Delayed neutron counting of uranium pins using moderated neutrons from portable DD neutron generator in high frequency mode.

Delayed neutron counting is often used to verify the characteristics of nuclear material. Use of portable neutron generators in high frequency pulsing mode enables effective analysis with higher counting efficiency and lower radiation protection demands. The paper deals with delayed neutron counting of uranium pins with P385 portable DD neutron generator in polyethylene-based setup. Counting is performed in high frequency mode of neutron generator. The viability of such measurement in the frequency range from 100 Hz to 250 Hz was demonstrated and optimal pulsing parameters for P385 neutron generator were found. Delayed neutron counting was then performed for two types of uranium rods. Delayed neutrons were counted both inbetween neutron pulses during neutron generation and once the emission of neutrons stopped. The results were further validated by Monte Carlo (MCNP) simulations. For the geometry of studied rods, the MCNP calculation were done to calculate the dependence of the response to delayed neutrons on rod enrichment, to show the viability to use the method for rod enrichment verification. An option of using cadmium inset in irradiation channel to overcome the effect of self-shielding for samples with higher enrichment was proposed, experimentally tested, and evaluated through MCNP calculations.

Comparison of several Li doped scintillators for measurement of neutron and γ radiation integral quantities.

Five types of scintillators for parallel detection of neutrons and γ were tested for their pulse shape discrimination ability. All detectors are based on 6Li (n,alpha)3H reaction. LiI, ZnS and polystyrene were used as scintillators. Tests were performed at different neutron and γ mixed fields (AmBe, Cf) using a small Hamamatsu photomultiplier and a Picoscope digitizer. A polystyrene-based ZnS + LiF detector was proven to be the most suitable for neutron-γ separation and similar sensitivity for both types of radiation to construct a single device to measure integral quantities neutrons and γ (fluence, ambient dose equivalent) in parallel. The ZnS + LiF detector based on plexiglass, is in principle useable also, but has low sensitivity to γ radiation. The tested Li glass or LiI crystal does not have the ability to separate neutron and γ. With amplitude discrimination it can be used as a neutron detector for a much simpler probe for a Bonner spectrometer.

NEUTRON DOSE ASSESSMENT USING SAMPLES OF HUMAN BLOOD AND HAIR.

The unique feature of nuclear accidents with neutron exposure is the induced radioactivity in body tissues. For dosimetry purposes, the most important stable isotopes occurring in human body, which can be activated by neutrons, are 23 Na and 32 S. The respective activation reactions are as follows:23Na(n,γ)24Na and32S(n,p)32P. While sodium occurs in human blood, sulfur is present in human hair. In order to verify the practical feasibility of this dosimetry technique in conditions of our laboratory, samples of human blood and hair were irradiated in a channel of a training reactor VR-1.24Na activity was measured by gamma-ray spectrometry.32P activity in hair was measured by means of a proportional counter. Based on neutron-spectrum calculation, relationships between neutron dose and induced activity were derived for both blood and hair.

On similarity of various reactor spectra and 235U prompt fission neutron spectrum.

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 - 6 MeV. Above this energy the reactor spectra becomes identical with the 235U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra.

DOSIMETRY WITH SALT IN MIXED RADIATION FIELDS OF PHOTONS AND NEUTRONS.

Salt (NaCl) represents a radiation sensitive material with a considerable potential for dosimetry in mixed radiation fields of photons and neutrons. In consequence of a gamma radiation exposure, it exhibits a strong luminescence signal following stimulation with blue light. Optically stimulated luminescence (OSL) technique can be used for measurement. Photon dose reconstruction can be done using so-called Single-Aliquot Regenerative-dose protocol. However, a part of OSL signal is caused by NaCl self-irradiation resulting from neutron reactions. This NaCl neutron sensitivity is comparable with neutron sensitivity of TLD-700. Neutron dose can be determined based on neutron activation of NaCl via reaction 23Na(n,γ)24Na. A relation between neutron dose and activity of 24Na can be derived. The total dose is thus determined based on the combination of results of OSL measurements and gamma activity measurements. Practical feasibility of this approach was experimentally verified for salt samples irradiated in a channel of a training reactor.

Equipment for neutron measurements at VR-1 Sparrow training reactor.

The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented.