Dosimetry for Use in Preparedness and Response to Radiological and Nuclear Emergency

2.1 Basic physical quantities

The basic dosimetric quantities include the particle fluence ΦR, the kerma KR, and the absorbed dose DRof radiation R. Basic quantities characterize the field of radiation and its interaction with a medium where the human body could be present:

1. The fluence, ΦR, is the quotient of dNRby ds, where dNRis the number of particles R incident on a sphere of cross-sectional area ds. Unit of fluence is m⁻².

2. The kerma, KR, for ionizing uncharged particles R, is the quotient of dEtrby dm, where dEtris the mean sum of the initial kinetic energies of all the charged particles liberated in a mass dm of a material by the uncharged particles incident on dm. Unit of kerma is J/kg. The special name for the unit of kerma is gray (Gy).

3. The absorbed dose, DR, is the quotient of dε¯by dm, where dε¯is the mean energy imparted by ionizing radiation to matter of mass dm. Unit of absorbed dose is J/kg. The special name for the unit of absorbed dose is gray (Gy).

The ICRU provides exact definition of these quantities in [3].

2.2 Protection quantities

Protection quantities are the dosimetric quantities, which characterize the irradiation of the human. GSR Part 7 [1] states that in response to emergency, “consideration shall be given to actions to be taken to avoid or to minimize severe deterministic effects and to reduce the risk of stochastic effects. Deterministic effects shall be evaluated on the basis of relative biological effectiveness (RBE) weighted absorbed dose in tissue or organ. Stochastic effects in a tissue or organ shall be evaluated on the basis of equivalent dose in tissue or organ. The detriment associated with the occurrence of stochastic effects in individuals in an exposed population shall be evaluated on the basis of the effective dose” (para. 4.28 of GSR Part 7 [1]). Dosimetric quantities of RBE-weighted absorbed dose in tissue or organ T, ADT; equivalent dose in tissue or organ T, HT; and effective dose E are defined in GSG-2 [4] and GSR Part 3 [5] and recommended for protection purposes.

The overall theory on the development of biological health effects is not yet developed. The protection quantities are essentially used for evaluation of consequences of human exposure to ionizing radiation in terms of developing deterministic and stochastic health effects separately as listed in Table 1. Figure 1 presents relationship between protection quantities and basic physical quantities.

Radiation protection quantities could not be measured directly. In particular irradiation conditions, they could be calculated by taking into account estimates of basic protection quantities (results of their measurement or calculation), geometry of irradiation, characteristics of irradiated person, etc. For radiation protection purposes, these quantities are defined for reference persons representing different groups of the public and workers in [6] and standard irradiation geometries for external exposure in [7].

Definitions of these quantities can be found in the GSR Part 3 [5], 2007 Recommendations of the International Commission on Radiological Protection (ICRP) [8], and EPR Publication [9].

The determination of RBE-weighted dose ADTin a tissue or organ T involves the use of tissue-specific and radiation-specific factors RBET,Ras a multiplier of absorbed dose for radiation R, to reflect the relative biological effectiveness of the radiation in inducing severe deterministic effect in organ T at high doses:

where DT,Ris the average absorbed dose in the tissue or organ T for radiation R.

Factor RBET,Rdepends on the quality of radiation and macroscopic distribution of energy of radiation imparted to matter of affected organ or tissue [9]. The role of such heterogeneity is significant in the case of internal exposure [9, 10]. The RBE-weighted absorbed dose in tissue or organ is defined for use together with radiobiological model for assessing the risk of developing severe deterministic effects in critical organs and tissues at high doses [9, 10, 11, 12].

The determination of equivalent dose HTin a tissue or organ T involves the use of a radiation weighting factor wRas a multiplier of absorbed dose for radiation R, to reflect the relative biological effectiveness of the radiation in inducing stochastic effects at low doses:

The equivalent dose in tissue or organ is defined for use together with radiobiological model for assessing the risk of developing stochastic effects in organs and tissues at low doses of external [13] and internal [14] exposure.

The determination of effective dose E involves the use of a tissue weighting factor wTas a multiplier of equivalent dose for tissue T, to account for the different sensitivities of different tissues or organs to the induction of stochastic effects and production of radiation detriment [8, 15]:

In accordance with the ICRP, the radiation detriment is the total harm to health experienced by an exposed group and its descendants as a result of the group’s exposure to a radiation source. Detriment is a multidimensional concept. Its principal components are the stochastic quantities: probability of attributable fatal cancer, weighted probability of attributable nonfatal cancer, weighted probability of severe heritable effects, and length of life lost if the harm occurs [8]. The 2007 Recommendations of the ICRP state in para. (153) that the main and primary uses of effective dose in radiological protection for both occupational workers and the general public are:

• prospective dose assessment for planning and optimization of protection; and

• retrospective dose assessment for demonstrating compliance with dose limits, or for comparing with dose constraints or reference levels.

In this regard, organ or tissue equivalent doses, not effective doses, are required for assessing the probability of cancer induction in exposed individuals.

The recommended values of RBET,R,wRand wTare based on a review of published biological and epidemiological studies and are given in the definitions of protection quantities in GSR Part 3 [5].

2.3 Operational quantities

Since radiation protection quantities cannot be measured directly, the ICRU introduced operational quantities for practical use in radiation protection where exposure due to external sources is concerned. Definitions of these quantities can be found in [16, 17, 18, 19, 20] and in GSR Part 3 [5]. The operational quantities provide an estimate of effective or equivalent dose in tissue or organ in such a way that avoids overestimation in most radiation fields encountered in practice [7]. Radiation quality factor Q(L) is used in calculating the operational dose equivalent quantities used in monitoring [21]. The quality factor characterizes the biological effectiveness of the radiation type, based on the ionization density along the tracks of charged particles in tissue. Q is defined as a function of the unrestricted linear energy transfer, L_m (often denoted as L or linear energy transfer, LET), of charged particles in water. Detailed evaluation of the relationship between the physical, protection, and operational quantities was conducted by a joint task group of the ICRP and ICRU [7].

Strongly penetrating radiation and weakly penetrating radiation are considered in radiation dosimetry and are differentiated as follows. For most practical purposes, it may be assumed that strongly penetrating radiation includes photons of energy above about 12 keV, electrons of energy more than about 2 MeV, and neutrons. It may be also assumed that weakly penetrating radiation includes photons of energy below about 12 keV, electrons of energy less than about 2 MeV, and massive charged particles as protons and alpha particles [5].

The operational quantity for individual monitoring is the personal dose equivalent HPd. Any statement of personal dose equivalent has to include a specification of the reference depth d. In order to simplify the notation, d is assumed to be expressed in millimeters.

For strongly penetrating radiation, the reference depth for controlling the radiation detriment in planned exposure situation is 10 mm. Personal dose equivalent HP10provides a conservative estimate of effective dose to adult for strongly penetrating radiation.

For weakly penetrating radiation, the reference depth for controlling stochastic effects due to irradiation of the basal membrane of the skin is 0.07 mm, and the deterministic effect in lens of the eye is 3 mm [15, 22, 23].

For monitoring of the lens of the eye, a depth of 3 mm is recommended by the International Commission on Radiation Units and Measurements [3], so the operational quantity to be used is HP3. In practice, however, the use of HP3has not yet been implemented for routine individual monitoring. In specific cases, when actual workplace radiation fields are known, monitoring of the eye through dosimeters calibrated for HP0.07could be acceptable. In [23], it is stated that HP0.07can be considered a good operational quantity for the lens of the eye for exposures to fields for which most of the dose is due to photons, including X radiation. In such cases, it has to be borne in mind that the uncertainty associated with the estimation of equivalent dose will be higher.

The operational quantities recommended for workplace monitoring are defined in a phantom known as the ICRU sphere [18]. This is a sphere of 30 cm diameter made of tissue equivalent material with a density of 1 g/cm³ and an elemental composition (by mass) of 76.2% oxygen, 11.1% carbon, 10.1% hydrogen, and 2.6% nitrogen.

The two quantities recommended by the International Commission on Radiation Units and Measurements for workplace monitoring [3] are the ambient dose equivalent H∗dand the directional dose equivalent H′dΩ.

The ambient dose equivalent H∗dat a point in a radiation field is the dose equivalent that would be produced by the corresponding aligned and expanded field in the ICRU sphere, at a depth d on the radius opposing the direction of the aligned field.

The expanded field (see Figure 2(b)) is one in which the fluence and its angular and energy distribution are the same throughout the volume of interest as in the actual field at the point of reference (see Figure 2(a)). In the expanded and aligned field (see Figure 2(c)), the fluence and its energy distribution are the same as in the expanded field, but the fluence is unidirectional [12].

Any statement of ambient dose equivalent has to include a specification of the reference depth d. For strongly penetrating radiation, the recommended depth is 10 mm. When measuring H∗10, the radiation fields have to be uniformed over the sensitive volume of the instrument, and the instrument has to have an isotropic response.

The directional dose equivalent H′dΩat a point in a radiation field is the dose equivalent that would be produced by the corresponding expanded field in the ICRU sphere, at a depth d on a radius in a specified direction Ω. Any statement of directional dose equivalent has to include a specification of the reference depth d and the direction Ωof the radiation. For strongly penetrating radiation and weakly penetrating radiation, the recommended depths are 10 mm and 0.07 mm, respectively.

If the field is unidirectional, the direction Ωis specified as the angle between the radius opposing the incident field and the specified radius. When the specified radius is parallel to the radiation field (i.e., when Ω= 0°) the quantity H′d0is equal to H∗d. When measuring H′dΩ, the radiation fields have to be uniformed over the dimensions of the instrument, and the instrument has to have the appropriate directional response.

4.1 Estimation of protection quantities characterizing an external exposure

Properties of external radiation entering the body of individual R, basic physical characteristics of radiation field, such as particle fluence ΦR, the kerma KR, and geometry of irradiation G, provide a basis for estimation of the protection quantities from external exposure. For idealized irradiation geometries, the ICRP in [7] presents the relationship between basic physical quantities and protection quantities for monoenergetic photons, electrons and positrons, neutrons, protons, muons, pions, and helium ions in six considered irradiation geometries: AP, anteroposterior; PA, posteroanterior; LLAT, left lateral; RLAT, right lateral; ROT, rotational; and ISO, isotropic.

As presented in Figure 4, the major characteristic of a field of radiation is a fluence of radiation ΦR. For monoenergetic radiation, an ambient dose equivalent and air kerma are proportional to fluence:

For non-monoenergetic radiation

where ΦRεRdεRis the fluence of radiation R with energy between εRand εR+dεR, kaεRis the dose coefficient of air kerma of radiation R with energy εR, and h∗10εRis the dose coefficient of ambient dose equivalent of radiation R with energy εR.

For irradiation geometry G, one has

where kaεRis the dose coefficient of air kerma of radiation R with energy εR, hP10εRGis the dose coefficient of personal dose equivalent of radiation R with energy εRand irradiation geometry G, AdTεRGis the dose coefficient of RBE-weighted dose in organ T of radiation R with energy εRand irradiation geometry G, hTεRGis the dose coefficient of equivalent dose in organ T of radiation R with energy εRand irradiation geometry G, and eεRGis the dose coefficient of effective dose of radiation R with energy εRand irradiation geometry G.

For photon radiation, values of AdTεRGand hTεRGare numerically equal when the same organ or tissue and irradiation geometry are considered.

The internet resource [25] provides tools for evaluation of absorbed dose in different organs from point or volumetric radioactive sources inside or outside the human body.

The publication [26] provides dose coefficients for exposure to bulk sources that are ground, water body and cloud containing radioactive material.

The publication [27] presents a compendium of neutron spectra, which could be used for estimation of protection quantities in accordance with Eqs. (7)–(10). The estimates show that for fission neutrons scattered from the concrete walls of the facility, soil, or from the air surrounding the facility (skyshine), the value of AdTεRGis numerically equal to 1/5 of hTεRGwhen the same organ or tissue and irradiation geometry are considered. This linkage provide the possibility to use results of operational or routine monitoring of doses in workers for estimation of ADRedmarrowin emergency exposure situation as required by GSR Part 7 and presented in Table 4.

As presented in [7], dose coefficients mentioned above are proportional to the photon energy in range of (0.1–6) MeV. Therefore, for the same irradiation geometry

where εjis any photon energy from range of (0.1–6) MeV.

For mentioned range of photon’ energy, the exposure is proportional to the kerma free-in-air (air kerma) or exposure. Thus, in the same point of field of photon radiation, exposure or kerma in air for an exposure of 100 R is 0.876 Gy [28]. This linkage provides the possibility to use old devices such as exposure meters (R-meters) in environmental monitoring for estimation of air kerma and protection quantities as given in Eq. (11).

The protection quantities ADT,HTand Ereceived from exposure due to external sources can be also estimated from the operational quantities by using the following equations:

For strongly penetrating radiation, the critical organ for controlling the development of the severe deterministic effects in individual is the red marrow [1, 4]. The ICRU did not recommend depth for controlling the dose in the red marrow. For practical reasons, monitoring of the red marrow through dosimeters calibrated for HP10could be acceptable. The RBE-weighted absorbed dose in the red marrow received from exposure due to external sources can be estimated from the operational quantities by using the following equation:

For weakly penetrating radiation and in emergency exposure situation, the reference depth for controlling severe deterministic effects due to irradiation of the derma of the skin is 0.4 mm and for controlling the severe deterministic effects due to irradiation of shallow soft tissue is 5 mm [1, 4, 9, 10]. The ICRU did not recommend operational quantity for controlling the dose in the skin derma or shallow soft tissue in emergency exposure situation. For practical reasons, monitoring of the skin derma and shallow soft tissue through dosimeters calibrated for HP0.07could be acceptable. The RBE-weighted absorbed dose in the skin derma or in shallow soft tissue received from exposure due to external sources can be conservatively estimated from the operational quantities by using the following equations:

Based on the foregoing, individual monitoring in planned, emergency and existing exposure situations through individual dosimeters calibrated for HP10and HP0.07could be acceptable.

4.2 Estimation of protection quantities characterizing an internal exposure

Internal doses cannot be measured directly; they can only be calculated from intake of radioactive substance through particular route such as respiratory system in the case of inhalation, gastrointestinal tract in the case of ingestion or wound, and undamaged skin in the case of contamination. The individual intake also could not be directly measured and can only be inferred from individual measurements of other quantities, such as measurements of activity in the body or in excretion samples or activity concentration in foodstuff or the environment. In circumstances where individual monitoring is inappropriate, inadequate, or not feasible, the occupational exposure of workers may be assessed on the basis of workplace monitoring and other relevant information such as location and durations of exposure. Individual measurements include measurements made by both direct and indirect methods. Methods for the measurement of activity content in the body, such as whole-body, lung, or thyroid counting, are examples of direct methods. Measurements of activity in collected biological samples or measurements made using personal air sampling are examples of indirect methods.

The conceptual framework for the assessment of doses from individual or environmental measurements is illustrated in Figure 5.

Chemical and physical properties of radioactive substance S containing radionuclide, the route of intake into the human body P, and value of intake IS,Pcomposite a base for estimation of the protection quantities from internal exposure.

The measurable characteristics of internal exposure are quantities of body content or excretion rate of radionuclide M and concentration of radioactive substance in the environmental media CtS.

According to the scheme in Figure 5, the value of the intake of radioactive substance S could be obtained by multiplying the integrated over time the measured concentration of radioactive substance in the environmental media CtSby the appropriate value of vPg:

where vPgis the consumption rate of an individual of age group g through route P and t is the period of consumption.

According to the scheme in Figure 5, the value of the intake of radioactive substance S is obtained by dividing the measured body content or excretion rate M by the appropriate value of mtgS,P:

where mtgS,Pis the fraction of an intake that remains in the body (for direct methods) or that is being excreted from the body (for indirect methods) at time t after the intake. This fraction depends on the radionuclide, its chemical and physical form in substance S, the route of intake P, the age group of age g, and the time t [29, 30].

Special attention has to be paid to the interpretation of bioassay measurements after the use of means for blocking the uptake of radionuclides or for enhancing their excretion, such as the administration of diuretics, laxatives, or blocking or chelating agents, as well as after the removal of contamination and/or surgical intervention at a wound site. These techniques influence and modify the biokinetic behavior of the incorporated radionuclides, thus invalidating the use of the standardized modeling approach for estimating intake and dose from the bioassay measurements.

In such cases, alternative approaches have to be employed, such as discarding data on excretion for excretion samples collected during the period in which excretion rates may be assumed to have been influenced by the treatment or modifying the standard models in order to take into account the effect of the treatment. Examples of analyses performed after the administration of the chelating agent Ca-DTPA (a calcium salt of diethylenetriaminepentaacetic acid) in cases of accidental intakes of actinides can be found in [31, 32, 33]. Bioassay measurements for dose assessment purposes are performed after a certain time period, posttreatment with Ca-DTPA, until the excretion of the radionuclide stabilizes in urine samples. Influence of iodine thyroid blocking on biokinetic of iodine in human body is discussed in [34]. Influence of administration of Prussian blue {Fe₄[Fe(CN)₆]₃} on biokinetic of Cs is discussed in [35].

The specific protection quantities of committed RBE-weighted organ dose ADTΔg, committed equivalent organ dose HTτg, and committed effective dose Eτgare used for evaluation of dose in an internal dosimetry:

where AdTΔgS,Pand hTτgS,Pare respectively the committed RBE-weighted or equivalent dose in the organ or tissue T due to intake of 1 Bq of radionuclide substance S through pathway P, by the group of age g. These dose factors are the time integrals of the relevant dose rates:

where AḋTtgS,Pis the RBE-weighted dose rate in organ or tissue T due to intake of 1 Bq of radionuclide substance S through pathway P and ḣTtgS,Pis the equivalent dose rate in organ or tissue T due to intake of 1 Bq of radionuclide substance S through pathway P. For estimation of committed RBE-weighted dose, Δis taken to be 30 days. The τ is the integration time elapsed after an intake of radioactive substances. When τ is not specified, it will be taken to be 50 years for adults and the time to age 70 years for intakes by children and persons younger than 20.

Various biokinetic models for calculating the values of mtgS,P,AdTΔgS,Pand hTτgS,Phave been developed.

Values of mtgS,Pat selected times for a subset of radionuclides have been reported by the ICRP in graphical and tabular form [36, 37]. A compilation of mtgS,Pby workers in emergency and planned exposure situation is presented in various publications of the IAEA [29, 38]. The Internet resources containing the values of mtgS,Pdifferent radioactive substances are presented in [39].

Special attention has to be paid to the patterns of inhalation intake in nuclear emergency. Chemical form of radionuclides in aerosol particles and their behavior in the human respiratory tract could be very specific [40, 41] and significantly different from these observed in planned exposure situation and presented in Table II.2C of GSR Part 3 [5].

A compilation of dose coefficients AdTΔgS,Pfor intakes of radionuclides by adults is presented in publications of the IAEA [9, 29].

A compilation of dose coefficients eτgS,Pfor ingestion and inhalation intakes of radionuclides by workers is presented in ICRP Publication 119 [42]. A compilation of hTτgS,Pand eτgS,Pfor workers and all age groups of members of the public can also be found in the ICRP database in [43]. Data for fetus or embryo due to inhalation or ingestion of radioactive substance by mother are presented in [44, 45] and for infant in [46, 47]. These dose coefficients are based on the calculation methods and parameters given in ICRP Publication 60 [15]. The current published values of eτgS,Pwill be superseded in due course by new values [6, 36, 42] based on updated biokinetic models and on the methods of calculation and the parameters given in ICRP Publication 103 [8]. A compilation of mtgS,Wond,hTτgS,Woundand eτgS,Woundfor penetration of radioactive substance through wound is presented in [48, 49].

4.3 Estimation of protection quantities characterizing a total exposure

The protection quantities E and HTrelate to the sum of the effective doses or equivalent doses, respectively, received from exposure due to external sources within a given time period and the committed effective doses or committed equivalent doses, respectively, from exposure due to intakes of radionuclides occurring within the same time period.

The total equivalent dose in organ or tissue T received or committed during a given time period in individuals from age group g can be estimated from the operational quantities by using the following equation:

where HP10is the personal dose equivalent from external exposure and HTτgis assessed by Eq. (21).

The total effective dose Ereceived or committed during a given time period in individuals from age group g can be estimated from the operational quantities by using the following equation:

where Eτgis assessed by Eq. (22).

In the calculation of the committed dose from specific radionuclides, allowance may need to be made for the characteristics of the material taken into the body through inhalation, ingestion, and contaminated wound.