximum incidence of tumours of the pituitary is found after irradiation with a dose of 42.5 rad.- At this same dose leucosis and tumour of the

  19. Standardization of high-dose measurement of electron and gamma ray absorbed doses and dose rates

    International Nuclear Information System (INIS)

    McLaughlin, W.L.

    1985-01-01

    Intense electron beams and gamma radiation fields are used for sterilizing medical devices, treating municipal wastes, processing industrial goods, controlling parasites and pathogens, and extending the shelf-life of foods. Quality control of such radiation processes depends largely on maintaining measurement quality assurance through sound dosimetry procedures in the research leading to each process, in the commissioning of that process, and in the routine dose monitoring practices. This affords documentation as to whether satisfactory dose uniformity is maintained throughout the product and throughout the process. Therefore, dosimetry at high doses and dose rates must in many radiation processes be standardized carefully, so that 'dosimetry release' of a product is verified. This standardization is initiated through preliminary dosimetry intercomparison studies such as those sponsored recently by the IAEA. This is followed by establishing periodic exercises in traceability to national or international standards of absorbed dose and dose rate. Traceability is achieved by careful selection of dosimetry methods and proven reference dosimeters capable of giving sufficiently accurate and precise 'transfer' dose assessments: (1) they must be calibrated or have well-established radiation-yield indices; (2) their radiation response characteristics must be reproducible and cover the dose range of interest; (3) they must withstand the rigours of back-and-forth mailing between a central standardizing laboratory and radiation processing facilities, without excessive errors arising due to instabilities, dosimeter batch non-uniformities, and environmental and handling stresses. (author)

  20. Quantitative radiation dose-response relationships for normal tissues in man - I. Gustatory tissues response during photon and neutron radiotherapy

    International Nuclear Information System (INIS)

    Mossman, K.L.

    1982-01-01

    Quantitative radiation dose-response curves for normal gustatory tissue in man were studied. Taste function, expressed as taste loss, was evaluated in 84 patients who were given either photon or neutron radiotherapy for tumors in the head and neck region. Patients were treated to average tumor doses of 6600 cGy (photon) or 2200 cGy intervals for photon patients and 320-cGy intervals for neutron patients during radiotherapy. The dose-response curves for photons and neutrons were analyzed by fitting a four-parameter logistic equation to the data. Photon and neutron curves differed principally in their relative position along the dose axis. Comparison of the dose-response curves were made by determination of RBE. At 320 cGy, the lowest neutron dose at which taste measurements were made, RBE = 5.7. If this RBE is correct, then the therapeutic gain factor may be equal to or less than 1, indicating no biological advantage in using neutrons over photons for this normal tissue. These studies suggest measurements of taste function and evaluation of dose-response relationships may also be useful in quantitatively evaluating the efficacy of chemical modifiers of radiation response such as hypoxic cell radiosensitizers and radioprotectors

  1. 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).

  2. Radiation protection aspects of a high flux, fast neutron generator

    International Nuclear Information System (INIS)

    DeLuca, P.M.; Torti, R.P.; Chenevert, G.M.; Tesmer, J.R.; Kelsey, C.A.

    1976-01-01

    During the development and operation of a gas target, DT neutron generator for use in cancer therapy, two radiation hazards were routinely encountered - personnel exposure to neutrons and to tritium. The principal hazard was irradiation by fast neutrons. By assembling the source below ground level, adding shielding and the use of a controlled access, key identification interlock, the neutron hazard has been reduced. With the present source strength of 2 x 10 12 n/sec, an average neutron dose rate in the control room of 20 mrem/hr was measured- a level compatible with a limited run schedule. The second hazard was exposure to tritium in both gaseous and solid forms. A target inventory of 90 Ci, and overall inventory of 500 Ci, and the need to modify and repair the generator present significant potential hazard due to tritium exposure. The use of protective gloves, wipe tests, urine assays, continuous room air monitoring, and equipment decontamination minimized personnel exposure and effectively confined contamination. The dose due to tritium has been ∼ 0.5 rem/year and negligible spread of contamination has occurred

  3. Comparison of doses delivered in clinical trials of neutron capture therapy in the USA

    International Nuclear Information System (INIS)

    Albritton, J.R.; Binns, P.J.; Riley, K.J.; Coderre, J.A.; Harling, O.K.; Kiger, W.S. III

    2006-01-01

    A combined 81 brain tumor patients have been treated in dose escalation trials of Neutron Capture Therapy (NCT) at Harvard-MIT and Brookhaven National Laboratory (BNL). Pooling the clinical outcomes from these trials will permit evaluation with more statistical rigor. However, differences in physical and computational dosimetry between the institutions make direct comparison of the clinical dosimetry difficult. This paper describes work performed to normalize the BNL clinical dosimetry to that of Harvard-MIT for combined dose response analysis. This normalization involved analysis of MIT measurements and calculations using the BNL treatment planning system (TPS), BNCT - Rtpe, for two different phantoms. The BNL TPS was calibrated to dose measurements made by MIT at the BMRR in the BNL calibration phantom, a Lucite cube, and then validated by MIT dose measurements at the BMMR in an ellipsoidal water phantom. Treatment plans for all BNL patients were recomputed using the newly determined TPS calibration, yielding reductions in reported mean brain doses of 19% on average in the initial 15 patients and 31% in the latter 38 patients. These reductions in reported doses have clinically significant implications for those relying on reported BNL doses as a basis for initial dose selection in clinical studies. (author)

  4. Application of semiconductor MOSFET and pin diode dosimeters to epithermal neutron beam dose distribution measurements in phantoms

    International Nuclear Information System (INIS)

    Carolan, M.G.; Wallace, S.A.; Allen, B.J.; Rosenfeld, A.B.; Mathur, J.N.

    1996-01-01

    For any clinical application of Boron Neutron Capture Therapy (BNCT) fast and accurate dose calculations will be required for treatment planning. Such calculations are also necessary for the planning and interpretation of results from pre-clinical and clinical trials where the speed of calculation is not so critical. A dose calculation system based on the MCNP Monte Carlo Neutron transport code has been developed by Wallace. This system takes image data from CT scans and constructs a voxel based geometrical model for input into MCNP. To validate the calculations, a number of phantoms were constructed and exposed in the HB11 epithermal neutron beam at the HFR of the CEC Joint Research Centre in Petten. The doses recorded by arrays of PIN diode neutron dosimeters and MOSFET gamma dosimeters in these phantoms were compared with the calculated results from the MCNP dose planning system. Initial results have been reported elsewhere. Poster 197. (author)

  5. High sensitivity neutron bursts detecting system

    International Nuclear Information System (INIS)

    Shyam, A.; Kaushik, T.C.; Srinivasan, M.; Kulkarni, L.V.

    1993-01-01

    Technique and instrumentation to detect multiplicity of fast neutrons, emitted in sharp bursts, has been developed. A bank of 16 BF 3 detectors, in an appropriate thermalising assembly, efficiency ∼ 16%, is used to detect neutron bursts. The output from this setup, through appropriate electronics, is divided into two paths. The first is directly connected to a computer controlled scalar. The second is connected to another similar scalar through a delay time unit (DTU). The DTU design is such that once it is triggered by a count pulse than it does not allow any counts to be recorded for a fixed dead time set at ∼ 100 μs. The difference in counts recorded directly and through DTU gives the total number of neutrons produced in bursts. This setup is being used to study lattice cracking, anomalous effects in solid deuterium systems and various reactor physics experiments. (author). 3 refs., 1 fig

  6. Influence of dose rate on the transformation of Syrian hamster embryo cells by fission-spectrum neutrons

    International Nuclear Information System (INIS)

    Jones, C.A.; Sedita, B.A.; Hill, C.K.; Elkind, M.M.

    1988-01-01

    Several explanations for this neutron dose-rate effect have been proposed, but further investigation is necessary to determine the mechanisms involved. In all cell transformation studies to date the immortalized, aneuploid 10T1/2 cell-line has been used. These cells may be premalignant; thus their response characteristics and, in particular, the nature of the transformation event, might differ from that in a normal, fibroblast cell. One reason for the present study was to determine whether the low-dose-rate effect of fission neutrons could be demonstrated in normal cells. If so, a normal cell system, which would more closely resemble a normal in vivo system, could be used for mechanistic studies. We chose Syrian hamster embryo (SHE) fibroblasts which are normal, diploid cells with a limited life span in culture. Upon exposure to low doses of ionizing radiation, the fraction of the cells that are transformed can be identified in a standard 8--10 day colony assay by examining their clonal morphology. Transformed cells form colonies with a dense, criss-crossed or piled-up structure. A high percentage of the transformed colonies can be further propagated and will acquire additional neoplastic characteristics; i.e., anchorage independence, immortality, altered proteolytic activity, karyotype alterations, and finally, tumorigenicity

  7. Influence of dose rate on the transformation of Syrian hamster embryo cells by fission-spectrum neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jones, C.A.; Sedita, B.A.; Hill, C.K.; Elkind, M.M.

    1988-01-01

    Several explanations for this neutron dose-rate effect have been proposed, but further investigation is necessary to determine the mechanisms involved. In all cell transformation studies to date the immortalized, aneuploid 10T1/2 cell-line has been used. These cells may be premalignant; thus their response characteristics and, in particular, the nature of the transformation event, might differ from that in a normal, fibroblast cell. One reason for the present study was to determine whether the low-dose-rate effect of fission neutrons could be demonstrated in normal cells. If so, a normal cell system, which would more closely resemble a normal in vivo system, could be used for mechanistic studies. We chose Syrian hamster embryo (SHE) fibroblasts which are normal, diploid cells with a limited life span in culture. Upon exposure to low doses of ionizing radiation, the fraction of the cells that are transformed can be identified in a standard 8--10 day colony assay by examining their clonal morphology. Transformed cells form colonies with a dense, criss-crossed or piled-up structure. A high percentage of the transformed colonies can be further propagated and will acquire additional neoplastic characteristics; i.e., anchorage independence, immortality, altered proteolytic activity, karyotype alterations, and finally, tumorigenicity.

  8. Dosimetric systems of high dose, dose rate and dose uniformity in food and medical products

    International Nuclear Information System (INIS)

    Vargas, J.; Vivanco, M.; Castro, E.

    2014-08-01

    In the Instituto Peruano de Energia Nuclear (IPEN) we use the chemical dosimetry Astm-E-1026 Fricke as a standard dosimetric system of reference and different routine dosimetric systems of high doses, according to the applied doses to obtain the desired effects in the treated products and the doses range determined for each type of dosimeter. Fricke dosimetry is a chemical dosimeter in aqueous solution indicating the absorbed dose by means an increase in absorbance at a specific wavelength. A calibrated spectrophotometer with controlled temperature is used to measure absorbance. The adsorbed dose range should cover from 20 to 400 Gy, the Fricke solution is extremely sensitive to organic impurities, to traces of metal ions, in preparing chemical products of reactive grade must be used and the water purity is very important. Using the referential standard dosimetric system Fricke, was determined to March 5, 2013, using the referential standard dosimetric system Astm-1026 Fricke, were irradiated in triplicate Fricke dosimeters, to 5 irradiation times (20; 30; 40; 50 and 60 seconds) and by linear regression, the dose rate of 5.400648 kGy /h was determined in the central point of the irradiation chamber (irradiator Gamma cell 220 Excel), applying the decay formula, was compared with the obtained results by manufacturers by means the same dosimetric system in the year of its manufacture, being this to the date 5.44691 kGy /h, with an error rate of 0.85. After considering that the dosimetric solution responds to the results, we proceeded to the irradiation of a sample of 200 g of cereal instant food, 2 dosimeters were placed at the lateral ends of the central position to maximum dose and 2 dosimeters in upper and lower ends as minimum dose, they were applied same irradiation times; for statistical analysis, the maximum dose rate was 6.1006 kGy /h and the minimum dose rate of 5.2185 kGy /h; with a dose uniformity of 1.16. In medical material of micro pulverized bone for

  9. Calculated intensity of high-energy neutron beams

    International Nuclear Information System (INIS)

    Mustapha, B.; Nolen, J.A.; Back, B.B.

    2004-01-01

    The flux, energy and angular distributions of high-energy neutrons produced by in-flight spallation and fission of a 400 MeV/A 238 U beam and by the break-up of a 400 MeV/A deuteron beam are calculated. In both cases very intense secondary neutron beams are produced, peaking at zero degrees, with a relatively narrow energy spread. Such secondary neutron beams can be produced with the primary beams from the proposed rare isotope accelerator driver linac. The break-up of a 400 kW deuteron beam on a liquid-lithium target can produce a neutron flux of >10 10 neutrons/cm 2 /s at a distance of 10 m from the target

  10. Bench mark spectra for high-energy neutron dosimetry

    International Nuclear Information System (INIS)

    Dierckx, R.

    1986-01-01

    To monitor radiation damage experiments, activation detectors are commonly used. The precision of the results obtained by the multiple foil analysis is largely increased by the intercalibration in bench-mark spectra. This technique is already used in dosimetry measurements for fission reactors. To produce neutron spectra similar to fusion reactor and high-energy high-intensity neutron sources (d-Li or spallation), accelerators can be used. Some possible solutions as p-Be and d-D 2 O neutron sources, useful as bench-mark spectra are described. (author)

  11. High speed motion neutron radiography of dynamic events

    International Nuclear Information System (INIS)

    Robinson, A.H.; Barton, J.P.

    1983-01-01

    The development of a technique that permits neutron radiographic analysis of dynamic processes over a period lasting from one to ten milliseconds is described. The key to the technique is the use of a neutron pulse broad enough to span the duration of a brief event and intense enough to allow recording of the results on a high-speed movie film at frame rates of 10,000 frames/sec. Some typical application results in ballistic studies and two-phase flow are shown and discussed. The use of scintillator screens in the high-speed motion neutron radiography system is summarized and the statistical limitations of the technique are discussed

  12. Organ and Effective Dose Coefficients for Cranial and Caudal Irradiation Geometries: Neutrons

    Science.gov (United States)

    Veinot, K. G.; Eckerman, K. F.; Hertel, N. E.; Hiller, M. M.

    2017-09-01

    With the introduction of new recommendations by ICRP Publication 103, the methodology for determining the protection quantity, effective dose, has been modified. The modifications include changes to the defined organs and tissues, the associated tissue weighting factors, radiation weighting factors, and the introduction of reference sex-specific computational phantoms (ICRP Publication 110). Computations of equivalent doses in organs and tissues are now performed in both the male and female phantoms and the sex-averaged values used to determine the effective dose. Dose coefficients based on the ICRP 103 recommendations were reported in ICRP Publication 116, the revision of ICRP Publication 74 and ICRU Publication 57. The coefficients were determined for the following irradiation geometries: anterior-posterior (AP), posterior-anterior (PA), right and left lateral (RLAT and LLAT), rotational (ROT), and isotropic (ISO). In this work, the methodology of ICRP Publication 116 was used to compute dose coefficients for neutron irradiation of the body with parallel beams directed upward from below the feet (caudal) and directed downward from above the head (cranial). These geometries may be encountered in the workplace from personnel standing on contaminated surfaces or volumes and from overhead sources. Calculations of organ and tissue absorbed doses for caudal and cranial exposures to neutrons ranging in energy from 10-9 MeV to 10 GeV have been performed using the MCNP6 radiation transport code and the adult reference voxel phantoms of ICRP Publication 110. At lower energies the effective dose per particle fluence for cranial and caudal exposures is less than AP orientations while above about 30 MeV the cranial and caudal values are greater.

  13. Nonlinear model of high-dose implantation

    International Nuclear Information System (INIS)

    Danilyuk, A.

    2001-01-01

    The models of high-dose implantation, using the distribution functions, are relatively simple. However, they must take into account the variation of the function of distribution of the implanted ions with increasing dose [1-4]. This variation takes place owing to the fact that the increase of the concentration of the implanted ions results in a change of the properties of the target. High-dose implantation is accompanied by sputtering, volume growth, diffusion, generation of defects, formation of new phases, etc. The variation of the distribution function is determined by many factors and is not known in advance. The variation within the framework of these models [1-4] is taken into account in advance by the introduction of intuitive assumptions on the basis of implicit considerations. Therefore, these attempts should be regarded as incorrect. The model prepared here makes it possible to take into account the sputtering of the target, volume growth and additional declaration on the implanted ions. Without any assumptions in relation to the variation of the distribution function with increasing dose. In our model it is assumed that the type of distribution function for small doses in a pure target substance is the same as in substances with implanted ions. A second assumption relates to the type of the distribution function valid for small doses in the given substances. These functions are determined as a result of a large number of theoretical and experimental investigations and are well-known at the present time. They include the symmetric and nonsymmetric Gauss distribution, the Pearson distribution, and others. We examine implantation with small doses of up to 10 14 - 10 15 cm -2 when the accurately known distribution is valid

  14. High dose gamma-ray standard

    International Nuclear Information System (INIS)

    Macrin, R.; Moraru, R.

    1999-01-01

    The high gamma-ray doses produced in a gamma irradiator are used, mainly, for radiation processing, i.e. sterilization of medical products, processing of food, modifications of polymers, irradiation of electronic devices, a.s.o. The used absorbed doses depend on the application and cover the range 10 Gy to 100 MGy. The regulations in our country require that the response of the dosimetry systems, used for the irradiation of food and medical products, be calibrated and traceable to the national standards. In order to be sure that the products receive the desired absorbed dose, appropriate dosimetric measurements must be performed, including the calibration of the dosemeters and their traceability to the national standards. The high dose gamma-ray measurements are predominantly based on the use of reference radiochemical dosemeters. Among them the ferrous sulfate can be used as reference dosemeter for low doses (up to 400 Gy) but due to its characteristics it deserves to be considered a standard dosemeter and to be used for transferring the conventional absorbed dose to other chemical dosemeters used for absorbed doses up to 100 MGy. The study of the ferrous sulfate dosemeter consisted in preparing many batches of solution by different operators in quality assurance conditions and in determining for all batches the linearity, the relative intrinsic error, the repeatability and the reproducibility. The principal results are the following: the linear regression coefficient: 0.999, the relative intrinsic error: max.6 %, the repeatability (for P* = 95 %): max.3 %, the reproducibility (P* = 95%): max.5 %. (authors)

  15. High-quality single crystals for neutron experiments

    Indian Academy of Sciences (India)

    studies and our collaborative research projects with other UK and international groups will be discussed. Keywords. Crystal growth; floating zone method; neutron scattering. ... of single crystals of new materials is a highly competitive business.

  16. An assessment of the secondary neutron dose in the passive scattering proton beam facility of the national cancer center

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Eun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Gyuseong [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Se Byeong [Proton Therapy Center, National Cancer Center, Goyang (Korea, Republic of)

    2017-06-15

    The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a 3He neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from 4.942 ± 0.031 mSv/Gy at the end of the field to 0.324 ± 0.006 mSv/Gy at 150 cm in axial distance.

  17. Determination of neutron dose from criticality accidents with bioassays for sodium-24 in blood and phosphorus-32 in hair

    International Nuclear Information System (INIS)

    Feng, Y.; Miller, L.F.; Brown, K.S.; Casson, W.H.; Mei, G.T.; Thein, M.

    1993-06-01

    A comprehensive review of accident neutron dosimetry using blood and hair analysis was performed and is summarized in this report. Experiments and calculations were conducted at Oak Ridge National Laboratory (ORNL) and the University of Tennessee (UT) to develop measurement techniques for the activity of 24 Na in blood and 32 P in hair for nuclear accident dosimetry. An operating procedure was established for the measurement of 24 Na in blood using an HPGe detector system. The sensitivity of the measurement for a 20-mL sample is 0.01-0.02 Gy of total neutron dose for hard spectra and below 0.005 Gy for soft spectra based on a 30- to 60-min counting time. The operating procedures for direct counting of hair samples are established using a liquid scintillation detector. Approximately 0.06-0.1 Gy of total neutron dose can be measured from a 1-g hair sample using this procedure. Detailed procedures for chemical dissolution and ashing of hair samples are also developed. A method is proposed to use blood and hair analysis for assessing neutron dose based on a collection of 98 neutron spectra. Ninety-eight blood activity-to-dose conversion factors were calculated. The calculated results for an uncollided fission spectrum compare favorably with previously published data for fission neutrons. This nuclear accident dosimetry system makes it possible to estimate an individual's neutron dose within a few hours after an accident if the accident spectrum can be approximated from one of 98 tabulated neutron spectrum descriptions. If the information on accident and spectrum description is not available, the activity ratio of 32 P in hair and 24 Na in blood can provide information related to the neutron spectrum for dose assessment

  18. Recent high-accuracy measurements of the 1S0 neutron-neutron scattering length

    International Nuclear Information System (INIS)

    Howell, C.R.; Chen, Q.; Gonzalez Trotter, D.E.; Salinas, F.; Crowell, A.S.; Roper, C.D.; Tornow, W.; Walter, R.L.; Carman, T.S.; Hussein, A.; Gibbs, W.R.; Gibson, B.F.; Morris, C.; Obst, A.; Sterbenz, S.; Whitton, M.; Mertens, G.; Moore, C.F.; Whiteley, C.R.; Pasyuk, E.; Slaus, I.; Tang, H.; Zhou, Z.; Gloeckle, W.; Witala, H.

    2000-01-01

    This paper reports two recent high-accuracy determinations of the 1 S 0 neutron-neutron scattering length, a nn . One was done at the Los Alamos National Laboratory using the π - d capture reaction to produce two neutrons with low relative momentum. The neutron-deuteron (nd) breakup reaction was used in other measurement, which was conducted at the Triangle Universities Nuclear Laboratory. The results from the two determinations were consistent with each other and with previous values obtained using the π - d capture reaction. The value obtained from the nd breakup measurements is a nn = -18.7 ± 0.1 (statistical) ± 0.6 (systematic) fm, and the value from the π - d capture experiment is a nn = -18.50 ± 0.05 ± 0.53 fm. The recommended value is a nn = -18.5 ± 0.3 fm. (author)

  19. Boron neutron capture therapy using mixed epithermal and thermal neutron beams in patients with malignant glioma-correlation between radiation dose and radiation injury and clinical outcome

    International Nuclear Information System (INIS)

    Kageji, Teruyoshi; Nagahiro, Shinji; Matsuzaki, Kazuhito; Mizobuchi, Yoshifumi; Toi, Hiroyuki; Nakagawa, Yoshinobu; Kumada, Hiroaki

    2006-01-01

    Purpose: To clarify the correlation between the radiation dose and clinical outcome of sodium borocaptate-based intraoperative boron neutron capture therapy in patients with malignant glioma. Methods and Materials: The first protocol (P1998, n = 8) prescribed a maximal gross tumor volume (GTV) dose of 15 Gy. In 2001, a dose-escalated protocol was introduced (P2001, n 11), which prescribed a maximal vascular volume dose of 15 Gy or, alternatively, a clinical target volume (CTV) dose of 18 Gy. Results: The GTV and CTV doses in P2001 were 1.1-1.3 times greater than those in P1998. The maximal vascular volume dose of those with acute radiation injury was 15.8 Gy. The mean GTV and CTV dose in long-term survivors with glioblastoma was 26.4 and 16.5 Gy, respectively. A statistically significant correlation between the GTV dose and median survival time was found. In the 11 glioblastoma patients in P2001, the median survival time was 19.5 months and 1- and 2-year survival rate was 60.6% and 37.9%, respectively. Conclusion: Dose escalation contributed to the improvement in clinical outcome. To avoid radiation injury, the maximal vascular volume dose should be <12 Gy. For long-term survival in patients with glioblastoma after boron neutron capture therapy, the optimal mean dose of the GTV and CTV was 26 and 16 Gy, respectively

  20. First observations of power MOSFET burnout with high energy neutrons

    International Nuclear Information System (INIS)

    Oberg, D.L.; Wert, J.L.; Normand, E.; Majewski, P.P.; Wender, S.A.

    1996-01-01

    Single event burnout was seen in power MOSFETs exposed to high energy neutrons. Devices with rated voltage ≥400 volts exhibited burnout at substantially less than the rated voltage. Tests with high energy protons gave similar results. Burnout was also seen in limited tests with lower energy protons and neutrons. Correlations with heavy-ion data are discussed. Accelerator proton data gave favorable comparisons with burnout rates measured on the APEX spacecraft. Implications for burnout at lower altitudes are also discussed

  1. Thermal neutron scattering studies of condensed matter under high pressures

    International Nuclear Information System (INIS)

    Carlile, C.J.; Salter, D.C.

    1978-01-01

    Although temperature has been used as a thermodynamic variable for samples in thermal neutron scattering experiments since the inception of the neutron technique, it is only in the last decade that high pressures have been utilised for this purpose. In the paper the problems particular to this field of work are outlined and a review is made of the types of high-pressure cells used and the scientific results obtained from the experiments. 103 references. (author)

  2. Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Manish K.; Alajo, Ayodeji B., E-mail: alajoa@mst.edu

    2016-08-11

    Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

  3. Qualitative dose response of the normal canine head to epithermal neutron irradiation with and without boron capture

    International Nuclear Information System (INIS)

    DeHaan, C.E.; Gavin, P.R.; Kraft, S.L.; Wheeler, F.J.; Atkinson, C.A.

    1992-01-01

    Boron Neutron Capture Therapy is being re-evaluated for the treatment of intracranial tumors. Prior to human clinical trials, determination of normal tissue tolerance is critical. Dogs were chosen as a large animal model for the following reasons. Dogs can be evaluated with advanced imaging, diagnostic and therapeutic modalities. Dogs are amenable to detailed neurologic examination and subtle behavioral changes are easily detected. Specifically, Labrador retrievers were chosen for their large body and head size. The dogs received varying doses of epithermal neutron irradiation and boron neutron capture irradiation using an epithermal neutron source. The dogs were closely monitored for up to one year post irradiation

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

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  5. Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

    1996-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

  6. Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

    International Nuclear Information System (INIS)

    Aizawa, O.

    1993-01-01

    In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

  7. The new high flux neutron source FRM-2 in Munich

    International Nuclear Information System (INIS)

    Roegler, H.J.; Wierheim, G.

    2002-01-01

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

  8. NIST high-dose calibration services

    International Nuclear Information System (INIS)

    Humphreys, J.C.

    1989-01-01

    There is a need for the standardization of high-dose measurements used in the radiation-processing industry in order to provide assured traceability to national standards. NIST provides dosimetry calibration services to this industry. One of these services involves administration of known absorbed doses of gamma rays to customer-supplied dosimeters. The dosimeters are packaged to provide electron equilibrium conditions and are irradiated in a standard 60 Co calibration facility; this provides a calibration of that batch of dosimeters. Another service consists of supplying to a customer calibrated transfer dosimeters for irradiation with the customer's radiation source. The irradiated transfer dosimeters are then returned to NIST for analysis; the results are reported to the customer, providing a calibration of the dose rate of the customer's source. (orig.)

  9. Recent developments on micrometric fission chambers for high neutron fluxes

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Recent developments on micrometric fission chambers for high neutron fluxes

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  11. Determination of the dose equivalents due to neutrons produced during therapeutic irradiations with a Varian CLINAC 2500; Determinacion de la dosis equivalente debida a neutrones producidos durante irradiaciones terapeuticas con un Varian CLINAC 2500

    Energy Technology Data Exchange (ETDEWEB)

    Carrillo, Ricardo E [Centro de Investigacion y de Estudios Avanzados del IPN, Unidad Merida, Yucatan (Mexico)

    1991-07-01

    This experiment it was designed to quantify that so important it is the dose equivalent deposited by the neutron flow that is generated by photonuclear reactions during therapeutic irradiations with X rays of produced high-energy for an accelerator Varian CLINAC 2500. This accelerator type is routinely used in the Department of Radiotherapy of the Hospital of the University of Wisconsin, E.U. The equivalent dose was measured in diverse towns of the room of irradiations using the activation of thin sheets of gold put in the center of plastic recipients full with water. In general, the recipients were 1 m or more than the floor and at distances still bigger than the walls. The irradiations were made using photons with the highest energy that you can select with this team - 24 MeV. The due equivalent dose to neutrons taken place here by the energy photons used they were measured and reported. (author)

  12. High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy.

    Science.gov (United States)

    Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Bisyakoev, M; Eliyahu, I; Feinberg, G; Hazenshprung, N; Kijel, D; Nagler, A; Silverman, I

    2011-12-01

    A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks. Copyright © 2011 Elsevier Ltd. All rights reserved.

  13. Calculation of dose distribution for 252Cf fission neutron source in tissue equivalent phantoms using Monte Carlo method

    International Nuclear Information System (INIS)

    Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong

    2001-01-01

    Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf

  14. Neutron shielding properties of a new high-density concrete

    International Nuclear Information System (INIS)

    Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

    2008-01-01

    The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

  15. High-performance instruments in neutron arena of JHP. Preliminary version

    International Nuclear Information System (INIS)

    Furusaka, M.; Itoh, S.; Otomo, T.; Arai, M.

    1996-05-01

    This report is a preliminary report of high-performance instruments in neutron arena of JHP (Japan Hadron Project). This report consists of as follows; neutron intensity of neutron arena, development of neutron sources in neutron arena, experimental devices and instrumentation. (J.P.N.)

  16. Neutron detection in a high gamma-ray background with EJ-301 and EJ-309 liquid scintillators

    International Nuclear Information System (INIS)

    Stevanato, L.; Cester, D.; Nebbia, G.; Viesti, G.

    2012-01-01

    Using a fast digitizer, the neutron–gamma discrimination capability of the new liquid scintillator EJ-309 is compared with that obtained using standard EJ-301. Moreover the capability of both the scintillation detectors to identify a weak neutron source in a high gamma-ray background is demonstrated. The probability of neutron detection is PD=95% at 95% confidence level for a gamma-ray background corresponding to a dose rate of 100 μSv/h.

  17. Neutron powder diffraction under high pressure at J-PARC

    International Nuclear Information System (INIS)

    Utsumi, Wataru; Kagi, Hiroyuki; Komatsu, Kazuki; Arima, Hiroshi; Nagai, Takaya; Okuchi, Takuo; Kamiyama, Takashi; Uwatoko, Yoshiya; Matsubayashi, Kazuyuki; Yagi, Takehiko

    2009-01-01

    It is expected that high-pressure material science and the investigation of the Earth's interior will progress greatly using the high-flux pulse neutrons of J-PARC. In this article, we introduce our plans for in situ neutron powder diffraction experiments under high pressure at J-PARC. The use of three different types of high-pressure devices is planned; a Paris-Edinburgh cell, a new opposed-anvil cell with a nano-polycrystalline diamond, and a cubic anvil high-pressure apparatus. These devices will be brought to the neutron powder diffraction beamlines to conduct a 'day-one' high-pressure experiment. For the next stage of research, we propose construction of a dedicated beamline for high-pressure material science. Its conceptual designs are also introduced here.

  18. Future directions in high-pressure neutron diffraction

    Science.gov (United States)

    Guthrie, M.

    2015-04-01

    The ability to manipulate structure and properties using pressure has been well known for many centuries. Diffraction provides the unique ability to observe these structural changes in fine detail on lengthscales spanning atomic to nanometre dimensions. Amongst the broad suite of diffraction tools available today, neutrons provide unique capabilities of fundamental importance. However, to date, the growth of neutron diffraction under extremes of pressure has been limited by the weakness of available sources. In recent years, substantial government investments have led to the construction of a new generation of neutron sources while existing facilities have been revitalized by upgrades. The timely convergence of these bright facilities with new pressure-cell technologies suggests that the field of high-pressure (HP) neutron science is on the cusp of substantial growth. Here, the history of HP neutron research is examined with the hope of gleaning an accurate prediction of where some of these revolutionary capabilities will lead in the near future. In particular, a dramatic expansion of current pressure-temperature range is likely, with corresponding increased scope for extreme-conditions science with neutron diffraction. This increase in coverage will be matched with improvements in data quality. Furthermore, we can also expect broad new capabilities beyond diffraction, including in neutron imaging, small angle scattering and inelastic spectroscopy.

  19. Proton induction linacs as high-intensity neutron sources

    International Nuclear Information System (INIS)

    Keefe, D.; Hoyer, E.

    1981-01-01

    Proton induction linacs are explored as high intensity neutron sources. The induction linac - concept, properties, experience with electrons, and possibilities - and its limitations for accelerating ions are reviewed. A number of proton induction linac designs are examined with the LIACEP program and general conclusions are given. Results suggest that a proton induction accelerator of the lowest voltage, consistent with good neutron flux, is preferred and could well be cost competitive with the usual rf linac/storage ring designs. (orig.)

  20. High yield neutron generators using the DD reaction

    Energy Technology Data Exchange (ETDEWEB)

    Vainionpaa, J. H.; Harris, J. L.; Piestrup, M. A.; Gary, C. K.; Williams, D. L.; Apodaca, M. D.; Cremer, J. T. [Adelphi technology, 2003 E. Bayshore Rd. 94061, Redwood City, CA (United States); Ji, Qing; Ludewigt, B. A. [Lawrence Berkeley National Lab, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Jones, G. [G and J Enterprise, 1258 Quary Ln, Suite F, Pleasanton California 94566 (United States)

    2013-04-19

    A product line of high yield neutron generators has been developed at Adelphi technology inc. The generators use the D-D fusion reaction and are driven by an ion beam supplied by a microwave ion source. Yields of up to 5 Multiplication-Sign 10{sup 9} n/s have been achieved, which are comparable to those obtained using the more efficient D-T reaction. The microwave-driven plasma uses the electron cyclotron resonance (ECR) to produce a high plasma density for high current and high atomic ion species. These generators have an actively pumped vacuum system that allows operation at reduced pressure in the target chamber, increasing the overall system reliability. Since no radioactive tritium is used, the generators can be easily serviced, and components can be easily replaced, providing essentially an unlimited lifetime. Fast neutron source size can be adjusted by selecting the aperture and target geometries according to customer specifications. Pulsed and continuous operation has been demonstrated. Minimum pulse lengths of 50 {mu}s have been achieved. Since the generators are easily serviceable, they offer a long lifetime neutron generator for laboratories and commercial systems requiring continuous operation. Several of the generators have been enclosed in radiation shielding/moderator structures designed for customer specifications. These generators have been proven to be useful for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA) and fast neutron radiography. Thus these generators make excellent fast, epithermal and thermal neutron sources for laboratories and industrial applications that require neutrons with safe operation, small footprint, low cost and small regulatory burden.

  1. The effect of a paraffin screen on the neutron dose at the maze door of a 15 MV linear accelerator.

    Science.gov (United States)

    Krmar, M; Nikolić, D; Kuzmanović, A; Kuzmanović, Z; Ganezer, K

    2013-08-01

    The purpose of this study was to explore the effects of a paraffin screen located at various positions in the maze on the neutron dose equivalent at the maze door. The neutron dose equivalent was measured at the maze door of a room containing a 15 MV linear accelerator for x-ray therapy. Measurements were performed for several positions of the paraffin screen covering only 27.5% of the cross-sectional area of the maze. The neutron dose equivalent was also measured at all screen positions. Two simple models of the neutron source were considered in which the first assumed that the source was the cross-sectional area at the inner entrance of the maze, radiating neutrons in an isotropic manner. In the second model the reduction in the neutron dose equivalent at the maze door due to the paraffin screen was considered to be a function of the mean values of the neutron fluence and energy at the screen. The results of this study indicate that the equivalent dose at the maze door was reduced by a factor of 3 through the use of a paraffin screen that was placed inside the maze. It was also determined that the contributions to the dosage from areas that were not covered by the paraffin screen as viewed from the dosimeter, were 2.5 times higher than the contributions from the covered areas. This study also concluded that the contributions of the maze walls, ceiling, and floor to the total neutron dose equivalent were an order of magnitude lower than those from the surface at the far end of the maze. This study demonstrated that a paraffin screen could be used to reduce the neutron dose equivalent at the maze door by a factor of 3. This paper also found that the reduction of the neutron dose equivalent was a linear function of the area covered by the maze screen and that the decrease in the dose at the maze door could be modeled as an exponential function of the product φ·E at the screen.

  2. High energy fast neutrons from the Harwell variable energy cyclotron. II. Biologic studies in mammalian systems

    International Nuclear Information System (INIS)

    Berry, R.J.; Bance, D.A.; Barnes, D.W.H.; Cox, R.; Goodhead, D.T.; Sansom, J.M.; Thacker, J.

    1977-01-01

    A high energy fast neutron beam potentially suitable for radiotherapy has been described in a companion paper. Its biologic effects have been studied in the following experimental systems: clonal survival and mutation induction after irradiation in vitro in Chinese hamster cells and human diploid fibroblasts; survival of reproductive capacity in vivo of murine hemopoietic colony-forming cells and murine intestinal crypts after irradiation in vivo; survival of reproductive capacity in vivo after irradiation in vitro or in vivo of murine lymphocytic leukemia cells; acute intestinal death following total body irradiation of mice and guinea pigs; and hemopoietic death following total body irradiation of mice and guinea pigs. The relative biologic effectiveness of these high energy neutrons varied among the different biologic systems, and in several cases varied with the size of the radiation dose. The oxygen enhancement ratio was studied in murine lymphocytic leukemia cells irradiated under aerobic or hypoxic conditions in vitro and assayed for survival of reproductive capacity in vivo. Compared with x-rays, the potential therapeutic gain factor for these neutrons was about 1.5. This work represents a ''radiobiologic calibration'' program which it is suggested should be undertaken before new and unknown fast neutron spectra are used for experimental radiotherapy. The results are compared with biologic studies carried out at high energy fast neutron generators in the United States

  3. Theory of Pulsed Neutron Experiments in Highly Heterogeneous Multiplying Media

    International Nuclear Information System (INIS)

    Corno, S.E.

    1965-01-01

    In this work we investigate the time and space dependence of the neutron flux within a highly heterogeneous assembly, in which pulsed or sinusoidally modulated neutrons are injected. We consider, for the sake of simplicity, a device consisting of a cylindrical block of heavy moderator, along the axis of which a line-shaped region of fissionable material is located. The driving neutron source is assumed to be located on one of the end faces of the cylinder. The extent of the fissionable region allows us to deal with it as with an absorbing and multiplying singularity of the neutron field. As our attention is mostly concentrated on space and time variation of the neutron flux, rather crude approximations are assumed as far as the energy dependence of the neutron population is concerned. Within the limits of the age-diffusion theory, the response of the device to any neutron excitation may be found in closed form. For a sinusoidally modulated source of given frequency, it may easily be shown that, if the axial singularity were a purely absorbing one, the neutron waves being propagated along the device would possess a phase shift; a wavelength and an attenuation constant depending on the absorbing properties of the singularity. This picture becomes more and more complicated when neutron multiplication occurs. For this general case the solution derived in our paper obviously turns out to be dependent on both absorption and multiplication properties of the singularity. This circumstance suggests, among others, the idea of using a device of the type described above for testing fuel elements of heterogeneous reactors. (author) [fr

  4. High-dose erythropoietin for tissue protection

    DEFF Research Database (Denmark)

    Lund, Anton; Lundby, Carsten; Olsen, Niels Vidiendal

    2014-01-01

    BACKGROUND: The discovery of potential anti-apoptotic and cytoprotective effects of recombinant human erythropoietin (rHuEPO) has led to clinical trials investigating the use of high-dose, short-term rHuEPO therapy for tissue protection in conditions such as stroke and myocardial infarction....... Experimental studies have been favourable, but the clinical efficacy has yet to be validated. MATERIALS AND METHODS: We have reviewed clinical studies regarding the use of high-dose, short-term rHuEPO therapy for tissue protection in humans with the purpose to detail the safety and efficacy of r...... no effect of rHuEPO therapy on measures of tissue protection. Five trials including 1025 patients reported safety concerns in the form of increased mortality or adverse event rates. No studies reported reduced mortality. CONCLUSIONS: Evidence is sparse to support a tissue-protective benefit of r...

  5. Dose-incidence relationships for exencephalia, anophthalmia and prenatal mortality in mouse embryos irradiated with fission neutrons or 250 kV X-rays

    Energy Technology Data Exchange (ETDEWEB)

    Friedberg, W.; Faulkner, D.N.; Neas, B.R.; Hanneman, G.D.; Darden, E.B. Jr.; Deal, R.B. Jr.; Parker, D.E.

    1987-08-01

    Groups of pregnant mice were irradiated at selected times between 10.00 hours on gestation day 7 and 16.00 hours on day 8. Each group received 0.39 Gy of neutrons or 1.60 Gy of X-rays, or was sham irradiated. We identified a period of high susceptibility of the embryos to radiation-induced exencephalia, anophthalmia and prenatal mortality early in gestation day 8. Dose-incidence relationships in this period were investigated with 0.19-0.48 Gy of neutrons and with 0.40-2.00 Gy of X-rays.

  6. Dose-incidence relationships for exencephalia, anophthalmia and prenatal mortality in mouse embryos irradiated with fission neutrons or 250 kV X-rays

    International Nuclear Information System (INIS)

    Friedberg, W.; Faulkner, D.N.; Neas, B.R.; Hanneman, G.D.; Deal, R.B. Jr.; Parker, D.E.

    1987-01-01

    Groups of pregnant mice were irradiated at selected times between 10.00 hours on gestation day 7 and 16.00 hours on day 8. Each group received 0.39 Gy of neutrons or 1.60 Gy of X-rays, or was sham irradiated. We identified a period of high susceptibility of the embryos to radiation-induced exencephalia, anophthalmia and prenatal mortality early in gestation day 8. Dose-incidence relationships in this period were investigated with 0.19-0.48 Gy of neutrons and with 0.40-2.00 Gy of X-rays. (author)

  7. Influence of the neutron flux shape on the value of absorbed neutron dose; Uticaj oblika neutronskog spektra na vrednost apsorbovane doze neutrona

    Energy Technology Data Exchange (ETDEWEB)

    Miric, I; Miric, P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    This paper deals with the study od specific doses dependence on the type and approximation procedures of neutron spectra. Values of specific dose rates (dose per neutron cm{sub 2}) were analysed for neutron spectra from RB reactor in Vinca, Crac facility in Valduc (France) and HPRR reactor in Oak Ridge (USA). Data used in this analysis were obtained by methods used in Harwell (AERE), Oak Ridge (ORNL), Chalk River (AECL), CEN de Cadarache (CEA) and in the Boris Kidric Institute (IBK). Specific absorbed neutron doses were determined for each of the estimated spectra and presented in the form of kerma/(n.cm{sup -2}) and rad/((n.cm{sup -2}) units. The obtained results have shown the influence of the flux approximation procedure on the values of conversion factors for obtaining neutron doses from neutron flux. U okviru ovog rada radjeno je na ispitivanju zavisnosti specificnih doza od vrste i nacina aproksimacije neutronskog spektra. U radu su analizirane vrednosti specificnih doza (doza po n.cm{sup -2}) za neutronske spektre koji se dobijaju oko sledecih nuklearnih postrojenja: reaktora RB u Vinci, postrojenja CRAC u Valduc-u (Francuska), reaktora HPRR u Oak Ridge-u (SAD). Za analizu su korisceni podaci dobijeni metodama koje se koriste u nuklearnim centrima Harwell (AERE), Oak Ridge-u (ORNL), Chalk River-u (AECL), CEN de Cadarache (CEA) i Institutu Boris Kidric (IBK). Za svaki procenjeni spektar odredjene su specificne apsorbovane doze neutrona izrazene u kerma/(n.cm{sup -2}) i rad/(n.cm{sup -2}) jedinicama. Dobijeni rezultati su pokazali koliko nacin aproksimacije spektra utice na vrednost konverzionih faktora koji sluze za prelazak sa fluksa na dozu neutrona (author)

  8. Determination of the energy dose of a neutron beam using a ionization chamber. A compendium

    International Nuclear Information System (INIS)

    Schraube, H.; Alberts, W.G.; Brede, H.; Burgkhardt, B.; Piesch, E.; Doerschel, B.; Heinzelmann, M.; Hess, A.; Hoefert, M.

    2003-01-01

    This report is addressed to scientists and technicians, who are engaged in dosimetry of fast neutrons, especially for purposes of percutaneous radiation therapy. The range of mean energies of the radiation sources may be approximately between 1 MeV and 50 MeV. The report exhibits a compendium, which describes the basic methods and procedures for the determination of energy dose in tissue in a phantom or free-in-air, where applicable. Furthermore, requirements for monitor, test devices and phantom materials are described. The calculation methods are comprehensibly derived and supplemented with numerical data. A detailed analysis of experimental uncertainties is completed with practical examples. (orig.) [de

  9. Attenuation of the neutron and γ ray dose in concrete channels

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1983-08-01

    The calculations of the γ and neutron dose in concrete channels is described. The method is based on the Monte Carlo procedure. One series of results obtained in straight channels shows the influence of the source spectra and geometry and thus the channel form. A second series shows the attenuation produced by bends along the length of the channel; the variation of the branch length is also studied. The results are generalised and represented by a simple formula. The parameters are adjusted to the curves obtained by the Monte Carlo programme. (G.T.H.)

  10. Radiation doses from radiation sources of neutrons and photons by different computer calculation

    International Nuclear Information System (INIS)

    Siciliano, F.; Lippolis, G.; Bruno, S.G.

    1995-12-01

    In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoric background are concerned

  11. Crystal growth in EPDM by chemi-crystallisation as a function of the neutron irradiation dose and flux level

    International Nuclear Information System (INIS)

    Lambri, O.A.; Salvatierra, L.M.; Sanchez, F.A.; Matteo, C.L.; Sorichetti, P.A.; Celauro, C.A.

    2005-01-01

    Neutron irradiation at room temperature were performed on EPDM (ethylene-propylene-diene monomer) in two different nuclear reactors at different fluxes. The effect of the irradiation on the chain arrangement in the polymer, as a function of the dose is discussed. Different crystal concentrations and crystal shapes, developed by chemi-crystallisation, are obtained depending on the neutron dose. In addition the radiation damage degree in the polymer depends both on the dose and the flux level. Dynamical mechanical analysis, swelling studies, X-ray diffraction, differential thermal analysis and infrared studies were employed as experimental techniques

  12. Neutron dose rate in the upper part of a PWR containment. Comparison between measurements and TRIPOLI-2 calculations

    International Nuclear Information System (INIS)

    Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.

    1984-01-01

    Conception of a reactor building requires large openings in the primary concrete shield for a postulated loss-of-coolant accident. Through these openings neutrons escape and produce dose rates in several parts of the reactor building. Some calculations using ANISN, DOT and essentially TRIPOLI-2 codes allow to compute the neutron dose rates at several places such as reactor containment operating floor and containment annulus. Some complementary shields are provided and the instrumentations are placed in area where the dose rate is lower. Comparisons are presented between measurements and calculations

  13. Neutron dose study with bubble detectors aboard the International Space Station as part of the Matroshka-R experiment

    International Nuclear Information System (INIS)

    Machrafi, R.; Garrow, K.; Ing, H.; Smith, M. B.; Andrews, H. R.; Akatov, Yu; Arkhangelsky, V.; Chernykh, I.; Mitrikas, V.; Petrov, V.; Shurshakov, V.; Tomi, L.; Kartsev, I.; Lyagushin, V.

    2009-01-01

    As part of the Matroshka-R experiments, a spherical phantom and space bubble detectors (SBDs) were used on board the International Space Station to characterise the neutron radiation field. Seven experimental sessions with SBDs were carried out during expeditions ISS-13, ISS-14 and ISS-15. The detectors were positioned at various places throughout the Space Station, in order to determine dose variations with location and on/in the phantom in order to establish the relationship between the neutron dose measured externally to the body and the dose received internally. Experimental data on/in the phantom and at different locations are presented. (authors)

  14. Measurement of the energy spectrum of cosmic-ray induced neutrons aboard an ER-2 high-altitude airplane

    CERN Document Server

    Goldhagen, P E; Kniss, T; Reginatto, M; Singleterry, R C; Van Steveninck, W; Wilson, J W

    2002-01-01

    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 aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the atmospheric ionizing radiation (AIR) project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (t...

  15. High Energy Neutron Induced Gamma Production

    International Nuclear Information System (INIS)

    Brown, D.A.; Johnson, M.; Navratil, P.

    2007-01-01

    N Division has an interest in improving the physics and accuracy of the gamma data it provides to its customers. It was asked to look into major gamma producing reactions for 14 MeV incident neutrons for several low-Z materials and determine whether LLNL's processed data files faithfully represent the current state of experimental and theoretical knowledge for these reactions. To address this, we surveyed the evaluations of the requested materials, made recommendations for the next ENDL release and noted isotopes that will require further experimental study. This process uncovered several major problems in our translation and processing of the ENDF formatted evaluations, most of which have been resolved

  16. Unique furnace system for high-energy-neutron experiments

    International Nuclear Information System (INIS)

    Panayotou, N.F.; Green, D.R.; Price, L.S.

    1982-03-01

    The low flux of high energy neutron sources requires optimum utilization of the available neutron field. A furnace system has been developed in support of the US DOE fusion materials program which meets this challenge. Specimens positioned in two temperature zones just 1 mm away from the outside surface of a neutron window in the furnace enclosure can be irradiated simultaneously at two independent, isothermal (+- 1 0 C) temperatures. The temperature difference between these closely spaced isothermal zones is controllable from 0 to 320 0 C and the maximum temperature is 400 0 C. The design of the system also provides a controlled specimen environment, rapid heating and cooling and easy access to heaters and thermocouples. This furnace system is in use at the Rotating Target Neutron Source-II of Lawrence Livermore National Laboratory

  17. 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

  18. High frame-rate neutron radiography of dynamic events

    International Nuclear Information System (INIS)

    Bossi, R.H.; Robinson, A.H.; Barton, J.P.

    1981-01-01

    A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames/second. Synchronization has provided high-speed-motion neutron radiographs for evaluation of the firing cycle of 7.62 mm munition rounds within a steel rifle barrel. The system has also been used to demonstrate the ability to produce neutron radiographic movies of two-phase flow. The equipment uses the Oregon State University TRIGA reactor capable of pulsing to 3000 MW peak power, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16 mm high speed movie camera. The peak neutron flux incident at the object position is approximately 4 x 10 11 n/cm 2 s with a pulse, full width at half maximum, of 9 ms. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on the image quality. Modulation transfer function analysis has been used to assist in the evaluation of the system performance

  19. High frame-rate neutron radiography of dynamic events

    International Nuclear Information System (INIS)

    Bossi, R.H.; Robinson, A.H.; Barton, J.P.

    1983-01-01

    A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames/second. Synchronization has provided high-speed-motion neutron radiographs for evaluation of the firing cycle of 7.62 mm munition rounds within a steel rifle barrel. The system has also been used to demonstrate the ability to produce neutron radiographic movies of two phase flow. The equipment uses the Oregon State University TRIGA reactor capable of pulsing to 3000 MW peak power, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16 mm high speed movie camera. The peak neutron flux incident at the object position is approximately 4 x 10 11 n/cm 2 s with a pulse, full width at half maximum, of 9 ms. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on the image quality. Modulation transfer function analysis has been used to assist in the evaluation of the system performance. (Auth.)

  20. Capability of NIPAM polymer gel in recording dose from the interaction of 10B and thermal neutron in BNCT

    International Nuclear Information System (INIS)

    Khajeali, Azim; Reza Farajollahi, Ali; Kasesaz, Yaser; Khodadadi, Roghayeh; Khalili, Assef; Naseri, Alireza

    2015-01-01

    The capability of N-isopropylacrylamide (NIPAM) polymer gel to record the dose resulting from boron neutron capture reaction in BNCT was determined. In this regard, three compositions of the gel with different concentrations of 10 B were prepared and exposed to gamma radiation and thermal neutrons. Unlike irradiation with gamma rays, the boron-loaded gels irradiated by neutron exhibited sensitivity enhancement compared with the gels without 10 B. It was also found that the neutron sensitivity of the gel increased by the increase of concentration of 10 B. It can be concluded that NIPAM gel might be suitable for the measurement of the absorbed dose enhancement due to 10 B and thermal neutron reaction in BNCT. - Highlights: • Three compositions of NIPAM gel with different concentration of 10 B have been exposed by gamma and thermal neutron. • The vials containing NIPAM gel have been irradiated by an automatic system capable of providing for dose uniformity. • Suitability of NIPAM polymer gel in measuring radiation doses in BNCT has been investigated.

  1. High dose rate brachytherapy for oral cancer

    International Nuclear Information System (INIS)

    Yamazaki, Hideya; Yoshida, Ken; Yoshioka, Yasuo; Shimizutani, Kimishige; Koizumi, Masahiko; Ogawa, Kazuhiko; Furukawa, Souhei

    2013-01-01

    Brachytherapy results in better dose distribution compared with other treatments because of steep dose reduction in the surrounding normal tissues. Excellent local control rates and acceptable side effects have been demonstrated with brachytherapy as a sole treatment modality, a postoperative method, and a method of reirradiation. Low-dose-rate (LDR) brachytherapy has been employed worldwide for its superior outcome. With the advent of technology, high-dose-rate (HDR) brachytherapy has enabled health care providers to avoid radiation exposure. This therapy has been used for treating many types of cancer such as gynecological cancer, breast cancer, and prostate cancer. However, LDR and pulsed-dose-rate interstitial brachytherapies have been mainstays for head and neck cancer. HDR brachytherapy has not become widely used in the radiotherapy community for treating head and neck cancer because of lack of experience and biological concerns. On the other hand, because HDR brachytherapy is less time-consuming, treatment can occasionally be administered on an outpatient basis. For the convenience and safety of patients and medical staff, HDR brachytherapy should be explored. To enhance the role of this therapy in treatment of head and neck lesions, we have reviewed its outcomes with oral cancer, including Phase I/II to Phase III studies, evaluating this technique in terms of safety and efficacy. In particular, our studies have shown that superficial tumors can be treated using a non-invasive mold technique on an outpatient basis without adverse reactions. The next generation of image-guided brachytherapy using HDR has been discussed. In conclusion, although concrete evidence is yet to be produced with a sophisticated study in a reproducible manner, HDR brachytherapy remains an important option for treatment of oral cancer. (author)

  2. High dose rate brachytherapy for oral cancer.

    Science.gov (United States)

    Yamazaki, Hideya; Yoshida, Ken; Yoshioka, Yasuo; Shimizutani, Kimishige; Furukawa, Souhei; Koizumi, Masahiko; Ogawa, Kazuhiko

    2013-01-01

    Brachytherapy results in better dose distribution compared with other treatments because of steep dose reduction in the surrounding normal tissues. Excellent local control rates and acceptable side effects have been demonstrated with brachytherapy as a sole treatment modality, a postoperative method, and a method of reirradiation. Low-dose-rate (LDR) brachytherapy has been employed worldwide for its superior outcome. With the advent of technology, high-dose-rate (HDR) brachytherapy has enabled health care providers to avoid radiation exposure. This therapy has been used for treating many types of cancer such as gynecological cancer, breast cancer, and prostate cancer. However, LDR and pulsed-dose-rate interstitial brachytherapies have been mainstays for head and neck cancer. HDR brachytherapy has not become widely used in the radiotherapy community for treating head and neck cancer because of lack of experience and biological concerns. On the other hand, because HDR brachytherapy is less time-consuming, treatment can occasionally be administered on an outpatient basis. For the convenience and safety of patients and medical staff, HDR brachytherapy should be explored. To enhance the role of this therapy in treatment of head and neck lesions, we have reviewed its outcomes with oral cancer, including Phase I/II to Phase III studies, evaluating this technique in terms of safety and efficacy. In particular, our studies have shown that superficial tumors can be treated using a non-invasive mold technique on an outpatient basis without adverse reactions. The next generation of image-guided brachytherapy using HDR has been discussed. In conclusion, although concrete evidence is yet to be produced with a sophisticated study in a reproducible manner, HDR brachytherapy remains an important option for treatment of oral cancer.

  3. Update on pediatric resuscitation drugs: high dose, low dose, or no dose at all.

    Science.gov (United States)

    Sorrentino, Annalise

    2005-04-01

    Pediatric resuscitation has been a topic of discussion for years. It is difficult to keep abreast of changing recommendations, especially for busy pediatricians who do not regularly use these skills. This review will focus on the most recent guidelines for resuscitation drugs. Three specific questions will be discussed: standard dose versus high-dose epinephrine, amiodarone use, and the future of vasopressin in pediatric resuscitation. The issue of using high-dose epinephrine for cardiopulmonary resuscitation refractory to standard dose epinephrine has been a topic of debate for many years. Recently, a prospective, double-blinded study was performed to help settle the debate. These results will be reviewed and compared with previous studies. Amiodarone is a medication that was added to the pediatric resuscitation algorithms with the most recent recommendations from the American Heart Association in 2000. Its use and safety will also be discussed. Another topic that is resurfacing in resuscitation is the use of vasopressin. Its mechanism and comparisons to other agents will be highlighted, although its use in the pediatric patient has not been thoroughly studied. Pediatric resuscitation is a constantly evolving subject that is on the mind of anyone taking care of sick children. Clinicians are continually searching for the most effective methods to resuscitate children in terms of short- and long-term outcomes. It is important to be familiar with not only the agents being used but also the optimal way to use them.

  4. 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

  5. Hardening of single crystals of magnesium by low neutron doses at 77 K

    International Nuclear Information System (INIS)

    Gonzalez, H.C.

    1984-01-01

    Radiation hardening in Mg single crystals at 77 K is studied with a microtensile machine operating in-situ in the CNEA reactor facility RA1. Experimental results show that the dose dependence of the yield stress is similar to that previously observed in irradiated Cu and Zn. The radiation-induced yield stress, due to the presence of radiation obstacles operating alone, increases according to a 0.5 power law. It adds algebraically to the athermal component of the initial yield stress, but is not exactly additive to the other thermally activated mechanisms. For doses higher than 4.5 x 10 16 neutrons/cm 2 , a strong instability in the deformation is observed. Post irradiation experiments in tensile tests performed with a hard machine show a continuous stress drop. This effect is attributed to the dislocation channeling phenomenon which takes place during the tensile test. (author)

  6. The Influence of Used Construction Material and Its Thickness on the Neutron Dose Rate Around the Linear Accelerator - Experimental Results

    International Nuclear Information System (INIS)

    Krpan, I; Miklavcic, I.; Poje, M.; Radolic, V.; Vukovic, B.; Zivkovic, A.; Faj, D.; Ivkovic, A.

    2013-01-01

    Since linear accelerators for medical radiotherapy do not have active radioactive sources it makes them adequate from the radioprotection point of view. However, when operating at the energy higher than 10 MeV, they can become a source of unwanted neutron radiation in the giant dipole resonance reaction between the photon beam and the accelerator head material. Neutrons created in this reaction are almost isotropic in direction with an energy range between 700 keV and 1 MeV. During the accelerator installation and different phases of the construction work around the accelerator, a neutron dose rate at several important locations was investigated. Both passive (solid state nuclear track etched detectors - CR 39 and/or LR-115 with the 10B foil) and active detectors (Thermo Biorem FHT 752) were used. A higher photon dose rate was measured around the accelerator facility. An effective photon dose reduction was achieved using steel plates. However, this was the secondary source of neutrons in the reaction between the photons and steel plates, since higher values were measured. Neutron reduction was done by additional layers of barite concrete. A very conservative assessment of the effective dose was done for the medical personnel inside the control room. At the accelerator extreme operating regime (fixed accelerator direction - gantry angle, highest energy possible used), the neutron dose rate in the control room of 12 μSv/h was measured. Knowing the number of working days and number of patients per technician (per day), an exposure to the neutron dose of 1,1 mSv per year was calculated.(author)

  7. Neutronic calculations in support of the design of the ITER High Resolution Neutron Spectrometer

    International Nuclear Information System (INIS)

    Moro, F.; Esposito, B.; Marocco, D.; Villari, R.; Petrizzi, L.; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Dapena, M.

    2011-01-01

    This paper presents the results of neutronic calculations performed to address important issues related to the optimization of the ITER HRNS (High resolution Neutron Spectrometer) design, in particular concerning the definition of the collimator and the choice of the detector system. The calculations have been carried out using the MCNP5 Monte Carlo code in a full 3-D geometry. The HRNS collimation system has been included in the latest MCNP ITER 40 o model (Alite-4). The ITER scenario 2 reference DT plasma fusion neutron source peaked at 14.1 MeV with Gaussian energy distribution has been used. Neutron fluxes and energy spectra (>1 MeV) have been evaluated at different positions along the HRNS collimator and at the detector location. The noise-to-signal ratio (i.e. the ratio of collided to uncollided neutrons), the breakdown of the collided spectrum into its components, the dependency on the first wall aperture and the gamma-ray spectra at the detector position have also been analyzed. The impact of the results on the design of the HRNS diagnostic system is discussed.

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

    International Nuclear Information System (INIS)

    Hajnal, F.; Wilson, J.

    1992-01-01

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

  9. 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.)

  10. Photo neutron dose equivalent rate in 15 MV X-ray beam from a Siemens Primus Linac

    Directory of Open Access Journals (Sweden)

    A Ghasemi

    2015-01-01

    Full Text Available Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF 3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER utilizing conversion factors of American Association of Physicist in Medicine′s (AAPM report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10 -6 (3 m from isocenter in +Y direction, 0 × 0 field size and 8.36 × 10 -8 Sv/min (in maze, 40 × 40 field size, respectively. For fast neutrons, maximum DERs using 9" and 3" moderators were 1.6 × 10 -5 and 1.74 × 10 -5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size, respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons.

  11. Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bartlett, D. [National Radiological Protection Board, Chilton (United Kingdom); Drake, P. [Vattenfall AB, Vaeroebacka (Sweden); Lindborg, L. [Swedish Radiation Protection Inst., Stockholm (Sweden); Klein, H. [Physikalisch-Technische Bundesanstalt, Braunschweig (Germany); Schmitz, Th. [Forschungszentrum Juelich GmbH, Juelich (Germany); Tichy, M

    1999-06-01

    Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification.

  12. Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

    International Nuclear Information System (INIS)

    Bartlett, D.; Drake, P.; Lindborg, L.; Klein, H.; Schmitz, Th.; Tichy, M.

    1999-06-01

    Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification

  13. Hexagonal boron nitride neutron detectors with high detection efficiencies

    Science.gov (United States)

    Maity, A.; Grenadier, S. J.; Li, J.; Lin, J. Y.; Jiang, H. X.

    2018-01-01

    Neutron detectors fabricated from 10B enriched hexagonal boron nitride (h-10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer on both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h-10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.

  14. High speed motion neutron radiography of two-phase flow

    International Nuclear Information System (INIS)

    Robinson, A.H.; Wang, S.L.

    1983-01-01

    Current research in the area of two-phase flow utilizes a wide variety of sensing devices, but some limitations exist on the information which can be obtained. Neutron radiography is a feasible alternative to ''see'' the two-phase flow. A system to perform neutron radiographic analysis of dynamic events which occur on the order of several milliseconds has been developed at Oregon State University. Two different methods have been used to radiograph the simulated two-phase flow. These are pulsed, or ''flash'' radiography, and high speed movie neutron radiography. The pulsed method serves as a ''snap-shot'' with an exposure time ranging from 10 to 20 milliseconds. In high speed movie radiography, a scintillator is used to convert neutrons into light which is enhanced by an optical intensifier and then photographed by a high speed camera. Both types of radiography utilize the pulsing capability of the OSU TRIGA reactor. The principle difficulty with this type of neutron radiography is the fogging of the image due to the large amount of scattering in the water. This difficulty can be overcome by using thin regions for the two-phase flow or using heavy water instead of light water. The results obtained in this paper demonstrate the feasibility of using neutron radiography to obtain data in two-phase flow situations. Both movies and flash radiographs have been obtained of air bubbles in water and boiling from a heater element. The neutron radiographs of the boiling element show both nucleate boiling and film boiling. (Auth.)

  15. High temperature neutron diffraction study of LaPO4

    International Nuclear Information System (INIS)

    Mishra, S.K.; Mittal, R.; Ningthoujam, R.S.; Vatsa, R.K.; Hansen, T.

    2016-01-01

    We report high temperature powder neutron diffractions study in LaPO 4 using high-flux D20 neutron diffractometer in the Institut Laue-Langevin, France. The measurements were carried out in high resolution mode (incident neutron wavelength 1.36 A) at various temperature upto 900°C. CarefuI inspection of temperature dependence of diffraction data showed appearance and disappearance of certain Bragg's reflections above 1273 K. It is a signature of structural phase transition. Rietveld refinement of the powder diffraction data revealed that diffraction patterns at and above 800°C could be indexed using the monoclinic structure with P21/n space group. Detail analysis for identify the water molecules is under investigation. (author)

  16. High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

    International Nuclear Information System (INIS)

    Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.

    2011-01-01

    Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240 Pu . On the other hand, identification of shielded uranium requires active methods using neutron or photon sources . Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials . In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers . Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1x10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2x10 4 n/cm 2 s.

  17. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  18. Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10); Caracterização de campos de nêutrons térmicos para a calibração de monitores de nêutrons em termos da grandeza equivalente de dose ambiente H⁎(10)

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W. [Instituto de Radioproteção e Dosimetria (LNMRI/IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratório Nacional de Metrologia das Radiações Ionizantes; Astuto, Achilles, E-mail: larissapaizante@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    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 {sup 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{sup 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.

  19. Radiation damage in molybdenum and tungsten in high neutron fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Veljkovic, S; Milasin, N [Institute of Nuclear Sciences Boris Kidric, Department of Reactor Materials, Vinca, Beograd (Serbia and Montenegro)

    1964-04-15

    The effects of radiation on molybdenum and tungsten in high neutron fluxes are presented. The changes induced, particularly defects with a high migration activation energy, are analyzed. The correlation of these changes with the basic concepts of radiation damage in solids is considered. An attempt is made to relate the defects studied with the changes in macroscopic properties (author)

  20. Radiation damage in molybdenum and tungsten in high neutron fluxes

    International Nuclear Information System (INIS)

    Veljkovic, S.; Milasin, N.

    1964-01-01

    The effects of radiation on molybdenum and tungsten in high neutron fluxes are presented. The changes induced, particularly defects with a high migration activation energy, are analyzed. The correlation of these changes with the basic concepts of radiation damage in solids is considered. An attempt is made to relate the defects studied with the changes in macroscopic properties (author)

  1. Methods and applications in high flux neutron imaging

    International Nuclear Information System (INIS)

    Ballhausen, H.

    2007-01-01

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

  2. 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.

  3. High dose rate endobronchial brachytherapy - treatment technique

    International Nuclear Information System (INIS)

    Carvalho, Heloisa de Andrade; Aisen, Salim; Haddad, Cecilia Maria Kalil; Nadalin, Wladimir; Pedreira Junior, Wilson Leite; Chavantes, Maria Cristina

    1998-01-01

    High dose rate endobronchial brachytherapy is efficient in symptom relief due to obstructive endobronchial malignancies. However, it's role in survival improvement for patients with lung cancer is not yet established. The use of this treatment in increasing, specially in the developing countries. The purpose of this paper is to present the treatment technique used in the Radiotherapy Department of the Hospital da Clinicas, University of Sao Paulo, based on an experience of 60 cases treated with 180 procedures. Some practical suggestions and rules adopted in the Department are described. The severe complications rate is 6.7%, demonstrating an adequate patient selection associated with the technique utilized. (author)

  4. Neuroprotective potential of high-dose biotin.

    Science.gov (United States)

    McCarty, Mark F; DiNicolantonio, James J

    2017-11-01

    A recent controlled trial has established that high-dose biotin supplementation - 100 mg, three times daily - has a stabilizing effect on progression of multiple sclerosis (MS). Although this effect has been attributed to an optimization of biotin's essential cofactor role in the brain, a case can be made that direct stimulation of soluble guanylate cyclase (sGC) by pharmacological concentrations of biotin plays a key role in this regard. The utility of high-dose biotin in MS might reflect an anti-inflammatory effect of cGMP on the cerebral microvasculature, as well on oligodendrocyte differentiation and on Schwann cell production of neurotrophic factors thought to have potential for managing MS. But biotin's ability to boost cGMP synthesis in the brain may have broader neuroprotective potential. In many types of neurons and neural cells, cGMP exerts neurotrophic-mimetic effects - entailing activation of the PI3K-Akt and Ras-ERK pathways - that promote neuron survival and plasticity. Hippocampal long term potentiation requires nitric oxide synthesis, which in turn promotes an activating phosphorylation of CREB via a pathway involving cGMP and protein kinase G (PKG). In Alzheimer's disease (AD), amyloid beta suppresses this mechanism by inhibiting sGC activity; agents which exert a countervailing effect by boosting cGMP levels tend to restore effective long-term potentiation in rodent models of AD. Moreover, NO/cGMP suppresses amyloid beta production within the brain by inhibiting expression of amyloid precursor protein and BACE1. In conjunction with cGMP's ability to oppose neuron apoptosis, these effects suggest that high-dose biotin might have potential for the prevention and management of AD. cGMP also promotes neurogenesis, and may lessen stroke risk by impeding atherogenesis and hypertrophic remodeling in the cerebral vasculature. The neuroprotective potential of high-dose biotin likely could be boosted by concurrent administration of brain

  5. High temperature superconductors for fusion magnets -influence of neutron irradiation

    International Nuclear Information System (INIS)

    Chudy, M.; Eisterer, M.; Weber, H. W.

    2010-01-01

    In this work authors present the results of study of influence of neutron irradiation of high temperature superconductors for fusion magnets. High temperature superconductors (type of YBCO (Yttrium-Barium-Copper-Oxygen)) are strong candidates to be applied in the next step of fusion devices. Defects induced by fast neutrons are effective pinning centres, which can significantly improve critical current densities and reduce J c anisotropy. Due to induced lattice disorder, T c is reduced. Requirements for ITER (DEMO) are partially achieved at 64 K.

  6. High-pressure cell for simultaneous dielectric and neutron spectroscopy

    DEFF Research Database (Denmark)

    Sanz, Alejandro; Hansen, Henriette Wase; Jakobsen, Bo

    2018-01-01

    In this article, we report on the design, manufacture, and testing of a high-pressure cell for simultaneous dielectric and neutron spectroscopy. This cell is a unique tool for studying dynamics on different time scales, from kilo- to picoseconds, covering universal features such as the α relaxation......, a cylindrical capacitor is positioned within the bore of the high-pressure container. The capacitor consists of two concentric electrodes separated by insulating spacers. The performance of this setup has been successfully verified by collecting simultaneous dielectric and neutron spectroscopy data...

  7. Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

    International Nuclear Information System (INIS)

    Ku, L.; Kolibal, J.G.

    1982-06-01

    The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented

  8. The relative biological effectiveness of fractionated doses of fast neutrons (42 MeVd→Be) for normal tissues. Pt. 3

    International Nuclear Information System (INIS)

    Rezvani, M.; Hopewell, J.W.; Robbins, M.E.C.; Hamlet, R.; Barnes, D.W.H.; Sansom, J.M.; Adams, P.J.V.

    1990-01-01

    The effect of single and fractionated doses of fast neutrons (42 MeV d→Bc ) on the early and late radiation responses of the pig lung have been assessed by the measurement of changes in lung function using a 133 Xe washout technique. The results obtained for irradiation schedules with fast neutrons have been compared with those after photon irradiation. There was no statistically significant difference between the values for the relative biological effectiveness (RBE) for the early and late radiation response of the lung. The RBE of the neutron beam increased with decreasing size of dose/fraction with an upper limit value of 4.39 ± 0.94 for infinitely small X-ray doses per fraction. (author)

  9. Fusion materials high energy-neutron studies. A status report

    International Nuclear Information System (INIS)

    Doran, D.G.; Guinan, M.W.

    1980-01-01

    The objectives of this paper are (1) to provide background information on the US Magnetic Fusion Reactor Materials Program, (2) to provide a framework for evaluating nuclear data needs associated with high energy neutron irradiations, and (3) to show the current status of relevant high energy neutron studies. Since the last symposium, the greatest strides in cross section development have been taken in those areas providing FMIT design data, e.g., source description, shielding, and activation. In addition, many dosimetry cross sections have been tentatively extrapolated to 40 MeV and integral testing begun. Extensive total helium measurements have been made in a variety of neutron spectra. Additional calculations are needed to assist in determining energy dependent cross sections

  10. Response function measurement of plastic scintillator for high energy neutrons

    International Nuclear Information System (INIS)

    Sanami, Toshiya; Ban, Syuichi; Takahashi, Kazutoshi; Takada, Masashi

    2003-01-01

    The response function and detection efficiency of 2''φ x 2''L plastic (PilotU) and NE213 liquid (2''NE213) scintillators, which were used for the measurement of secondary neutrons from high energy electron induced reactions, were measured at Heavy Ion Medical Accelerator in Chiba (HIMAC). High energy neutrons were produced via 400 MeV/n C beam bombardment on a thick graphite target. The detectors were placed at 15 deg with respect to C beam axis, 5 m away from the target. As standard, a 5''φ x 5''L NE213 liquid scintillator (5''NE213) was also placed at same position. Neutron energy was determined by the time-of-flight method with the beam pickup scintillator in front of the target. In front of the detectors, veto scintillators were placed to remove charged particle events. All detector signals were corrected with list mode event by event. We deduce neutron spectrum for each detectors. The efficiency curves for pilotU and 2''NE213 were determined on the bases of 5 N E213 neutron spectrum and its efficiency calculated by CECIL code. (author)

  11. Neutron scattering at the high-flux isotope reactor

    International Nuclear Information System (INIS)

    Cable, J.W. Chakoumakos, B.C.; Dai, P.

    1995-01-01

    The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

  12. 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.

  13. A parametric model to describe neutron spectra around high-energy electron accelerators and its application in neutron spectrometry with Bonner Spheres

    Science.gov (United States)

    Bedogni, Roberto; Pelliccioni, Maurizio; Esposito, Adolfo

    2010-03-01

    Due to the increased interest of the scientific community in the applications of synchrotron light, there is an increasing demand of high-energy electron facilities, testified by the construction of several new facilities worldwide. The radiation protection around such facilities requires accurate experimental methods to determine the dose due to prompt radiation fields. Neutron fields, in particular, are the most complex to measure, because they extend in energy from thermal (10 -8 MeV) up to hundreds MeV and because the responses of dosemeters and survey meters usually have large energy dependence. The Bonner Spheres Spectrometer (BSS) is in practice the only instrument able to respond over the whole energy range of interest, and for this reason it is frequently used to derive neutron spectra and dosimetric quantities in accelerator workplaces. Nevertheless, complex unfolding algorithms are needed to derive the neutron spectra from the experimental BSS data. This paper presents a parametric model specially developed for the unfolding of the experimental data measured with BSS around high-energy electron accelerators. The work consists of the following stages: (1) Generation with the FLUKA code, of a set of neutron spectra representing the radiation environment around accelerators with different electron energies; (2) formulation of a parametric model able to describe these spectra, with particular attention to the high-energy component (>10 MeV), which may be responsible for a large part of the dose in workplaces; and (3) implementation of this model in an existing unfolding code.

  14. A parametric model to describe neutron spectra around high-energy electron accelerators and its application in neutron spectrometry with Bonner Spheres

    International Nuclear Information System (INIS)

    Bedogni, Roberto; Pelliccioni, Maurizio; Esposito, Adolfo

    2010-01-01

    Due to the increased interest of the scientific community in the applications of synchrotron light, there is an increasing demand of high-energy electron facilities, testified by the construction of several new facilities worldwide. The radiation protection around such facilities requires accurate experimental methods to determine the dose due to prompt radiation fields. Neutron fields, in particular, are the most complex to measure, because they extend in energy from thermal (10 -8 MeV) up to hundreds MeV and because the responses of dosemeters and survey meters usually have large energy dependence. The Bonner Spheres Spectrometer (BSS) is in practice the only instrument able to respond over the whole energy range of interest, and for this reason it is frequently used to derive neutron spectra and dosimetric quantities in accelerator workplaces. Nevertheless, complex unfolding algorithms are needed to derive the neutron spectra from the experimental BSS data. This paper presents a parametric model specially developed for the unfolding of the experimental data measured with BSS around high-energy electron accelerators. The work consists of the following stages: (1) Generation with the FLUKA code, of a set of neutron spectra representing the radiation environment around accelerators with different electron energies; (2) formulation of a parametric model able to describe these spectra, with particular attention to the high-energy component (>10 MeV), which may be responsible for a large part of the dose in workplaces; and (3) implementation of this model in an existing unfolding code.

  15. Performance of thermoluminescent materials for high dose dosimetry

    International Nuclear Information System (INIS)

    Texeira, Maria I.; Cecatti, Sonia G.P.; Caldas, Linda V.E.

    2008-01-01

    Cases involving high-doses of ionizing radiation are becoming increasingly common.The objective of this work was to characterize thermoluminescent materials for the dosimetry of workers exposed to high doses. Samples of TLD-200, TLD-400 and TLD-800 pellets from Thermo Electron Corporation were studied in gamma high-doses. Dose-response curves were obtained for doses between 100 mGy and 100 Gy. The reproducibility, the lower detection limits and dose-response curves were obtained for all three materials. The different kinds of detectors show usefulness for dosimetry of workers exposed accidentally to high doses. (author)

  16. Neutron resonance transmission spectroscopy with high spatial and energy resolution at the J-PARC pulsed neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Shinohara, T.; Kai, T.; Ooi, M. [Japan Atomic Energy Agency, 2–4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kamiyama, T.; Kiyanagi, Y.; Shiota, Y. [Hokkaido University, Kita 13 Nishi 8 Kita-ku, Sapporo-shi, Hokkaido 060-8628 (Japan); McPhate, J.B.; Vallerga, J.V.; Siegmund, O.H.W. [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

    2014-05-11

    The sharp variation of neutron attenuation at certain energies specific to particular nuclides (the lower range being from ∼1 eV up to ∼1 keV), can be exploited for the remote mapping of element and/or isotope distributions, as well as temperature probing, within relatively thick samples. Intense pulsed neutron beam-lines at spallation sources combined with a high spatial, high-timing resolution neutron counting detector, provide a unique opportunity to measure neutron transmission spectra through the time-of-flight technique. We present the results of experiments where spatially resolved neutron resonances were measured, at energies up to 50 keV. These experiments were performed with the intense flux low background NOBORU neutron beamline at the J-PARC neutron source and the high timing resolution (∼20 ns at epithermal neutron energies) and spatial resolution (∼55 µm) neutron counting detector using microchannel plates coupled to a Timepix electronic readout. Simultaneous element-specific imaging was carried out for several materials, at a spatial resolution of ∼150 µm. The high timing resolution of our detector combined with the low background beamline, also enabled characterization of the neutron pulse itself – specifically its pulse width, which varies with neutron energy. The results of our measurements are in good agreement with the predicted results for the double pulse structure of the J-PARC facility, which provides two 100 ns-wide proton pulses separated by 600 ns, broadened by the neutron energy moderation process. Thermal neutron radiography can be conducted simultaneously with resonance transmission spectroscopy, and can reveal the internal structure of the samples. The transmission spectra measured in our experiments demonstrate the feasibility of mapping elemental distributions using this non-destructive technique, for those elements (and in certain cases, specific isotopes), which have resonance energies below a few keV, and with lower

  17. Immune reactivity after high-dose irradiation

    International Nuclear Information System (INIS)

    Gassmann, W.; Wottge, H.U.; von Kolzynski, M.; Mueller-Ruchholtz, W.

    1986-01-01

    Immune reactivity after total-body irradiation was investigated in rats using skin graft rejection as the indicator system. After sublethal irradiation with 10.5 Gy (approximately 50% lethality/6 weeks) the rejection of major histocompatibility complex allogeneic skin grafts was delayed significantly compared with nonirradiated control animals (28 versus 6.5 days). In contrast, skin grafts were rejected after 7.5 days in sublethally irradiated animals and 7 days in lethally irradiated animals if additional skin donor type alloantigens--namely, irradiated bone marrow cells--were given i.v. either simultaneously or with a delay of not more than 24 hr after the above conditioning regimen. These reactions were alloantigen-specific. They were observed in six different strain combinations with varying donors and recipients. Starting on day 2 after irradiation, i.v. injection of bone marrow gradually lost its effectivity and skin grafts were no longer rejected with uniform rapidity; skin donor marrow given on days 4 or 8 did not accelerate skin graft rejection at all. These data show that for approximately 1-2 days after high-dose total-body irradiation rats are still capable of starting a vigorous immune reaction against i.v.-injected alloantigens. The phenomenon of impaired rejection of skin grafted immediately after high-dose irradiation appears to result from the poor accessibility of skin graft alloantigens during the early postirradiation phase when vascularization of the grafted skin is insufficient

  18. High-dose vitamin C and cancer

    Directory of Open Access Journals (Sweden)

    Ahmet Unlu, M.D.

    2016-01-01

    Full Text Available Vitamin C (ascorbic acid, ascorbate is a basic compound that is of great importance with its role in various enzymatic reactions including the synthesis of collagen, as well as with its redox functions. Vitamin C has become the center of interest in cancer studies, in consequence of the facts that connective tissue changes and vitamin C deficiency were first alleged to be associated with cancer in the 1950s; and that high doses of vitamin C was asserted to be cytotoxic for cancer cells, later on. The results of the first study carried out in the 1970s were promising; but afterwards, the studies were ascertained to be faulty. Despite the positive results achieved from some laboratory and animal experiments, randomized clinical trials did not verify those findings, and no clear benefit of vitamin C for cancer treatment could be demonstrated. As for studies, where its use in combination with other cancer treatment regimens was assessed, conflicting results were obtained. Although intake of high doses of vitamin C is alleged to be harmless, based on that it is in the group of water soluble vitamins and is not stored in the body, there are many side effects and drug interactions reported in the literature. For now, it is better to abstain from this treatment, until the benefit of the treatment (if any is clearly demonstrated, considering the potential side effects and interactions.

  19. Electron scattering from high-momentum neutrons in deuterium

    International Nuclear Information System (INIS)

    Klimenko, A.V.; Kuhn, S.E.; Bueltmann, S.; Careccia, S.L.; Dharmawardane, K.V.; Dodge, G.E.; Guler, N.; Hyde-Wright, C.E.; Klein, A.; Tkachenko, S.; Weinstein, L.B.; Zhang, J.; Butuceanu, C.; Griffioen, K.A.; Baillie, N.; Fersch, R.G.; Funsten, H.; Egiyan, K.S.; Asryan, G.; Dashyan, N.B.

    2006-01-01

    We report results from an experiment measuring the semiinclusive reaction 2 H(e,e ' p s ) in which the proton p s is moving at a large angle relative to the momentum transfer. If we assume that the proton was a spectator to the reaction taking place on the neutron in deuterium, the initial state of that neutron can be inferred. This method, known as spectator tagging, can be used to study electron scattering from high-momentum (off-shell) neutrons in deuterium. The data were taken with a 5.765 GeV electron beam on a deuterium target in Jefferson Laboratory's Hall B, using the CEBAF large acceptance spectrometer. A reduced cross section was extracted for different values of final state missing mass W*, backward proton momentum p → s , and momentum transfer Q 2 . The data are compared to a simple plane wave impulse approximation (PWIA) spectator model. A strong enhancement in the data observed at transverse kinematics is not reproduced by the PWIA model. This enhancement can likely be associated with the contribution of final state interactions (FSI) that were not incorporated into the model. Within the framework of the simple spectator model, a 'bound neutron structure function' F 2n eff was extracted as a function of W* and the scaling variable x* at extreme backward kinematics, where the effects of FSI appear to be smaller. For p s >0.4 GeV/c, where the neutron is far off-shell, the model overestimates the value of F 2n eff in the region of x* between 0.25 and 0.6. A dependence of the bound neutron structure function on the neutron's 'off-shell-ness' is one possible effect that can cause the observed deviation

  20. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B N; Edwards, D J; Horsewell, A; Toft, P

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

  1. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    International Nuclear Information System (INIS)

    Singh, B.N.; Edwards, D.J.; Horsewell, A.; Toft, P.

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10 17 n/m 2 s (E > 1 MeV, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

  2. High energy neutron source for materials research and development

    International Nuclear Information System (INIS)

    Odera, M.

    1989-01-01

    Requirements for neutron source for nuclear materials research are reviewed and ESNIT, Energy Selective Neutron Irradiation Test facility proposed by JAERI is discussed. Its principal aims of a wide neutron energy tunability and spectra peaking at each energy to enable characterization of material damage process are demanding but attractive goals which deserve detailed study. It is also to be noted that the requirements make a difference in facility design from those of FMIT, IFMIF and other high energy intense neutron sources built or planned to date. Areas of technologies to be addressed to realize the ESNIT facility are defined and discussed. In order to get neutron source having desired spectral characteristics keeping moderate intensity, projectile and target combinations must be examined including experimentation if necessary. It is also desired to minimize change of flux density and energy spectrum according to location inside irradiation chamber. Extended target or multiple targets configuration might be a solution as well as specimen rotation and choice of combination of projectile and target which has minimum velocity of the center of mass. Though relevant accelerator technology exists, it is to be stressed that considerable efforts must be paid, especially in the area of target and irradiation devices to get ESNIT goal. Design considerations to allow hands-on maintenance and future upgrading possibility are important either, in order to exploit the facility fully for nuclear materials research and development. (author)

  3. High-pressure 3He gas scintillation neutron spectrometer

    International Nuclear Information System (INIS)

    Derzon, M.S.; Slaughter, D.R.; Prussin, S.G.

    1985-10-01

    A high-pressure, 3 He-Xe gas scintillation spectrometer has been developed for neutron spectroscopy on D-D fusion plasmas. The spectrometer exhibits an energy resolution of (121 +- 20 keV) keV (FWHM) at 2.5 MeV and an efficiency of (1.9 +- 0.4) x 10 -3 (n/cm 2 ) -1 . The contribution to the resolution (FWHM) from counting statistics is only (22 +- 3 keV) and the remainder is due predominantly to the variation of light collection efficiency with location of neutron events within the active volume of the detector

  4. High-resolution neutron-diffraction measurements to 8 kbar

    Science.gov (United States)

    Bull, C. L.; Fortes, A. D.; Ridley, C. J.; Wood, I. G.; Dobson, D. P.; Funnell, N. P.; Gibbs, A. S.; Goodway, C. M.; Sadykov, R.; Knight, K. S.

    2017-10-01

    We describe the capability to measure high-resolution neutron powder diffraction data to a pressure of at least 8 kbar. We have used the HRPD instrument at the ISIS neutron source and a piston-cylinder design of pressure cell machined from a null-scattering titanium zirconium alloy. Data were collected under hydrostatic conditions from an elpasolite perovskite La?NiMnO?; by virtue of a thinner cell wall on the incident-beam side of the cell, it was possible to obtain data in the instrument's highest resolution back-scattering detector banks up to a maximum pressure of 8.5 kbar.

  5. High precision neutron interferometer setup S18b

    International Nuclear Information System (INIS)

    Hasegawa, Y.; Lemmel, H.

    2011-01-01

    The present setup at S18 is a multi purpose instrument. It is used for both interferometry and a Bonse-Hart camera for USANS (Ultra Small Angle Neutron Scattering) spectroscopy with wide range tunability of wavelength. Some recent measurements demand higher stability of the instrument, which made us to propose a new setup dedicated particularly for neutron interferometer experiments requiring high phase stability. To keep both options available, we suggest building the new setup in addition to the old one. By extending the space of the present setup by 1.5 m to the upstream, both setups can be accommodated side by side. (authors)

  6. Self-powered neutron detector of high sensitivity

    International Nuclear Information System (INIS)

    Brixy, H.; Spillekothen, H.G.; Benninghofen, G.; Serafin, N.

    1983-01-01

    A self-powered neutron detector is proposed, consisting of three concentrically arranged electrically conducting tubes; where the central one forms the emitter and the inner and outer ones form the collector and where the tubes are electrically insulated from each other by insulating material. The emitter consists of a material with a high absorption cross-section for thermal neutrons, particularly of gadolinium, and is provided with an auxiliary emitter layer on the inside or the outside. With suitable dimensions and material, the auxiliary emitter layer increases the yield of electrons. (orig./HP) [de

  7. Neutron analysis of the fuel of high temperature nuclear reactors

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2014-10-01

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  8. Properties of neutron-rich hafnium high-spin isomers

    CERN Multimedia

    Tungate, G; Walker, P M; Neyens, G; Billowes, J; Flanagan, K; Koester, U H; Litvinov, Y

    It is proposed to study highly-excited multi-quasiparticle isomers in neutron-rich hafnium (Z=72) isotopes. Long half-lives have already been measured for such isomers in the storage ring at GSI, ensuring their accessibility with ISOL production. The present proposal focuses on:\\\\ (i) an on-line experiment to measure isomer properties in $^{183}$Hf and $^{184}$Hf, and\\\\ (ii) an off-line molecular breakup test using REXTRAP, to provide Hf$^{+}$ beams for future laser spectroscopy and greater sensitivity for the future study of more neutron-rich isotopes.

  9. Hair {sup 32}P measurement for body dose mapping in non-fatal exposures to fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Mianji, Fereidoun A. [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Iran Nuclear Regulatory Authority, Tehran (Iran, Islamic Republic of); Jafari, Sheyda; Zaryouni, Saiedeh [Bu-Ali Sina University, Faculty of Science, Hamedan (Iran, Islamic Republic of); Hajizadeh, Bardia [Iran Nuclear Regulatory Authority, Tehran (Iran, Islamic Republic of)

    2015-03-15

    Dosimetry bioassay methods are the backbone of a personal dosimetry in criticality accidents. Although methods like hair dosimetry and the use of activation foils (e.g., {sup 32}S) have been employed for decades, capabilities of different techniques, effects of hair type and neutron spectrum on the dose response, sensitivity and uncertainties of different techniques, etc., need more investigations. For this reason, the use of the {sup 32}S(n,p){sup 32}P reaction and hair samples for estimating non-fatal doses from fast neutrons was studied. The experiments were carried out with the hair samples attached on a RANDO phantom in a Cf-252 neutron field, in the dose range of about 0.05-1.15 Gy. In addition, the adequate post-accident preparation for hair samples including optimum conditioning and timing were investigated. Experimental results prove the good sensitivity and merit of the method for neutron quantification in the mentioned dose range for which other bioassay methods are of poor resolution and sensitivity. A rough estimation of the dose-response curve for Iranian hair was also derived. (orig.)

  10. Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

    International Nuclear Information System (INIS)

    Ueki, K.; Namito, Y.; Fuse, T.

    1986-01-01

    On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

  11. Estimation of dependence between mean of fractionation of photons and neutrons dose and intensity of post-irradiation reaction of mouse large intestine

    International Nuclear Information System (INIS)

    Gasinska, A.

    1995-01-01

    neutrons F r was 3% for early- and 16% for late - measured biological effects. The differences in effectivity of radiation with various LPE effect confirm WSB values calculated on the base of α and β coefficients and LQ model. WSB values were higher and more dependent on amount of fractionated dose for late effects. WSB for late reaction and low fraction dose (0.25 Gy of photons) was 9.6 and for early irradiation effects was 6.3. On the level of high fraction dose (25 Gy of photons) there was inverted dependence. WSB for early effects was higher (2.5) in comparison with late irradiation effects (1.8). Application of mathematical model for measuring of quantity of units responsible for intestine regeneration (tissue rescuing units - TRU) showed various K number for early reaction and do not confirmed big radiosensitivity of this tissue. Also for late response low radiosensitivity was confirmed. Histological investigation showed that target cells for early reaction are epithelial crypt cells, however for late reaction there are fibrocytis and myocytes

  12. Measurement of the relationship of 24 Na activity and the received neutron dose

    International Nuclear Information System (INIS)

    Gossio, S.; Carrelli, J.; Villella, A.; Soppe, E.

    2013-01-01

    In cases of criticality accidents it is required a fast dosimetric system that allows to evaluate the doses of the personnel involved. The reaction (n,y) with sodium presented in the body ( 23 Na), generates 24 Na, that emits two gamma of 1369KeV and 2754 KeV that can be measured using a whole body counter. The experienced were carried out with the irradiation of 252 Cf of a phantom with a solution of NaCl in water. After the irradiation it was measured the 24 Na activity in the whole body counter, which has a HPGe detector previously calibrated in energy and efficiency. Considering the correction by decay, the quantity of 23 Na presented in the body of an adult and elimination curve of 24 Na, it was established a coefficient of neutronic doses by unity of activity of 24 Na measured in the whole body counter. This method is useful for the retrospective estimation of the doses, as well as to carry out a radiological sorting in case of criticality

  13. Dynamics of elements in soil treated with increasing doses sewage sludge for instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Oliveira, Helder de; Mortatti, Jefferson; Vendramini, Diego; Lopes, Renato A.; Nolasco, Murilo M.; Sarries, Gabriel A.; Furlan, Adriana

    2007-01-01

    In this work the dynamics of the elements was analyzed The, Br, Ce, Co, Cr, Cs, Fe, Hf, La, In the, Sb, Sc, Sm, Ta, Th, U, Yb and Zn in a profile of a red-yellow latossolo, in the depths of 0-5, 5-10, 10-30 and 30-50 cm, and dose of the biosolid of 0, 25, 124 and 375 t ha -1 , of the station of treatment of sewer of Barueri, Sao Paulo. The experiment was carried out in areas of 3,05 m 2 in the times of 2,2; 4,0; 6,6; 14,3 and 21 months. For analysis of the elementary composition, it was used of the analysis technique by instrumental neutron activation analysis (INAA). The experiment was submitted under normal tropical conditions in a forest station in Itatinga, Sao Paulo, of the University of Sao Paulo. For better details, the factors depth, doses and times statistical analyses of the results of the elementary composition of the soil samples were made. For all the biossolid doses conditioned with polymeric and applied in the soil, the composition of 17 of the 18 elements in the soil were not altered, with exception for Cr in the studied times. The elements As, Br, Ce, Co, Fe, Hf, La, Sm, Ta, Th, U and Yb presented higher levels in the deepest layers of soil; already the elements Cr, In the, Sb and Zn presented higher concentrations in the superficial layers. (author)

  14. Neutron spin echo and high resolution inelastic spectroscopy

    International Nuclear Information System (INIS)

    Mezei, F.; Hungarian Academy of Sciences, Budapest. Central Research Inst. for Physics)

    1982-01-01

    The principles of neutrons spin echo (NSE) technique are considered. It is shown that the basis of NSE principle is a single step measurement of the change of the neutron velocity in the scattering process. The backscattering soectroscopy and the NSE techniques are compared. The NSF spectrometer is described. It is shown that 0.5 MeV energy resolution achieved in the NSE experiment is about 40 times superior to those achieved by the other techniques. The NSE technique has the unique feature that provides high resolution in neutron energy change independently of the monochromatization of the beam. The NSE instrument not only covers a wider dynamic range on a pulsed source that on a continuous one, but also collects data more efficiently

  15. Genetic injury in hybrid male mice exposed to low doses of 60CO γ-rays or fission neutrons. 1

    International Nuclear Information System (INIS)

    Grahn, D.; Carnes, B.A.; Farrington, B.H.; Lee, C.H.

    1984-01-01

    Young adult male B6CF 1 mice were exposed to single whole body doses of fission neutrons or 60 Co γ rays. Postspermatogonial dominant lethal injury, incidence of reciprocal chromosome translocations induced in spermatogonia, incidence of abnormal epididymal sperm 4-6 weeks after exposure, and testis weight loss 3-6 weeks after exposure were all measured. Significant effects were seen at 1 and 2.5 rad of neutrons consistent with extrapolation from higher doses, with the exception of dominant lethal mutations, which occurred in significant excess of expectation. Dose-response functions were linear or linear-quadratic, depending upon end point, radiation quality, and dose range. For translocation frequencies, the D 2 term was negative for neutron and positive for γ-ray irradiations. RBE values varied with dose and end point. For testis weight loss and abnormal sperm over the full dose range, the RBEs were between 5 and 6. They were between 7 and 9 at lower doses (< 10 rad) for translocations. RBEs for postimplantation and total dominant lethal rates were 5-6 above 10 rad and 10-14 below 10 rad. The RBEs for preimplant losses were between 15 and 25 above 10 rad and possibly higher below 10 rad, although the data are statistically 'noisy'. (Auth.)

  16. Intense neutron source: high-voltage power supply specifications

    International Nuclear Information System (INIS)

    Riedel, A.A.

    1980-08-01

    This report explains the need for and sets forth the electrical, mechanical and safety specifications for a high-voltage power supply to be used with the intense neutron source. It contains sufficient information for a supplier to bid on such a power supply

  17. Nuclear radiation detectors using high resistivity neutron transmutation doped silicon

    International Nuclear Information System (INIS)

    Gessner, T.; Irmer, K.

    1983-01-01

    A method for the production of semiconductor detectors based on high resistivity n-type silicon is described. The n-type silicon is produced by neutron irradiation of p-type silicon. The detectors are produced by planar technique. They are suitable for the spectrometry of alpha particles and for the pulse count measurement of beta particles at room temperature. (author)

  18. The study of high precision neutron moisture gauge

    International Nuclear Information System (INIS)

    Liu Shengkang; Bao Guanxiong; Sang Hai; Zhu Yuzhen

    1993-01-01

    The principle, structure and calibration experiment of the high precision neutron moisture gauge (insertion type) are described. The gauge has been appraised. The precision of the measuring moisture of coke is lower than 0.5%, and the range of the measuring moisture is 2%-12%. The economic benefit of the gauge application is good

  19. Investigation of Acrylic Acid at High Pressure using Neutron Diffraction

    DEFF Research Database (Denmark)

    Johnston, Blair F.; Marshall, William G.; Parsons, Simon

    2014-01-01

    This article details the exploration of perdeuterated acrylic acid at high pressure using neutron diffraction. The structural changes that occur in acrylic acid-d4 are followed via diffraction and rationalised using the Pixel method. Acrylic acid undergoes a reconstructive phase transition to a new...

  20. Investigation of Methacrylic Acid at High Pressure Using Neutron Diffraction

    DEFF Research Database (Denmark)

    Marshall, William G.; Urquhart, Andrew; Oswald, Iain D. H.

    2015-01-01

    This article shows that pressure can be a low-intensity route to the synthesis of polymethacrylic acid. The exploration of perdeuterated methacrylic acid at high pressure using neutron diffraction reveals that methacrylic acid exhibits two polymorphic phase transformations at relatively low...

  1. RBE/absorbed dose relationship of d(50)-Be neutrons determined for early intestinal tolerance in mice

    International Nuclear Information System (INIS)

    Gueulette, J.; Wambersie, A.

    1978-01-01

    RBE/absorbed dose relationship of d(50)-Be neutrons (ref.: 60 Co) was determined using intestinal tolerance in mice (LD50) after single and fractionated irradiation. RBE is 1.8 for a single fraction (about 1000 rad 60 Co dose); it increases when decreasing dose and reaches the plateau value of 2.8 for a 60 Co dose of about 200 rad. This RBE value is used for the clinical applications with the cyclotron 'Cyclone' at Louvain-la-Neuve [fr

  2. High dose radiation damage in nuclear energy structural materials investigated by heavy ion irradiation simulation

    International Nuclear Information System (INIS)

    Zheng Yongnan; Xu Yongjun; Yuan Daqing

    2014-01-01

    Structural materials in ITER, ADS and fast reactor suffer high dose irradiations of neutrons and/or protons, that leads to severe displacement damage up to lOO dpa per year. Investigation of radiation damage induced by such a high dose irradiation has attracted great attention along with the development of nuclear energy facilities of new generation. However, it is deeply hampered for the lacking of high dose neutron and proton sources. Irradiation simulation of heavy ions produced by accelerators opens up an effective way for laboratory investigation of high dose irradiation induced radiation damage encountered in the ITER, ADS, etc. Radiation damage is caused mainly by atomic displacement in materials. The displacement rate of heavy ions is about lO 3 ∼10 7 orders higher than those of neutrons and protons. High displacement rate of heavy ions significantly reduces the irradiation time. The heavy ion irradiation simulation technique (HIIS) technique has been developed at China Institute of Atomic Energy and a series of the HIIS experiments have been performed to investigate radiation damage in stainless steels, tungsten and tantalum at irradiation temperatures from room temperature to 800 ℃ and in the irradiation dose region up to 100 dpa. The experimental results show that he radiation swelling peak for the modified stainless steel appears in the temperature region around 580 ℃ and the radiation damage is more sensitive to the temperature, the size of the radiation induced vacancy cluster or void increase with the increasing of the irradiation dose, and among the three materials the home-made modified stainless steel has the best radiation resistant property. (authors)

  3. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  4. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  5. Quality control of 192Ir high dose rate after loading brachytherapy dose veracity

    International Nuclear Information System (INIS)

    Feng Zhongsu; Xu Xiao; Liu Fen

    2008-01-01

    Recently, 192 Ir high dose rate (HDR) afterloading are widely used in brachytherapy. The advantage of using HDR systems over low dose rate systems are shorter treatment time and higher fraction dose. To guarantee the veracity of the delivery dose, several quality control methods are deseribed in this work. With these we can improve the position precision, time precision and dose precision of the brachytherapy. (authors)

  6. Quick evaluation of the neutron dose following a criticality accident by measurement of sodium 24 activity

    International Nuclear Information System (INIS)

    Tabardel, R.; Ricourt, A.; Parmentier, N.

    1984-07-01

    In order to quickly sort out the irradiated individuals following a criticality accident, the neutron dose can be evaluated quickly by measuring the sodium-24 activity induced in the human body. The report supplies the information necessary for this evaluation from the response of various detectors of current use in radiation protection. The first part describes the method of evaluation of sodium-24 activity (A) given by the reading (M) of each instrument. The second part describes the method of kerma evaluation from the measured sodium-24 activity. The third part is an experimental application of the method of kerma evaluation from the sodium-24 activity measured in a phantom irradiated in the SILENE reactor flux. The results given by radiation protection instruments are in good agreement with the calculated values for a front exposure and demonstrate the usefulness of measuring the induced sodium-24 activity by radiation protection instruments of current use [fr

  7. Dose conversion coefficients for neutron exposure to the lens of the human eye

    International Nuclear Information System (INIS)

    Manger, Ryan P.; Bellamy, Michael B.; Eckerman, Keith F.

    2011-01-01

    Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10 -9 to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

  8. Dose conversion coefficients for neutron exposure to the lens of the human eye

    International Nuclear Information System (INIS)

    Manger, R. P.; Bellamy, M. B.; Eckerman, K. F.

    2012-01-01

    Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10 -9 to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations. (authors)

  9. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1993-12-31

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

  10. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1994-12-31

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

  11. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    International Nuclear Information System (INIS)

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  12. Chemical dosimetry principles in high dose dosimetry

    International Nuclear Information System (INIS)

    Mhatre, Sachin G.V.

    2016-01-01

    In radiation processing, activities of principal concern are process validation and process control. The objective of such formalized procedures is to establish documentary evidence that the irradiation process has achieved the desired results. The key element of such activities is inevitably a well characterized reliable dosimetry system that is traceable to recognized national and international dosimetry standards. Only such dosimetry systems can help establish the required documentary evidence. In addition, industrial radiation processing such as irradiation of foodstuffs and sterilization of health careproducts are both highly regulated, in particular with regard to dose. Besides, dosimetry is necessary for scaling up processes from the research level to the industrial level. Thus, accurate dosimetry is indispensable

  13. Onyx as radiation detector for high doses

    International Nuclear Information System (INIS)

    Teixeira, Maria Inês; Souza, Divanizia N.; Caldas, Linda V.E.

    2011-01-01

    A study of the thermoluminescent (TL) characteristics of white, black and stripped onyx samples is reported in this work. Onyx is a variety of chalcedony, a form of quartz. The onyx stone is considered nobler than marble. The irradiations were performed using a Gamma-Cell 220 system ( 60 Co). The TL emission curves presented two peaks around 150 °C and 210 °C for all samples. The dose–response curves showed a sublinear behavior between 0.5 Gy and 5 kGy, and the lower detection limit for the white onyx pellets was 1.5 mGy. The main dosimetric characteristics were studied, and the material showed good performance for high dose dosimetry.

  14. The application of high dose food irradiation

    International Nuclear Information System (INIS)

    Bruyn, I. De

    1997-01-01

    During the 1950's to end 1970's the United States Army developed the basic methodology to produce shelf stable irradiated meat, seafood and poultry products. These products are normally packed without gravy, sauce or brine, as liquid is not required to sterilize the product as in the canning process. This leads to the distinctive 'dried cooked' taste normally associated with roasts opposed to the casserole taste usually associated with tinned meats. The meats are cooked, chilled, portioned, vacuum packed and irradiated to the required minimum dose of 25 to 45 kGy (depending on the product) at a temperature of between -20 and -40 Centigrade to ensure absolute sterility even under tropical conditions. The product is packaged in a high quality four layer laminate pouch and will therefore not rust or burst even under adverse weather conditions. The product can be guaranteed for more than two years as long as the integrity of the packaging is maintained. (Author)

  15. The application of high dose food irradiation

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