Radiation Protection Quantities Research Articles

NORMAN-05 conversion coefficients for monoenergetic neutrons below 20 MeV

In recent years voxel models have been used to investigate the validity of radiation protection quantities evaluated, till now, employing stylized reference models. An ICRP task group (DOCAL) is developing a male and female reference voxel models to estimate a new set of radiation protection quantities. It should be taken into account that, due the intrinsic subject-dependent anatomical differences, the various developed voxel models can imply a certain variability in the corresponding sets of estimated conversion coefficients. Therefore a qualitative evaluation of the range of such discrepancies should be of interest in terms of the sensitivity of organ absorbed dose to the specific model used for its computation. The NORMAN-05 model, updated at ENEA-Radiation Protection Institute on the basis of the original NORMAN model, developed by P. Dimbylow at NRPB-UK (now HPA) in the 1990s, is a good representation of the standard Caucasian-type man and could be a valuable tool for comparison with other voxel models in order to state the inherent uncertainties associated to the estimated sets of conversion coefficients. The present paper summarizes the results obtained with NORMAN-05 and the Monte Carlo code MCNPX for neutron conversion coefficients calculation, in the energy domain from thermal to 20 MeV, in the kerma approximation that, for such energy range, as stated in the ICRU/ICRP recommendations, is a satisfactory approach.

Characterisation of mixed neutron photon workplace fields at nuclear facilities by spectrometry (energy and direction) within the EVIDOS project

Within the EC project EVIDOS, 17 different mixed neutron-photon workplace fields at nuclear facilities (boiling water reactor, pressurised water reactor, research reactor, fuel processing, storage of spent fuel) were characterised using conventional Bonner sphere spectrometry and newly developed direction spectrometers. The results of the analysis, using Bayesian parameter estimation methods and different unfolding codes, some of them especially adapted to simultaneously unfold energy and direction distributions of the neutron fluence, showed that neutron spectra differed strongly at the different places, both in energy and direction distribution. The implication of the results for the determination of reference values for radiation protection quantities (ambient dose equivalent, personal dose equivalent and effective dose) and the related uncertainties are discussed.

Preliminary studies on neutron conversion coefficients calculated with MCNPX in NORMAN voxel phantom

Effective dose is the main radiation protection quantity. Progresses in radiation studies brought ICRP to revise ICRP 60 recommendations. A new publication, already circulated in form of draft, is expected to change some aspects of effective dose evaluation method. The organ absorbed doses for neutrons at various energies and incidence angles, necessary to estimate the effective dose, have been published in ICRU 57 and ICRP 74 reports for ADAM and EVA analytical male and female phantoms and similar calculations were also performed, based on the MCNP code, for VIP-MAN voxel phantom. The NORMAN voxel phantom, developed on the basis of magnetic resonance data of an adult male at HPA (formerly NRPB), is an accurate model (with a voxel element of approximately 8 mm(3)), which well approximates the standard man and has been already employed for radiation protection studies with photons. In the present paper, a modified version, called NORMAN-05, including a new organ, the salivary glands (as suggested in the mentioned ICRP draft), and a more detailed skeletal description, especially devoted to red bone marrow dose evaluation, has been employed with the Monte Carlo code MCNPX to calculate neutron conversion coefficients from thermal energies to 20 MeV. Some preliminary results, for antero-posterior and postero-anterior irradiation conditions, are presented and compared with the available published data.

Measurements and simulations of the radiation exposure to aircraft crew workplaces due to cosmic radiation in the atmosphere

As required by the European Directive 96/29/Euratom, radiation exposure due to natural ionizing radiation has to be taken into account at workplaces if the effective dose could become more than 1 mSv per year. An example of workers concerned by this directive is aircraft crew due to cosmic radiation exposure in the atmosphere. Extensive measurement campaigns on board aircrafts have been carried out to assess ambient dose equivalent. A consortium of European dosimetry institutes within EURADOS WG5 summarized experimental data and results of calculations, together with detailed descriptions of the methods for measurements and calculations. The radiation protection quantity of interest is the effective dose, E (ISO). The comparison of results by measurements and calculations is done in terms of the operational quantity ambient dose equivalent, H(10). This paper gives an overview of the EURADOS Aircraft Crew In-Flight Database and it presents a new empirical model describing fitting functions for this data. Furthermore, it describes numerical simulations performed with the Monte Carlo code FLUKA-2005 using an updated version of the cosmic radiation primary spectra. The ratio between ambient dose equivalent and effective dose at commercial flight altitudes, calculated with FLUKA-2005, is discussed. Finally, it presents the aviation dosimetry model AVIDOS based on FLUKA-2005 simulations for routine dose assessment. The code has been developed by Austrian Research Centers (ARC) for the public usage (http://avidos.healthphysics.at).

EUROMET supplementary comparisons for photon and beta radiation--actual state and first results

In January 2004, two EUROMET supplementary comparisons started, coordinated by Physikalisch-Technische Bundesanstalt (PTB): one for the radiation protection quantity 'personal dose equivalent' Hp(10) using X-radiation, and one for the quantity 'absorbed dose in tissue' Dt(0.07) using beta reference fields. The focus of both comparisons lies on calibrations using radiation qualities at which the dissemination of the quantities considered here is difficult. Within the scope of the two comparisons, an ionisation chamber shall be calibrated. Together with a transfer chamber optimised to measure Hp(10) and Dt(0.07), respectively, for each comparison a complete electronic measuring system is circulated. These transfer devices, as well as, a specially developed test procedure, which has to be carried out by each participant to check the instruments, are described. For both comparisons, the results of those participants who had finished their measurements and delivered the measuring protocol until February 2005 were evaluated. As the comparisons are still running these first results are shown in an anonymous form. A first assessment of the data will be given.

Evaluation of individual dosimetry in mixed neutron and photon radiation fields (EVIDOS). Part II: conclusions and recommendations

The paper presents the main conclusions and recommendations derived from the EVIDOS project, which is supported by the European Commission within the 5th Framework Programme. EVIDOS aims at evaluating state of the art neutron dosimetry techniques in representative workplaces of the nuclear industry with complex mixed neutron-photon radiation fields. This analysis complements a series of individual papers which present detailed results and it summarises the main findings from a practical point of view. Conclusions and recommendations are given concerning characterisation of radiation fields, methods to derive radiation protection quantities and dosemeter results.

Monte Carlo evaluated parameters for internal dosimetry

The radiation doses received by individuals from radionuclides which enter the human body cannot be measured directly but must be inferred. In these calculations, several measurable quantities (such as the internal whole body burden or urine daily excretion) and quantities derived from models are employed. The Radiation Protection Quantities for internal dosimetry are, in principle, the same as for external dosimetry with the addition of quantities taking into account that the doses in the body are protracted. Other parameters are also necessary for the dose assessment, such as the SAFs (Specific Absorbed Fractions). All these quantities are calculated using Monte Carlo codes and complex anthropomorphic phantoms. Monte Carlo codes are also widely employed as useful tools during the calibration procedure for in vivo measurements. This paper summarises the role played by Monte Carlo modelling in these fields.

TH-B-I-611-01: Science Behind Proposed ICRP Recommendations: Dosimetric Quantities and Units Used in Radiation Protection

The International Commission on Radiological Protection (ICRP) and the International Commission on Radiation Units and Measurements (ICRU) have developed two types of special dosimetric quantities for radiation protection: those that are related to risk and demonstrate compliance with exposure limits, and operational quantities that are used in monitoring and measuring external exposure and intakes of radionuclides. All are based on mean absorbed dose with its distributions in time and in linear energy transfer (linear collision stopping power). In Publication 26 (1977), ICRP introduced the concepts of dose equivalent and effective dose equivalent to take into account the varying effectiveness of different radiations (the Q factors) and the variations in radiation sensitivity of different tissues (6 were specifically considered) for the induction of stochastic effects (the weighting factors, wT). In Publication 60 (1991), the ICRP introduced equivalent dose, effective dose and the subsidiary quantities committed dose and collective dose. Q factors were superseded by wR values and the number of tissues was increased to 12, with new values for wT. In 2005, the ICRP proposes the terms radiation‐weighted dose, effective dose, committed dose and collective effective dose. Based on new scientific evidence, wR values are unchanged for photons and alphas, but have changed for neutrons and protons. wT are different for the gonads, the breast an the “remainder” (the treatment of which has also changed); the number of tissues has increased to 14. The operational quantities for external exposure, defined by ICRU (1993, 2001), include ambient dose equivalent and directional dose equivalent for area monitoring, and personal dose equivalent for individual monitoring. For internal exposure, the ICRP proposes to use activity quantities in combination with dose coefficients based on physiological models and 4‐D computations. The unit for all the quantities listed above is the Sievert (although the ICRP would like a different unit for the radiation‐weighted dose). The weighting factors and the dosimetric quantities based on wR and wT relate only to stochastic health effects, cancer induction and hereditary disease. The risk factors, from which wR and wT values are derived, have been obtained from epidemiological and experimental radiobiological data in the medium and higher dose ranges and have been extrapolated to the lower dose ranges using the linear no threshold model, which bears a high degree of uncertainty. The assumed linearity and additivity are, however, necessary conditions for all the dosimetric considerations. Effective dose is to be used by regulators for occupationally exposed workers and members of the public, where doses are assumed to be low. At doses above about 0.5–1 Sv, tissue reactions (deterministic effects) may occur. At these levels, the dosimetric quantity to use is the absorbed dose in the irradiated tissue modified by the radiobiological effectiveness of the radiation for the biological endpoint of concern. The unit is the Gray. Effective dose should not be used for retrospective evaluation of exposed populations or to assess individual risks, as is the case in medical exposures, which are not subject to dose limitations. Educational Objectives: 1. To understand the dosimetric concepts involved in radiation protection. 2. To understand the uses of and the limitations and uncertainties associated with the dosimetric quantities.

Dose quantities in radiation protection and their limitations

For more than 50 years the quantity absorbed dose has been the basic physical quantity in the medical applications of ionising radiation as well as radiological protection against harm from ionising radiation. In radiotherapy relatively high doses are applied (to a part of the human body) within a short period and the absorbed dose is mainly correlated with deterministic effects such as cell killing and tissue damage. In contrast, in radiological protection one is dealing with low doses and low dose rates and long-term stochastic effects in tissue such as cancer induction. The dose quantity (absorbed dose) is considered to be correlated with the probability of cancer incidence and thus risk induced by exposure. ICRP has developed specific dosimetric quantities for radiological protection that allow the extent of exposure to ionising radiation from whole and partial body external radiation as well as from intakes of radionuclides to be taken into account by one quantity. Moreover, radiological protection quantities are designed to provide a correlation with risk of radiation induced cancer. In addition, operational dose quantities have been defined for use in measurements of external radiation exposure and practical applications. The paper describes the concept and considerations underlying the actual system of dose quantities, and discusses the advantage as well as the limitations of applicability of such a system. For example, absorbed dose is a non-stochastic quantity defined at any point in matter. All dose quantities in use are based on an averaging procedure. Stochastic effects and microscopic biological and energy deposition structures are not considered in the definition. Absorbed dose is correlated to the initial very short phase of the radiation interaction with tissue while the radiation induced biological reactions of the tissue may last for minutes or hours or even longer. There are many parameters other than absorbed dose that influence the process of cancer induction, which may influence the consideration of cells and/or tissues at risk which are most important for radiological protection.

Standards for standard-makers? A testing time.

The International Commission on Radiological Protection (ICRP) is now reviewing its recommendations with a view to publishing their revision in 2005. The last set of recommendations issued by the ICRP has caused some concern to neutron dosimetrists. This paper attempts to explain these concerns. Technological developments make it likely that exposure to high-linear energy transfer (LET) radiations will increase in the future. It is in the area of the dosimetry of high-LET radiations, particularly neutrons, where some experts feel that ICRP recommendations have been unclear. This paper discusses the process of setting protection limits in toxicology and its application to radiation protection. The development of radiation protection quantities and models is described, and the problems found with effective dose described. Suggestions for improvements are made that would enable effective dose to serve in two modes--both as a limiting quantity and also as the measurable (operational) quantity required by dosimetrists.

Adaptation of the present concept of dosimetric radiation protection quantities for external radiation to radiation protection practice.

The present concept of dosimetric radiation protection quantities for external radiation is reviewed. For everyday application of the concept some adaptations are recommended. The check of the compliance with dose limits should be performed either by the comparison with values of the respective operational quantities directly or by the calculation of the protection quantity by means of the operational quantity, the appertaining conversion coefficient and additional information of the radiation field. Only four operational quantities are regarded to be sufficient for most applications in radiation protection practice. The term equivalent should be used in the connection dose equivalent only. Proposals are made for names of frequently used operational quantities which are denoted up to now by symbols only.

Concepts and quantities in spectrometry and radiation protection.

Concepts and quantities in spectrometry and radiation protection.

Long-term monitoring of the onboard aircraft exposure level with a Si-diode based spectrometer

The radiation fields onboard aircraft are complex (EURADOS, 1996), and several methods are used to characterise them for radiation protection. We have tested a spectrometer based on Si-diode at different sources and accelerator facilities. The energy deposited in the diode is analysed to estimate the contribution of different radiations to dosimetry quantities. The spectrum of energy deposition events onboard aircraft is similar to that registered in the CERN high-energy reference field. We used this similarity to determine the correction factors to appreciate radiation protection quantities from the results of onboard measurements. During 2001–2002, the spectrometer was used to acquire measurements onboard commercial aircraft during five long-term exposures. All necessary flight parameters were acquired; thus permitting calculations of the onboard effective dose and/or ambient dose equivalent by means of both the CARI6 and the EPCARD codes and comparison with the results of the measurements. It was found that the apparent ambient dose equivalent values from measured data are in reasonable agreement with the results of calculations. Quantitative analysis of this agreement as a function of flight parameters (geomagnetic position, solar activity variations, etc.) is presented. During one flight, an important solar event GLE 60 on 15 April 2001 was recorded by the spectrometer. In some other cases the measurements during a Forbush decreases were acquired. These extremes were well registered by the equipment and the data obtained are analyzed.

The Impact of ICRP/ICRU Quantitites on High Energy Neutron Dosimetry: A Review

Unlike other fields of toxicology, radiation protection has a dual system of quantities, one set for assessment and the derivation of authorised limits and another set for monitoring radiation performance and compliance. Neutrons are an important or dominant constituent of the radiation field around high energy accelerators and the evolution of the radiation protection quantities used to measure neutrons is described. In 1990 ICRP introduced a new quantity, the effective dose. E. with which to express its protection limits. E represented a radical departure from previous advice of the Commission, particularly in the manner by which it weighted the absorbed dose deposited by high LET radiations. This advice had profound consequences for neutron dosimetry. Over the past decade analyses have revealed logical flaws and inconsistencies in the definition of effective dose. These are briefly discussed with most emphasis being placed on inconsistencies in radiation weighting. Suggestions are made with a view to resolving these inconsistencies.

Dose quantities in radiation protection and dosemeter calibration.

Dosimetric concepts and the definition of dose quantities for use in radiation protection were defined by the ICRP and the ICRU. Three types of quantities are of relevance for radiation protection purposes: basic physical quantities, protection quantities, and operational quantities. The physical quantities are universally accepted for the characterisation of radiation fields. These are defined in any point of the field, and the units are directly obtained by primary standards. The protection quantities form the basis for dose limitation. These quantities are not directly measurable. A set of measurable operational quantities provides, in general, a conservative estimate of the protection quantities. These operational quantities are recommended in routine monitoring of occupational exposure. The relation between these different types of quantities is discussed. In addition, the calibration procedure for dosemeters in terms of the operational quantities, including the use of calibration phantoms, is summarised.

In radiological protection, the protection quantities should be expressed in terms of measurable physical quantities.

It is somewhat of a truism that is expressed in the often quoted statement that you can't go home again. In the days when life was simpler, we could assume that a measurement performed with a portable survey instrument yielding a value for exposure rate in terms of roentgens/hour was a sufficient approximation to the value for the dose equivalent rate or absorbed dose rate at a point in a radiation field. As our understanding of radiological protection grew more sophisticated, we found it necessary to introduce additional rigor into the definitions of the radiation protection quantities. As mentioned earlier, we demand a lot from the definitions of these quantities. They need to satisfy many conditions, perhaps too many. When they are examined from the point of view of a metrologist, they may be considered to be deficient in many respects. When they are viewed by a radiobiologist, they may appear to be naïve or overly simplistic. A radiation protection technologist at a nuclear power plant may have difficulty understanding their sometimes-subtle implications. The ICRU and ICRP have worked for quite some time to incorporate all the necessary requirements into the definitions of the operational and protection quantities. From the arguments of our debaters, one might conclude that after careful consideration of past developments, additional refinements will be devised in the future.

Effective Doses and Organ Doses per Unit Fluence Calculated for Monoenergetic 0.1 MeV to 100 MeV Electrons by the MIRD-5 Phantom

To determine quantities for radiation protection against high energy electrons, such as those from LINAC, the organ doses and effective doses were calculated using EGS-4 Monte Carlo simulation code and a MIRD-5 mathematical human phantom. Conversion coefficients for the incident fluence to the organ dose equivalent, to the effective dose, and to the effective dose equivalent for monoenergetic electrons were obtained for AP and PA irradiation geometries at incident energies from 0.1 to 100 MeV. The agreement between the results of the present study of effective doses and those of other studies is acceptable within the statistical uncertainties. At energies lower than 10 MeV, the skin sensitive layer was taken into account.

Neutron Dose Equivalent Conversion Coefficients Have Changed in the Last Forty Years...Haven't They?

Conversion coefficients relating neutron fluence to dose equivalent were first made available forty years ago, and subsequently many versions of the coefficients have been published in the scientific literature. These conversion coefficients have been tabulated in a number of formats that related, for example, the fluence rate to maximum dose equivalent rate; the fluence rate, in a period of 40 h, producing a maximum dose equivalent of 1 mSv; and the fluence to ambient dose equivalent. The dose equivalent quantities have been specified in a number of phantoms, including: semi-infinite slabs of tissue-equivalent materials, right-circular and elliptical cylinders, and the ICRU sphere. Significant changes in the definitions of the radiation protection quantities have taken place. Interestingly, the neutron conversion coefficients have remained nearly invariant. It appears that the current set of conversion coefficients are adequate for radiation protection dosimetry and that, barring any unexpectedly large changes in neutron cross section or risk data, further changes are neither necessary nor desirable in the foreseeable future.

Editorial - Variables and Quantities in Radiation Protection

Editorial - Variables and Quantities in Radiation Protection

Radiological Impact of Low Level Solid Radioactive Waste Disposed of with Ordinary Hospital Refuse

A method for the measurement of radioactivity contained in cardboard boxes used to store hospital refuse was developed, testing its reliability with sources of known activity and geometry. The method was applied to a random sample of about 2000 boxes taken from those which are not considered to contain radioactive material and are disposed of without checks, in order to evaluate the amount of activity which is released into the environment. It was found that it is about 0.06% of the total activity administered to patients. Radiation protection quantities were calculated on this basis, showing that the radiological effect is small, not requiring routine analysis of all waste boxes produced in the hospital, but suggesting that measurements on a random sample are advisable to monitor the process.