Neutron KERMA factors of human tissues

A program to calculate the neutron KERMA in human tissues has been developed. The program was developed in Mathcad and contains the neutron kerma factors of those elements that are present in different human tissues. Having the elemental composition of any human tissue the neutron kerma can be easily calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic. Neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large.

Artificial neural networks technology for neutron spectrometry and dosimetry.

Artificial Neural Network Technology has been applied to unfold neutron spectra and to calculate 13 dosimetric quantities using seven count rates from a Bonner Sphere Spectrometer with a (6)LiI(Eu). Two different networks, one for spectrometry and another for dosimetry, were designed. To train and test both networks, 177 neutron spectra from the IAEA compilation were utilised. Spectra were re-binned into 31 energy groups, and the dosimetric quantities were calculated using the MCNP code and the fluence-to-dose conversion coefficients from ICRP 74. Neutron spectra and UTA4 response matrix were used to calculate the expected count rates in the Bonner spectrometer. Spectra and H(10) of (239)PuBe and (241)AmBe were experimentally obtained and compared with those determined with the artificial neural networks.

Neutron fluence rate measurement using prompt gamma rays.

A gamma ray spectrometer, with a 3('') Ø X 3('') NaI(Tl) detector, with a moderator sphere has been utilised to measure the neutron fluence rate, with this value the H(10) was estimated. When a neutron is captured by the hydrogen-based moderator, a 2.22 MeV prompt gamma ray is produced. In a multichannel analyser the net area under the 2.22 MeV photopeak is proportional to the total neutron fluence rate. The features of this system were determined by a Monte Carlo study that includes 3-, 5- and 10-inches diameter, water and polyethylene moderators and a (239)Pu-Be source. The prompt gamma response was extended to monoenergetic neutron sources. To verify the response, a (239)Pu-Be source in combination with a 10('') polyethylene sphere having a gamma-ray spectrometer with NaI(Tl) was utilised to estimate the neutron fluence rate and the H(10). These results were compared with neutron fluence rate and H(10) obtained using a Bonner sphere spectrometer and with the H(10) measured using a neutron remmeter.

Water-extended polyester neutron shield for a 252Cf neutron source.

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. During calculations, (252)Cf and shielding were modelled and the neutron spectra as well as the H(10) were calculated in four sites. The calculation was extended to include a water shielding, the source in vacuum and in air. Besides neutron shielding characteristics, the Kerma in air due to gammas emitted by (252)Cf and due to capture gamma rays in the shielding were included.

Campos neutrónicos producidos por un acelerador lineal para radioterapia / Neutron fields produced by linear accelerator for radiotherapy

Mediante una serie de medidas y cálculos Monte Carlo se han determinado las características dosimétricas y los espectros de los fotoneutrones que se producen en torno a un acelerador lineal para radioterapia de 18 MV. Las medidas se realizaron con dosímetros termoluminiscentes TLD 600 y TLD 700 que se expusieron desnudos y emparedados con Cd, así como dentro de una esfera de parafina y dentro de esferas Bonner.

Measurements and Monte Carlo calculations has been utilized to determine the dosimetric features as well as the neutron spectra of photoneutrons produced around an 18 MV linear accelerator for radiotherapy. Measurementes were carried out with bare and Cd covered thermolumiscent dosimeters, TLD600 and TLD700, as well as inside a paraffine moderator. TLD pairs were also utilized as thermal neutrons inside a Bonner sphere spectrometer.

Artificial neural networks in neutron dosimetry.

An artificial neural network (ANN) has been designed to obtain neutron doses using only the count rates of a Bonner spheres spectrometer (BSS). Ambient, personal and effective neutron doses were included. One hundred and eighty-one neutron spectra were utilised to calculate the Bonner count rates and the neutron doses. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra, UTA4 response matrix and fluence-to-dose coefficients were used to calculate the count rates in the BSS and the doses. Count rates were used as input and the respective doses were used as output during neural network training. Training and testing were carried out in the MATLAB environment. The impact of uncertainties in BSS count rates upon the dose quantities calculated with the ANN was investigated by modifying by +/-5% the BSS count rates used in the training set. The use of ANNs in neutron dosimetry is an alternative procedure that overcomes the drawbacks associated with this ill-conditioned problem.

Response of human lymphocytes to low gamma ray doses

Linfocitos humanos fueron irradiados en un campo de radiación gamma de baja intensidad para determinar la expresión de las proteínas de choque calórico en función de la dosis. Los linfocitos fueron obtenidos de individuos cuyo trabajo los identifica como ocupacionalmente expuesto y no ocupacionalmente expuestos. La identidad de las proteínas se realizó utilizando anticuerpos contra las proteínas Hsp25, Hsp60, Hsp70 y Hsp90. De éstas, solamente la proteína hsp70 fue detectada antes y después de la irradiación. Los linfocitos del personal ocupacionalmente expuesto y no ocupacionalmente expuesto expresaron, antes y después de la irradiación, solamente la proteína Hsp70. La cantidad de proteína resultó directamente proporcional al tiempo de irradiación. Después de una dosis gamma de 70.5 mGy, los linfocitos del individuo ocupacionalmente expuesto expresaron una mayor cantidad de proteína Hsp70 que la expresada por los linfocitos del personal no ocupacionalmente expuesto. Este hecho es indicio de que el individuo ocupacionalmente expuesto tiene una mayor tolerancia a los rayos gamma (gamma-tolerancia), inducida por un proceso de adaptación generada por su condición laboral