Neutron emission and dose distribution from natural carbon irradiated with a 12 MeV amu-1 12C5+ ion beam.

Measured neutron energy distribution emitted from a thick stopping target of natural carbon at 0°, 30°, 60° and 90° from nuclear reactions caused by 12 MeV amu-1 incident 12C5+ ions were converted to energy differential and total neutron absorbed dose as well as ambient dose equivalent H *(10) using the fluence-to-dose conversion coefficients provided by the ICRP. Theoretical estimates were obtained using the Monte Carlo nuclear reaction model code PACE and a few existing empirical formulations for comparison. Results from the PACE code showed an underestimation of the high-energy part of energy differential dose distributions at forward angles whereas the empirical formulation by Clapier and Zaidins (1983 Nucl. Instrum. Methods 217 489-94) approximated the energy integrated angular distribution of H *(10) satisfactorily. Using the measured data, the neutron doses received by some vital human organs were estimated for anterior-posterior exposure. The estimated energy-averaged quality factors were found to vary for different organs from about 7 to about 13. Emitted neutrons having energies above 20 MeV were found to contribute about 20% of the total dose at 0° while at 90° the contribution was reduced to about 2%.

PHOTON FIELD OF ~100-200 KEV FOR ENVIRONMENTAL DOSEMETER CALIBRATION.

As a reference photon field, several radionuclides have been used frequently, such as 241Am,137Cs and60 Co for calibration. These nuclides provide mono-energy photons for dosemeters covering few tens of keV-MeV. The main energy around 200 keV is important for both environmental and medical fields since the former should consider scattering photons and the later should measure photons from X-ray generator. In our previous work, a backscattered layout can provide a uniform photon field spectra and dose rate with an energy of 190 keV by using an affordable intensity 137 Cs gamma source. Several other quasi-monoenergetic photon fields in the range of 100-200 keV could be obtained by using several available gamma sources. Two calibrated environmental CsI(Tl) survey meters, Horiba PA-1000 and Mr. Gamma A2700, had been measured with the developed backscattered photon field to understand energy-dependent features in order to confirm dosemeter readings. Consequently, both scintillator instruments are sensitive for measurements of the relatively low dose rates at 190 keV.

Response of plastic scintillator to gamma sources.

In this study, we developed a method for directly determining the energy deposited over the entire energy range by monitoring the light output from a plastic scintillator under gamma irradiation. The relative light output was analyzed based on Birks' semi-empirical formula for ionization to obtain the quenching parameter as kB = 0.016 ± 0.0004 g cm-2 MeV-1. Comparisons of experimental and calculated results for the light output spectra showed that considering the quenching effect, background subtraction, source casing, and energy sampling were essential for achieving good agreement.

Evaluation of absolute measurement using a 4π plastic scintillator for the 4πß-γ coincidence counting method.

Absolute measurement by the 4πß-γ coincidence counting method was conducted by two photomultipliers facing across a plastic scintillator to be focused on ß ray counting efficiency. The detector was held with a through-hole-type NaI(Tl) detector. The results include absolutely determined activity and its uncertainty especially about extrapolation. A comparison between the obtained and known activities showed agreement within their uncertainties.

Differential cross sections of neutron-induced fragment-emission reactions for a microdosimetry study.

Experimental differential cross sections of fragment emission (p, d, t, alpha, Li, Be and B), which were obtained for tens of mega electron volt neutrons on carbon and aluminum, using a counter telescope array and a Bragg-curve counter specially developed for neutron-induced reactions, are presented and compared with theoretical calculations using various reaction models. A calculation with the ISOBAR and GEM models was found to reproduce the experimental data except for an underestimation in non-vaporation processes. Calculations of the energy deposition by neutrons in a thin silicon layer show significant differences among the model employed.

Simulation technique for extrapolation curves in 4πß-γ coincidence counting method using EGS5 code.

A simulation technique was developed for the extrapolation technique in 4πß-γ coincidence counting method. Simultaneous emissions of ß and γ rays were calculated using EGS5 code to obtain coincidence counting between both ß and γ channels. The simulated extrapolation curves were compared with experimental data obtained with (134)Cs measurements using a plastic scintillator in the ß channel. The variation of the extrapolation curves with γ-gate configuration was investigated by the simulation technique.

Study of the neutron beam line shield design for JSNS.

The JSNS, a spallation neutron source of J-PARC (JAERI-KEK Joint Project of the High Intensity Proton Accelerator) has 23 neutron beam lines. In the present study, a database was formulated for an optimum shielding design using the MCNP-X code. The calculations involved two steps. In the first step, the neutron distributions were created in the typical neutron beam line with a model that included the spallation neutron source target. The neutron currents evaluated flowed from the duct into the duct wall which was the boundary source for the bulk shield surrounding the beam line. In the second step, bulk-shield calculations were performed for the various shielding materials (iron, concrete, heavy concrete and so on) used and their composites up to thicknesses of 3 m. The results were compared with each other. Composite material shields of iron and such hydrogeneous materials as polyethylene or concrete were more effective. A typical design was prepared for a beam line within 25 m distance from a moderator, as a sample.

Arrangement of high-energy neutron irradiation field and shielding experiment using 4 m concrete at KENS.

An irradiation field of high-energy neutrons produced in the forward direction from a thick tungsten target bombarded by 500 MeV protons was arranged at the KENS spallation neutron source facility. In this facility, shielding experiment was performed with an ordinary concrete shield of 4 m thickness assembled in the irradiation room, 2.5 m downstream from the target centre. Activation detectors of bismuth, aluminium, indium and gold were inserted into eight slots inside the shield and attenuations of neutron reaction rates were obtained by measurements of gamma-rays from the activation detectors. A MARS14 Monte Carlo simulation was also performed down to thermal energy, and comparisons between the calculations and measurements show agreements within a factor of 3. This neutron field is useful for studies of shielding, activation and radiation damage of materials for high-energy neutrons, and experimental data are useful to check the accuracies of the transmission and activation calculation codes.

Advantages and limitations of thermal and epithermal neutron activation analysis of bulk samples.

Systematic investigations have been carried out to extend the thermal neutron activation method for elemental analysis of bulk samples. A new method developed for the determination of the flux perturbation factor renders the thermal and epithermal neutron activation analyses of bulky samples of unknown compositions possible both in hydrogenous and in graphite moderators. The flux perturbation, F, depression, H, and self-absorption, G, factors are given for different samples. The limitation of the epithermal neutron activation analysis for hydrogenous bulky samples is also discussed.