Response of Harshaw neutron thermoluminescence dosemeters in terms of the revised ICRP/ICRU recommendations

To monitor workers for external neutron radiation dose, the Y-12 National Security Complex utilises the thermoluminescence dosemeters (TLDs) manufactured by Harshaw. At Y-12, the majority of external dose to workers is due to low-energy photon and/or beta particles emitted from uranium and its progeny. However, some neutron dose is expected since neutrons are produced from (alpha,n) reactions in various compounds found at the plant, including UF4 and UF6. Neutron sources, such as 252Cf, are also used throughout the complex. The Harshaw neutron dosemeter consists of two gamma-sensitive elements (7Li) and two neutron-sensitive elements enriched in 6Li with various shielding/filter materials placed around each of them. In this work, the energy response of the dosemeter to neutrons has been calculated using the Monte Carlo transport code MCNP Version 4-C and, these results are compared with the measured response of the dosemeter to unmoderated and D2O-moderated 252Cf neutrons. The response of the dosemeter has also been determined in terms of the personal absorbed dose and personal dose equivalent as a function of neutron energy based on the recommendations of the ICRP Publication 60 and ICRU Report 49. The energy response of the dosemeter characteristics can be used to generate spectral conversion coefficients for routine neutron absorbed dose and dose equivalent calculations.     

IEC standards for individual monitoring of ionising radiation

This paper presents IEC/SC 45B 'Radiation protection instrumentation' and its standards for individual monitoring of ionising radiation: IEC 61526 Ed. 3 for active personal dosemeters and IEC 62387-1 for passive integrating dosimetry systems. The transposition of these standards as CENELEC (European) standards is also discussed together with the collaboration between IEC/SC 45B and ISO/TC 85/SC 2.     

The outlook for the application of electronic dosemeters as legal dosimetry

This study analyses new trends of a set of 12 electronic personal dosemeters in order to gain an overview of their main advantages and limitations. Physical characteristics and radiological, mechanical and environmental performance were tested according to IEC-61526 Standard requirements. The study highlights the different behaviour of the selected dosemeters. In particular, it is demonstrated that three of the tested devices fulfilled most of the established requirements, whereas another three of them presented important faults. The parameters that need more development are, in general, the response at low energy photon and beta radiation, and the dose rate alarm features. In some cases, mechanical problems as well as interference in the response due to external electromagnetic fields were also found. However, the results of the study foresee a promising future for the application of the newest personal electronic dosemeters as legal personal dosemeters and show the need for internationally agreed technical requirements within the European countries.     

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.     

Measured and calculated angular responses of panasonic UD-809 thermoluminescence dosemeters to neutrons

Multi-element thermoluminescence dosemeters (TLD), such as the Panasonic UD-809, are used in personal dosimetry. The Panasonic UD-809 dosemeter consists of one gamma sensitive and three neutron sensitive TLD elements with different filter materials. In this work, the neutron energy responses (the number of (n,alpha) reactions per neutron) of the neutron-sensitive TLD elements of the Panasonic UD-809 dosemeter were calculated using the MCNP Monte Carlo transport code. Experiments were performed in a calibration geometry with an unmoderated 252Cf neutron source. These measurements were made with the dosemeter placed on the centre front face of a polymethylmethacrylate (PMMA) slab phantom. The phantom was rotated in the horizontal plane from -90 to +90 degrees, in 15 degree increments. Good agreement between calculated and measured element responses was observed. The angular dependency of personal dose equivalent was also calculated for parallel beams of 252Cf neutrons and compared to the TLD element angular responses.     

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.     

Electronic personal dosemeters: the solution to problems of individual monitoring in mixed neutron/photon fields?

An overview is given showing the main principles of the present-day electronic neutron dosemeters. The radiological performance of the devices is described in a comparative way. This includes chiefly the personal dose equivalent Hp(10) response for monoenergetic neutrons and in practical fields with broad energy distributions and estimations of the low dose limit for neutrons.     

Comparison test of electronic dosemeters

To assist with a planned purchase of electronic dosemeters by the Swiss Federal Office for Civil Protection, the calibration laboratory of the Paul Scherrer Institute performed tests on 11 types of electronic dosemeters manufactured by 10 European and American companies. The technical specifications for the World Trade Organisation (WTO) tendering procedure were largely in accord with the specifications of the international standard IEC 61526. First tests were performed with samples from each type of dosemeter. The reproducibility of a dose of 0.1 mSv generated with 137Cs radiation at a dose rate of 2.1 mSv.h-1 was found adequate for all tested dosemeter types. The response for environmental levels of radiation showed a large variation, indicating insufficient background correction of some dosmeters. A very high dose rate of 10 Sv.h-1 provoked faulty dose readings for more than half of the tested dosemeters. Dosemeter response for low-energy photon radiation was satisfactory for two of the tested dosemeter types. Four dosemeter types were selected for extended technical tests. Three samples of each of these dosemeter types were purchased. For drop and temperature tests the specifications of the WTO tendering procedure outranged the specifications of the IEC standard. Whereas even at a temperature of -25 degrees C the tested dosemeters functioned normally, drops from a height of 2 m onto a wooden surface rendered the samples of two dosemeter types inoperative.     

Determination of personal dose equivalents in accelerator radiation fields

Values for the dose equivalent are required for radiation protection purposes, but determination of such values can be quite difficult for high energy radiations. The accurate determination of personal dose equivalents in accelerator radiation fields requires the propel use of appropriate radiological quantities and units, knowledge of the dose equivalent response of the personal dosemeters used, measurement or calculation of the fluence spectrum in the workplace and the fluence spectrum of the reference radiation used to calibrate the dosemeters, in addition to knowledge of the appropriate fluence-to-dose equivalent conversion coefficients. This information can then be used to select the appropriate dosemeters, set up the optimum calibration conditions, or to establish correction factors that account for differences in the calibration and workplace fluence spectra. High energy neutrons account for a significant fraction of the dose equivalent received by workers at accelerator facilities, and this work discusses the procedures and methods needed to determine dose equivalent produced by neutrons in the vicinity of high energy particle accclerators.     

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.     

Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces.     

Calibration of personal dosemeters in mixed neutron-photon fields: some problems and their solution

In neutron reference radiation fields, the conventional true value of the personal dose equivalent, H(p)(10), is derived from the spectral neutron fluence and recommended conversion coefficients. This procedure requires the phantom on which the personal dosemeter is mounted to be irradiated with a broad and parallel beam. In many practical situations, the change of the neutron fluence and/or the energy distribution over the surface of the phantom may not be neglected. For a selection of typical irradiation conditions in neutron reference radiation fields, the influence of this effect has been analysed using numerical methods. A further problem, which is of relevance for the calibration of dosemeters measuring both the neutron and the photon component of mixed fields, is the 'double counting' of the dose equivalent due to neutron-induced photons. The relevance of this conceptual problem for calibrations in mixed-field dosimetry was analysed.     

Assessment of neutron dosemeters around standard sources and nuclear fissile objects

In order to evaluate the neutron doses around nuclear fissile objects, a comparative study has been made on several neutron dosemeters: bubble dosemeters, etched-track detectors (CR-39) and 3He-filled proportional counters used as dose-rate meters. The measurements were made on the ambient and the personal dose equivalents H*(10) and Hp(10). Results showed that several bubble dosemeters should have been used due to a low reproducibility in the measurements. A strong correlation with the neutron energy was also found, with about a 30% underestimation of Hp(10) for neutrons from the PuBe source, and about a 9% overestimation for neutrons from the 252Cf source. Measurements of the nuclear fissile objects were made using the CR-39 and the dose-rate meters. The CR-39 led to an underestimation of 30% with respect to the neutron dose-rate meter measurements. In addition, the MCNP calculation code was used in the different configurations.     

Electronic personal neutron dosemeters for high energies: measurements, new developments and further needs

The results of measurements with neutron energies up to 60 MeV are shown for the personal neutron dosemeters Thermo Electron EPD-N2, ALOKA PDM-313 and the PTB prototype dosemeter DOS-2002. All dosemeters show dose equivalent responses that are about a factor of 10, too high at 60 MeV. A new prototype dosemeter-called DOS-2005-consisting of a detector with a thin effective layer of 6 microm has been set up at PTB. The dose equivalent response of this dosemeter and that of the newly developed dosemeter SAPHYDOSE-N was measured up to 19 MeV. Both dosemeters indicate a more flat response at high neutron energies. Further needs-optimisations, measurements and calculations-for use at high-energy accelerators and in space are discussed.     

The right choice: extremity dosemeter for different radiation fields

Measurements of weakly penetrating radiation in personal dosimetry present problems in the design of suitable detectors and in the interpretation of their readings. For the measurement of the individual beta radiation dose, personal dosemeters for the fingers/tips are required. LiF:Mg,Cu,P is a promising thermoluminescent (TL) material which allows the production of thin detectors with sufficient sensitivity. Dosimetric properties of two different types of extremity dosemeters, designed to measure the personal dose equivalent Hp(0.07), have been compared: LiF:Mg,Ti (TLD100) and LiF:Mg,Cu,P (TLD700H). A type test for energy response for photon and beta radiation according to ISO 4037-3 and ISO-6980 was carried out and the results for both dosemeters were compared. Simultaneous measurements with both types of dosemeters were performed at workplaces, where radiopharmaceuticals containing different radioisotopes are prepared and applied. Practices in these fields are characterized by handling of high activities at very small distances between source and skin. The results from the comparison of the two-dosemeter types are presented and analysed with respect to different radiation fields. Experiments showed a satisfactory sensitivity for the thinner dosemeter (TLD 700H) for detecting beta radiation at protection levels and a good energy response.     

Evaluation of two personal dosemeters in polyenergetic mono- and multi-directional neutron fields

The neutron dose-equivalent response of two commercially available electronic personal neutron dosemeters was studied in several laboratory-produced broad-spectrum neutron fields. Fluence-weighted mean energies ranged from 200 keV to 4 MeV; personal dose-equivalent rates ranged from 75 to 10 mSv h(-1); and angles of incidence were multidirectional, 0 degrees, 30 degrees and 60 degrees. Three of these fields have been shown previously to resemble ones found in CANDU (Canadian Deuterium Uranium is a registered trademark of the Atomic Energy of Canada Limited) power plant workplaces. Both dosemeters were found to perform reasonably well across the range of energy spectra and angles of incidence. One type of dosemeter displayed values of the personal dose equivalent that were, at worst, within a factor of approximately 2 of the reference values and, at best, within a few per cent of the reference values. The other type displayed values of the personal dose equivalent that were consistently within unity and 20% of the reference values. Although the radiological performance of one was found to be more accurate, this device was also found to be the less rugged of the two. Some of the data acquired in this work were compared with results previously published by others. There was consistency between these sets of data.     

Establishment of a simple neutron calibration field using a moderated 252Cf source: Part II. Experimental characterization of simple neutron calibration field

This paper describes the development and experimental standardization of neutron fields simply arranged for detector calibrations used for radiation control and environmental measurement. These fields are the following: (1) bare ²⁵²Cf fission field, (2) iron-moderated ²⁵²Cf field, (3) carbone-moderated Cf field, and (4) polyethylene-moderated ²⁵²Cf field. These fields are most suitable for calibrating the detectors used in and around nuclear and radiation facilities, since the fields are designed to simulate the typical neutron fields in and around the facilities.The direct neutron components of these fields have been standardized by the following two methods: (1) calculation by the ANISN code, and (2) measurements with and without a shadow shield by detectors standardized in the national standard field at the Electrotechnical Laboratory (ETL). The neutron emission rates of the ²⁵²Cf source have been calibrated also at ETL. We have standardized only direct components because of their independence of room size and peripheral structures. The standardized values are energy spectra and dose equivalent rates of the direct neutron components; the accuracies have also been evaluated to be 20% below 100 keV, 15% at 1 MeV, and 50% above 5 MeV. These fields including room scattered components have also been characterized especially to calibrate neutron detectors having sensitivity to low energy room scattered neutrons, because of large errors caused by shadow shield subtraction.     

Evaluation of spectrum measurement devices for operational use

Several neutron spectrometers manufactured by Bubble Technology Industries (BTI) were tested and evaluated in a variety of neutron fields. Findings and conclusions are presented for the following BTI instruments: a modification of the Rotational Spectrometer (ROSPEC) that includes a thermal and epithermal capability, the Simple Scintillation Spectrometer that is used in conjunction with the ROSPEC to extend its high-energy range, and the MICROSPEC N-Probe which is capable of providing a crude spectrum over the energy range from thermal to 18 MeV. The main objective of these measurements was to determine the accuracy of both the energy spectrum and dose equivalent information generated by these devices. In addition, the dose response of the Wide-Energy Neutron Detection Instrument (WENDI-II) was measured in all neutron fields relative to a bare ²⁵²Cf calibration. The performance of the WENDI-II rem meter was compared to the dose information generated by the neutron spectrometers. The instruments were irradiated to bare ²⁵²Cf and ²⁴¹AmBe sources, and in a series of moderated ²⁵²Cf fields using a standard D₂O sphere and a set of polyethylene spheres. The measured spectra were benchmarked with a set of detailed Monte Carlo calculations with the same energy bin structure as that of the instruments under test. These calculations allowed an absolute comparison to be made with the measurements on a bin by bin basis. The simulations included the effects of room return and source anisotropy.     

Current problems and expected improvements in personal neutron dosimetry

Many technological activities involve the potential for worker exposure to neutrons. The determination of neutron personal dose equivalent is difficult due to a number of factors including the materials and methods used to evaluate the response of personal dosemeters and the quantities for expressing dose equivalent. Nevertheless, recent progress has been made in the development of devices and techniques for the measurement and calibration of neutron personal dosemeters. The quantities and units used to express neutron dose equivalent are being improved and clarified. Therefore, it is expected that a number of remaining difficulties with neutron dosimetry will be mitigated.     

Current challenges in personal dosimetry at the US DOE Hanford site

An overview is presented of the dosimetry system, dose equivalent calculation methodology, and QA/QC practices used at the US Department of Energy Hanford site. It describes some of the problems encountered in accurately measuring dose equivalent quantities under a broad range of field conditions that do not necessarily correlate with laboratory calibration conditions and the approach taken to solve these problems. Personnel at Hanford are monitored with a combination of Harshaw model 8825 and 8816 thermoluminescence dosemeters and CR-39 etched track dosemeters. Extremities are monitored using the ICN MeasuRing loaded with a Harshaw XD740 chipstrate TLD. All dosemeters employ LiF:Mg,Ti elements that are read on-site with Harshaw model 8800 and 6600 TLD readers. CR-39 dosemeters are electrochemically etched in non-commercial etch chambers and counted with an automated track counting system developed by Pacific Northwest National Laboratory. Problems with over response of the 8825 with respect to Hp(0.07), under-response of the 8825 with respect to Hp(3), and over response of the 8825 with respect to Hp(10) in Hanford's 90Sr/90Y beta radiation fields are discussed. Approaches to measurement of the operational quantities for field conditions and algorithm solutions to the above problems are described. Methods used to calibrate the ring dosemeter for Hanford field conditions together with limitations of the ring dosemeter in measuring Hp(0.07) for extremities, particularly when covered with protective clothing, are also discussed.