Energy and fluence calibration of the neutron spectrometer ROSPEC at the IRSN AMANDE facility between 70 keV and 4.5 MeV.

The ROSPEC device is a multi-detector system, which has been designed by Bubble Technologies Industries (BTI at Chalk River, ON, Canada) to assess neutron spectra, and hence neutron dose quantities, at workplace fields. It is made up of six gaseous proportional counters that detect neutrons via the elastic (n,p) scattering (four hydrogenous counters) and with the (3)He(n,p)T reaction (two (3)He-filled counters). Results of the calibration of a similar rotating spectrometer (ROSPEC) have been described by Rosenstock et al.((1)). For energy and fluence calibration purposes, measurements were performed with the accelerator for metrology and neutron applications in external dosimetry (AMANDE) facility at the Laboratory of Neutron Metrology and Dosimetry (Institute of Radiation Protection and Nuclear Safety, IRSN, France). This facility provides monoenergetic neutron radiation fields from 2 keV to 20 MeV. Two kinds of experiments were carried out. First, the ROSPEC was used in its rotational mode for the ISO energies. Then, each detector was irradiated with all the available neutron energies, in a well defined position with the rotation of the device stopped. The energy values of the neutron beam were calculated using the TARGET code. A BC501-A liquid scintillation spectrometer provided the fluence values for energies beyond 1.2 MeV, a methane-filled SP2 counter from 800 keV to 1.4 MeV and an H(2)-filled SP2 counter from 144 to 800 keV. Reference data for 70 keV monoenergetic neutrons were obtained using the IRSN Long Counter. Results showed that the ROSPEC device was in agreement with the absolute neutron fluences within 10%. Moreover, the new energy calibration factors are in good agreement with those derived by BTI.

Photon contribution to ambient dose equivalent H(10) in the wide-spectrum neutron reference fields of the IRSN.

The photon contribution to ambient dose equivalent in several wide-spectrum reference neutrons fields of the Institute for Radiological Protection and Nuclear Safety were measured using a Geiger-Müller counter. For the investigated fields, the ratio of photon to neutron ambient dose equivalent ranged between 0.03 and 0.20. The results show that the Geiger-Müller tube is a versatile instrument for dosimetry in mixed photon-neutron fields if sufficient information for the calculation of corrections is available.

Characterization of neutron field in a NPP workplace.

At the Krsko Nuclear Power Plant (NPP), albedo dosimeters are used for personal neutron dosimetry. Spectrometric measurements allow determination of reference dosimetric values of realistic neutron fields to be used for calibration of albedo dosimeters. The Laboratory for Neutron Metrology and Dosimetry from the Institute for Radiological Protection and Nuclear Safety (IRSN) was in charge of characterising neutron fields in the plant at two representative points with high neutron and gamma dose rate. Calibration of the dosimeters in the workplace used to be performed only by a spherical survey meter. Based on the reference dosimetric values, the Plant Dosimetry Laboratory has verified the response of albedo dosimeters.

Bonner sphere neutron spectrometry at nuclear workplaces in the framework of the EVIDOS project.

The Institute for Radiological protection and Nuclear Safety was engaged in the EC funded EVIDOS project to provide reference spectrometry data using its Bonner sphere system. The data were processed by means of two unfolding codes, NUBAY and GRAVEL, both provided by the Physikalisch-Technische Bundesanstalt. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterised spectrum and provides posterior probability distributions for the parameters. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. The BS measurements were used to establish the neutron fluence energy distributions and reference values for the neutron ambient dose equivalent. As this quantity depends strongly on the high energy neutrons, a sensitivity analysis was done by unfolding the BS data with GRAVEL using the NUBAY solution spectrum as default with various changes in the parameters of the high energy peak. This new method of analysing Bonner sphere data allowed the determination of reliable neutron spectra, as well as a very good estimate of the corresponding integral quantities with small associated uncertainties.

French comparison exercise with the rotating neutron spectrometer, 'ROSPEC'.

The French laboratories in charge of 'neutron' dosimetry using the spectrometer 'ROSPEC', formed a working group in 2001. The participants began to study the behaviour of the instrument with a comparison exercise in broad energy neutron fields recommended by the International Organisation for Standardisation (ISO) and available at the LMDN in Cadarache. The complete version of the ROSPEC is made up of six spherical proportional counters fixed to a rotating platform. These counters cover different energy ranges which overlap each other to provide a link between the detectors, within the energy range from thermal neutrons to 4.5 MeV. The irradiation configurations chosen were ISO standard sources (252Cf, (252Cf+D2O)(/Cd), 241Am-Be) and the SIGMA facility. The results show that the 'thermal and epithermal' neutron fluence was widely overestimated by the spectrometer in all configurations.

Criticality accident dosimetry systems: an international intercomparison at the SILENE reactor in 2002.

In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue.

Reference dosimetry measurements for the international intercomparison of criticality accident dosimetry SILENE 9-21 June 2002.

An international intercomparison of criticality accident dosimetry systems took place in the SILENE reactor, in June 2002. Participants from 60 laboratories irradiated their dosemeters (physical and biological) using two different configurations of the reactor. In preparation for this intercomparison, the leakage radiation fields were characterised by spectrometry and dosimetry measurements using the ROSPEC spectrometer associated with a NE-213 scintillator, ionisation chambers, GM counters, diodes and thermoluminescence dosemeters (TLDs). For this intercomparison, a large area was required to irradiate the dosemeters both in free air and on phantoms. Therefore, measurements of the uniformity of the field were performed with activation detectors and TLDs for neutron and gammas, respectively. This paper describes the procedures used and the results obtained.

Retrospective dosimetry after criticality accidents using low-frequency EPR: a study of whole human teeth irradiated in a mixed neutron and gamma-radiation field.

In the context of accidental or intentional radiation exposures (nuclear terrorism), it is essential to separate rapidly those individuals with substantial exposures from those with exposures that do not constitute an immediate threat to health. Low-frequency electron paramagnetic resonance (EPR) spectroscopy provides the potential advantage of making accurate and sensitive measurements of absorbed radiation dose in teeth without removing the teeth from the potential victims. Up to now, most studies focused on the dose-response curves obtained for gamma radiation. In radiation accidents, however, the contribution of neutrons to the total radiation dose should not be neglected. To determine how neutrons contribute to the apparent dose estimated by EPR dosimetry, extracted whole human teeth were irradiated at the SILENE reactor in a mixed neutron and gamma-radiation field simulating criticality accidents. The teeth were irradiated in free air as well as in a paraffin head phantom. Lead screens were also used to eliminate to a large extent the contribution of the gamma radiation to the dose received by the teeth. The EPR signals, obtained with a low-frequency (1.2 GHz) spectrometer, were compared to dosimetry measurements at the same location. The contribution of neutrons to the EPR dosimetric signal was negligible in the range of 0 to 10 Gy and was rather small (neutron/gamma-ray sensitivity in the range 0-0.2) at higher doses. This indicates that the method essentially provides information on the dose received from the gamma-ray component of the radiation.