REFERENCE NEUTRON FIELDS IN CALIBRATION LABORATORY; SIMPLE DOSIMETRIC PARAMETERS AND THEIR CHANGES IN TIME.

The calibration laboratory revises periodically the dosimetric properties of its calibration fields. For reference neutron fields based on radionuclide sources, it is due to presence of isotopic impurities in the source ingredients. Consequently, in long term, the admixtures and their decay products become significant. Neglecting the impact of such effect leads to inconsistencies between the neutron emission rate observed during the measurements and the emission rate derived from the decay curve of the main isotope. In the Radiation Protection Measurements Laboratory at the National Centre for Nuclear Research the neutron fields of the bare sources of 252Cf, 241AmBe and 239PuBe have been examined for nearly 30 years, regularly, in fixed geometry. Additionally, at the particular point of a calibration bench, determination of neutron fluence rate, ambient dose equivalent rate and its scattered component, total neutron and gamma dose rate, gamma to total dose ratio, radiation quality factor and total tissue kerma are occasionally determined. In this article, we would like to present recently achieved results and their comparison with the measurement data formerly presented. The growth of the neutron emission of 239PuBe source, as well as discrepancies between the decay curves of relevant isotopes and the emission rate of 252Cf and 241AmBe will be presented.

MEASUREMENTS OF THE IONISING RADIATION LEVEL AT A NUCLEAR MEDICINE FACILITY PERFORMING PET/CT EXAMINATIONS.

This paper presents the results of radiation level measurements at workplaces in a nuclear medicine facility performing PET/CT examinations. This study meticulously determines the staff radiation exposure in a PET/CT facility by tracking the path of patient movement. The measurements of the instantaneous radiation exposure were performed using an electronic radiometer with a proportional counter that was equipped with the option of recording the results on line. The measurements allowed for visualisation of the staff's instantaneous exposure caused by a patient walking through the department after the administration of 18F-FDG. An estimation of low doses associated with each working step and the exposure during a routine day in the department was possible. The measurements were completed by determining the average radiation level using highly sensitive thermoluminescent detectors.

Study on the influence of the B4C layer thickness on the neutron flux and energy distribution shape in multi-electrode ionisation chamber.

A model of a multi-electrode ionisation chamber, with polypropylene electrodes coated with a thin layer of B4C was created within Monte Carlo N-Particle Transport Code (MCNPX) and Fluktuierende Kaskade (FLUKA) codes. The influence of the layer thickness on neutron absorption in B4C and on the neutron spectra in the consecutive intra-electrode gas volumes has been studied using the MCNPX and FLUKA codes. The results will be used for designing the new type of the ionisation chamber.

Characterisation of radiation field for irradiation of biological samples at nuclear reactor-comparison of twin detector and recombination methods.

Central Laboratory for Radiological Protection is involved in achieving scientific project on biological dosimetry. The project includes irradiation of blood samples in radiation fields of nuclear reactor. A simple facility for irradiation of biological samples has been prepared at horizontal channel of the nuclear reactor MARIA in NCBJ in Poland. The radiation field, composed mainly of gamma radiation and thermal neutrons, has been characterised in terms of tissue kerma using twin-detector technique and recombination chambers.

Low-level gamma and neutron monitoring based on use of proportional counter filled with 3He in polythene moderator: study of the responses to gamma and neutrons.

It has been shown that a proportional counter filled with (3)He placed centrally inside a polythene sphere opens a new possibility for measuring gamma photons and neutrons in the separate pulse-height windows. The responses to gamma and neutrons (in terms of ambient dose equivalent) of the detector assembly consisting of 203-mm polythene sphere with centrally positioned 40-mm diameter (3)He proportional counter have been studied. The response to secondary gammas from capture process in hydrogen has also been studied. The rather preliminary studies indicate that the proposed measuring system has very promising features as an ambient dose equivalent device for mixed gamma-neutron fields.

Investigations of recombination chambers for BNCT beam dosimetry.

A set of cylindrical recombination chambers, including a tissue-equivalent chamber and three graphite chambers filled with different gases-CO(2), N(2) and (10)BF(3), was designed for the dosimetry of therapeutic neutron radiation beams used for BNCT. The separation of the dose components is based on differences of the shape of the saturation curve depending on the LET spectrum of the investigated radiation. The measurements using all the chambers were performed in a reactor beam of NRI ReZ (Czech Republic) and in the reference radiation fields of a (252)Cf radiation source free in air or in filters.

Ionisation chamber containing boron as a neutron detector in medical accelerator fields.

A combination of the recombination principle of H(10) measurements with the use of the ionisation chambers containing boron has been presented, in order to increase the relative sensitivity of the chamber to neutrons by a factor close to the radiation quality factor of photoneutrons. Three types of the chambers were investigated. Two of them were filled with BF(3) and the third one contained electrodes covered with B(4)C. All the chambers were placed in paraffin moderators. The response of the chambers was investigated, depending on gas pressure and polarising voltage. The results showed that it was possible to obtain nearly the same response of the chamber to H(10) for photons and neutrons in a restricted energy range; however, further investigations are needed to make an optimum design.

Measurements of the neutron dose near a 15 Mv medical linear accelerator.

In this work, simplified recombination methods for routine estimation of dose equivalent in mixed (gamma and neutrons) radiation field outside the irradiation field of linear medical accelerators is considered. The author's earlier reported method of H(10) measurements, involving determination of the recombination index of radiation quality, Q(4) by tissue-equivalent recombination chamber was combined with the new method for determination of the photon to neutron dose ratio D(X)/D(n) from the ratio of ion collection efficiencies measured in the investigated radiation field and in two reference fields of gamma and neutron radiations. The method is suitable when the neutron contribution to the total absorbed dose, D(n)/D, is >3%.

Radionuclide neutron sources in calibration laboratory--neutron and gamma doses and their changes in time.

The calibration laboratory, having standard neutron fields of radionuclide sources, should perform regular measurements of fields' parameters in order to check their stability and to get knowledge of any changes. Usually, accompanying gamma radiation is not of serious concern, but some personal dosemeters, old neutron dose equivalent meters with scintillation detectors and the dose meters of mixed radiation require the determination of this component. In the Laboratory of Radiation Protection Measurements in the Institute of Atomic Energy, Poland, the fields of radionuclide neutron sources (252)Cf, (241)Am-Be and (239)Pu-Be were examined for nearly 20 y. A number of detectors and methods have been applied for the determination of neutron ambient dose equivalent rate and for the determination of neutron and gamma dose components. This paper presents the recent results of measurements of gamma and neutron dose and dose equivalent, compared with the results accumulated in nearly 20 y.

Recombination chambers filled with different gases--studies of possible application for BNCT beam dosimetry.

Recombination microdosimetric method (RMM), based on the phenomenon of initial recombination of ions is applied to determine the distribution of the absorbed dose versus linear energy transfer (LET). Usually, the recombination chambers used for RMM are filled with tissue-equivalent gas, but the response of the device can be adjusted to the actual needs by the use of different gases. Using a graphite chamber filled with nitrogen and 10BF3 it was shown that RMM can also be used with chambers containing these gases. This opens the possibility of designing a recombination chamber for the determination of the dose fractions due to gamma radiation, fast neutrons, neutron capture on nitrogen and high-LET particles from the (n,10B) reaction in simulated tissue with different contents of 10B. It was also necessary to improve the method for the determination of initial recombination at low polarising voltages, when volume-recombination and back-diffusion of ions are considerably high.