The use of passive environmental TLDs in the operation of the Spanish early warning network 'REVIRA'

As required by different international agreements, the regulatory body in Spain (Consejo de Seguridad Nuclear) implemented in 1992 a national automatic network (REVIRA) that continuously monitors radiation levels in order to give early warning of incidents having potential transboundary implications. The detector for environmental gamma-radiation dose rate is an active instrument based on a Geiger-Müller counter. However, the use of passive environmental dosemeters provides an additional low-cost dose estimate with an independent centralised calibration and even better basic features than active instruments. Since 1999, all 25 REVIRA stations have been monitored with passive TL environmental dosemeters based on LiF:Mg,Cu,P and operated according to the procedures established at Ciemat. This paper presents the obtained results and the further analysis considering differences in aspects such as photon energy response, inherent background or response to cosmic rays. The benefits of the use of passive environmental dosemeters in early warning networks are discussed.     

Das Geiger-Müller-Zählrohr

This article is a contribution to the history of the Geiger-Müller counter, based on the replication method. Along with a historical analysis of the Geiger-Müller counter I present and discuss my experience with replicas of early counters from 1928. The explicit and tacit knowledge that was gained during my construction und the experimentation process allows a different view of the development process in 1928. Against this background, I discuss Walter Müller’s role in the development and the significance of experimental work in Geiger’s laboratory by illustrating the status of data gained from indirect electrical measurements in the field of early radioactivity research. All things considered it was not the instrument itself, but Geiger’s influence and status that affected its establishment in the scientific community in an authoritative way.     

Application of semiconductors for dosimetry of fast-neutron therapy beam

Two types of ion implanted miniature p-i-n diodes were tested in a d(48.5) + Be fast-neutron beam produced in the Detroit superconducting cyclotron. The increase in forward voltage drop caused by neutron-induced damage was correlated with neutron dose measured in a water phantom. The neutron and gamma dose components were predetermined using twin detector (Tissue-equivalent ion chamber paired with miniature Geiger-Müller counter) method. The increase in the voltage drop for 1 mA injection current was monitored together with the cyclotron beam target current, thus the differential voltage drop could be defined precisely for given radiation dose. The average neutron sensitivities of tested diodes were 1.284 +/- 0.014 and 0.528 +/- 0.058 mV per cGy. The miniature detectors can be utilised in characterisation of small radiation fields and in the regions of high dose gradient as well as for in vivo dosimetry of the patients undergoing fast-neutron therapy.     

Measurements of photon spectra in mixed fields at a nuclear installation

This paper describes the measurements of photon spectra in mixed neutron/photon radiation fields at a few locations in a nuclear reactor. The measurements were performed inside the containment of reactor 4 at the Swedish reactor site Ringhals, with a Ge-detector (4%). The measurements were carried out as a part of a EURADOS project in co-operation with the Swedish authorities and the reactor operating company. The measurements showed that a large fraction of the photons are high-energy photons (up to 7.6 MeV). This implies that GM-based photon detectors will overread in these fields since this type of detector generally overestimates the ambient dose equivalent in 6–7 MeV photon fields. The measurements also indicated that the photon field was almost isotropic, which in turn implies that the effective dose as well as the personal dose equivalent will be lower than the ambient dose equivalent.     

A wide-range direction neutron spectrometer

A new device is presented which has been developed for measuring the energy and direction of distribution of neutron fluence in fields of broad energy spectra (thermal to 100 MeV) and with a high background of photon, electron and muon radiation. The device was tested in reference fields with different energy and direction distributions of neutron fluence. The direction-integrated fluence spectra agree fairly well with reference spectra. In all cases, the ambient and personal dose equivalent values calculated from measured direction-differential spectra are within 35% of the reference values. Independent measurements of the directional dose equivalent were performed with a directional dose equivalent monitor based on superheated drop detectors.     

Calibration and testing of a TLD dosemeter for area monitoring

The response of a TLD-600/TLD-700 area dosemeter has been characterized in neutron fields around the 590 MeV cyclotron ring at the Paul Scherrer Institute (PSI). The dosemeter is based on a cylindrical paraffin moderator with three of each type of TLD chip at the centre, and is intended to use for area monitoring around accelerator facilities. The dosemeter is calibrated in terms of ambient dose equivalent using a non-moderated 252Cf neutron source. The ambient dose equivalent response has been tested in five locations where the neutron fields and dose rates have been well characterized by Bonner sphere spectrometer and active neutron monitor measurements. The different spectrum shapes and dose rates in the five locations permit the comparison of the behavior of the active and passive dosemeters in these neutron fields.     

A chamber for determining the conventionally true value of Hp(10) and H*(10) needed by calibration laboratories

A secondary standard chamber for photon radiation developed for measuring directly the conventionally true value of the personal dose equivalent, Hp(10), in a slab phantom is now commercially available. In addition, this chamber can be used for determining the true value of the ambient dose equivalent, H*(10), in monodirectional radiation fields; for example, photon fields generated by X ray facilities. Once the chamber has been calibrated at the facility of the calibration laboratory, the true value of Hp(10) or H*(10) can be measured at other facilities without applying any conversion coefficients. For low energy photon fields the conversion coefficients are strongly dependent on the spectral distribution. For nominally the same radiation quality small spectral differences, caused, for example, by use of different X ray facilities, may lead to differences between the spectrum-averaged conversion coefficients from Ka to Hp(10) and H*(10), respectively, of up to several tens per cent. For this reason, tabulated conversion coefficients for low energy radiation fields cannot be used for calibration purposes, if the standard uncertainty is to be 2-5%. Direct measurement by the secondary standard chamber overcomes this problem.     

Monte Carlo simulations of the photon calibration fields at the underground laboratory of PTB

A unique photon calibration facility operated by Physikalisch-Technische Bundesanstalt (PTB) provides photon fields with area dose rates in the order of the natural environmental radiation and even below. This facility is located in an underground laboratory in the Asse salt mine at a depth of 490 m below ground, where the ambient dose equivalent rate is only 2 nSv h(-1). Radioactive sources of the nuclides (241)Am, (57)Co, (137)Cs, (60)Co and (226)Ra are used to generate photon fields with different characteristics. In the past, the basic properties of the photon field, especially the area dose rate at the reference point and the mean energy of the photon spectra, were calculated by using analytic methods. However, information about scattered photons is only accessible through an investigation of spectra by performing Monte Carlos simulations. Therefore, the photon spectra at the reference point of the calibration facility were calculated using the Monte Carlo transport code MCNP. The results obtained by using this method are of relevance for the traceability of the reference dose rate values to PTB's primary standards, as well as for the determination of the mean photon energy of the spectra. The latter was calculated with respect to the different quantities 'photon fluence', 'air kerma' and 'ambient dose equivalent'. The origin of the scattered component in the photon spectrum is investigated in detail by studying the photon field produced by the quasi-monoenergetic gamma emitter (137)Cs (E(γ) = 662 keV) under various geometrical conditions. Implications of the Monte Carlo simulations on the traceability of the dose rate reference values as well as on the assessment of uncertainties will be described.     

The use of the Monte Carlo simulation technique for the design of an H*(10) dosemeter based on TLD-100

The thermoluminescence (TL) detector material LiF:Mg,Ti (TLD-100) and appropriate filter materials were combined in order to design a passive dosemeter measuring the operational quantity ambient dose equivalent, H*(10), for monitoring low-dose external photon radiation fields. Using the Monte Carlo simulation technique, optimisations of energy dependent conversion coefficients from air kerma free-in-air compared to ICRU and ISO proposed values. h*K(10), were performed by varying dosemeter detector positioning. geometrical arrangements, and filter materials. Deviations smaller than 5% compared to h*K(10) between 30 keV and 2.5 MeV of primary photon energies were achieved by a dosemeter design consisting of a 15 microm Sn metal layer and a 5 mm PMMA layer surrounding the LiF detector. Subsequently performed free-air verification experiments carried out in well defined standard photon radiation fields showed an obviously TL-specific effect. An underestimation up to -15% of the modelled data at low photon energies was observed.     

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.     

The influence of the type of filling gas on the response of ionisation chambers to a mixed high-energy radiation field

Radiation protection dosimetry in radiation fields behind the shielding of high-energy accelerators such as CERN is a challenging task and the quantitative understanding of the detector response used for dosimetry is essential. Measurements with ionisation chambers are a standard method to determine absorbed dose (in the detector material). For applications in mixed radiation fields, ionisation chambers are often also calibrated in terms of ambient dose equivalent at conventional reference radiation fields. The response of a given ionisation chamber to the various particle types of a complex high-energy radiation field in terms of ambient dose equivalent depends of course on the materials used for the construction and the chamber gas used. This paper will present results of computational studies simulating the exposure of high-pressure ionisation chambers filled with different types of gases to the radiation field at CERN's CERN-EU high-energy reference field facility. At this facility complex high-energy radiation fields, similar to those produced by cosmic rays at flight altitudes, are produced. The particle fluence and spectra calculated with FLUKA Monte Carlo simulations have been benchmarked in several measurements. The results can be used to optimise the response of ionisation chambers for the measurement of ambient dose equivalent in high-energy mixed radiation fields.     

A mobile TEPC-based system to measure the contributions to H*(10) at flight altitudes

A very promising method to measure the ambient dose equivalent H*(10) at flight altitudes is to use Tissue Equivalent Proportional Counters (TEPC). The measured quantity is the lineal energy, y, which can be converted into equivalent dose as a good estimate of H*(10). According to the lineal energy transfer (LET) spectra one may even extract information about the composition of the radiation field. A new system was developed by adding a surrounding coincidence detector (CACS: Coincidence/Anti-Coincidence Shield) that allows one to identify the primary particle, which deposits energy in the TEPC, as neutral or charged. The entire system was calibrated in the neutron and high-energy photon reference fields at PTB. One of the results of these measurements is, the use of low- and high-LET calibration factors when performing measurements in mixed radiation fields. The TEPC/CACS system is now operated on-board aircrafts as a fixed or mobile dosimetry system.     

Individual monitoring in mixed neutron/photon fields using a single silicon detector

A device based on a single silicon detector of special converter/detector design optimised for the determination of the neutron dose equivalent is also used for the determination of the photon dose equivalent. While the neutron dose is determined on the basis of signals corresponding to energy depositions above 1.5 MeV, depositions between 80 keV and 150 keV are used for the photon dose equivalent. In this way, a photon response is achieved which varies by less than 30% in the energy region from 80 keV to 7 MeV for irradiation at normal incidence and at 60 degrees to the normal. The lower limit of detection is of the order of 1 microSv. Neutrons contribute to the photon reading by less than 2% in mixed fields with a comparable dose equivalent from neutrons and photons.     

Evaluation of dose equivalent by the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities

This paper presents the results of measurements made with the electronic personal neutron Saphydose-N during the four campaigns of the European contract EVIDOS (EValuation of Individual DOSimetry in mixed neutron and photon radiation fields). These measurements were performed at Institute for Radiological Protection and Nuclear Safety (IRSN) in France (C0), at the Krümmel Nuclear Power Plant in Germany (C1), at the VENUS Research Reactor and the Belgonucléaire fuel processing plant in Belgium (C2) and at the Ringhals Nuclear Power Plant in Sweden (C3). The results for Saphydose-N are compared with reference values for dose equivalent.     

Long-term measurements of H*(10) at aviation altitudes in the northern hemisphere

Monitoring the radiation field at aviation altitudes is achieved by the dosemeter system piDOS installed in a passenger aircraft. The basic detector is a 2 in. tissue-equivalent proportional counter (TEPC) mounted in a cabin-baggage-sized aluminium suitcase. The entire system was characterised in neutron and photon reference fields from which two calibration factors were determined according to the splitting of the measured dose deposition spectrum yd(y) into low-LET and high-LET regions. A total of 255 flights in the northern hemisphere was analysed. The dependencies of the ambient dose equivalent rates on altitude, latitude and solar activity have been determined. These new data extend the data base used in Germany for the validation of program codes to calculate the radiation exposure of air crew members.     

Shielding evaluation of a medical linear accelerator vault in preparation for installing a high-dose rate 252Cf remote afterloader

In support of the effort to begin high-dose rate 252Cf brachytherapy treatments at Tufts-New England Medical Center, the shielding capabilities of a clinical accelerator vault against the neutron and photon emissions from a 1.124 mg 252Cf source were examined. Outside the clinical accelerator vault, the fast neutron dose equivalent rate was below the lower limit of detection of a CR-39 etched track detector and below 0.14 +/- 0.02 muSv h(-1) with a proportional counter, which is consistent, within the uncertainties, with natural background. The photon dose equivalent rate was also measured to be below background levels (0.1 muSv h(-1)) using an ionisation chamber and an optically stimulated luminescence dosemeter. A Monte Carlo simulation of neutron transport through the accelerator vault was performed to validate measured values and determine the thermal-energy to low-energy neutron component. Monte Carlo results showed that the dose equivalent rate from fast neutrons was reduced by a factor of 100,000 after attenuation through the vault wall, and the thermal-energy neutron dose equivalent rate would be an additional factor of 1000 below that of the fast neutrons. Based on these findings, the shielding installed in this facility is sufficient for the use of at least 5.0 mg of 252Cf.     

Development of a light-weight portable neutron survey meter

A light-weight portable neutron survey meter was developed using a mixed organic gas counter for dose management at nuclear power plants and accelerator facilities. This survey meter, NSN31041, is ~2 kg in weight and W160×H250×L300 mm(3) in size, which is capable of measuring neutron ambient dose equivalent rate from thermal to 15 MeV neutrons. The neutron energy response of the survey meter is evaluated using continuous energy neutron sources of (252)Cf, (241)Am-Be, thermal neutrons generated from a graphite pile loading a (252)Cf source, concrete-moderated neutrons of (241)Am-Be source and D(2)O-moderated neutrons of (252)Cf source. The measured response data show very good agreement with neutron ambient dose equivalent within a 50 % deviation.     

Dose build up correction for radiation monitors in high-energy bremsstrahlung photon radiation fields

Conventional radiation monitors have been found to underestimate the personal dose equivalent in the high-energy bremsstrahlung photon radiation fields encountered near electron storage rings. Depth-dose measurements in a water phantom were carried out with a radiation survey meter in the bremsstrahlung photon radiation fields from a 450 MeV electron storage ring to find out the magnitude of the underestimation. Dose equivalent indicated by the survey meter was found to build up with increase in thickness of water placed in front of the meter up to certain depth and then reduce with further increase in thickness. A dose equivalent build up factor was estimated from the measurements. An absorbed dose build up factor in a water phantom was also estimated from calculations performed using the Monte Carlo codes, EGS-4 and EGSnrc. The calculations are found to be in very good agreement with the measurements. The studies indicate inadequacy of commercially available radiation monitors for radiation monitoring within shielded enclosures and in streaming high-energy photon radiation fields from electron storage rings, and the need for proper correction for use in such radiation fields.     

Intercomparison on measurements of the quantity personal dose equivalent Hp(10) in photon fields in the African region

An intercomparison was arranged by the International Atomic Energy Agency for the African region for personal dose equivalent. This was arranged with a view to helping member states to comply with dose limitation requirements and to harmonise the use of internationally agreed quantities and recommended assessment methods. The scope of the intercomparison was aimed at passive dosemeters, which determine the personal dose equivalent in photon radiation fields. The comparison protocol is presented together with the results.     

Monte Carlo simulation of the IRSN CANEL/T400 realistic mixed neutron-photon radiation field

The calibration of dosemeters and spectrometers in realistic neutron fields simulating those encountered at workplaces is of high necessity to provide true and reliable dosimetric information to the exposed nuclear workers. The CANEL assembly was set-up at IRSN to produce such neutron fields. It comprises a depleted uranium shell, to produce fission neutrons, then iron and water to moderate them and a polyethylene duct. The new presented CANEL facility is used with 3.3 MeV neutrons. Calculations were performed with the MCNP4C code to characterise this mixed neutron-photon expanded radiation field at the position where calibrations are usually performed. The neutron fluence energy and the direction distributions were calculated and the operational quantities were derived from these distributions. The photon fluence and corresponding ambient dose equivalent were also estimated. Comparison with experimental results showed an overall good agreement.