Department of Civil and Industrial Engineering
Master of Science in
Nuclear Engineering
S
ETTING UP A NEW PERSONAL DOSIMETRY
SYSTEM AT
SCK•CEN:
TESTING THE
I
NSTADOSE
2
DOSIMETER
Author
Alessio Parisi
Supervised by
Riccardo Ciolini (University of Pisa)
Olivier Van Hoey (SCK•CEN)
Filip Vanhavere (SCK•CEN)
2 I took a heavenly ride through our silence
I knew the moment had arrived For killing the past and coming back to life
3
A
BSTRACT
Personal dosimetry can be done by means of the thermoluminescent detectors (TLD). The nuclear research center SCK•CEN (Dutch: Studiecentrum voor Kernenergie; French: Centre d'Étude de l'énergie Nucléaire) uses a whole body TLD system based on Lithium Fluoride (LiF) with addition of Magnesium and Titanium. This SCK•CEN personal dosimetry system is nearly 30 years old, reaching the end of its life. Because the same system is no longer available on the market, a completely new dosimetry system will be needed.
One of the possible dosimetry system candidates is the one based on Direct Ion Storage (DIS) technology. This is a system in-between passive and active dosimetry and has some remarkable advantages compared to present systems. This dosimeter, commercially known as Instadose 2, can communicate its results through a Bluetooth connection and should no longer be sent back to the dosimetry service.
The objective of this work is to test this system for implementation in routine. This encompasses practical issues, like calibration and quality control, but also more formal type testing according to IEC International Standard 62387 “Radiation protection instrumentation – Passive integrating dosimetry systems for personal and environmental monitoring of photon and beta radiation”. This was done through a study on repeatability, angular and energy response and environmental influences.
The work is subdivided in four chapters. In the first chapter, the Introduction, is shown the different types and sources of ionizing radiation, their health effects on human body and the reason why it is important to measure and control the dose related to ionizing radiation exposure. Furthermore, in the final part of this chapter it is shown an overview of the tests that are typically required to assess the quality of a dosimetry system. The second chapters of this work shows the evolution of passive dosimetry system from TLD to the first DIS-based dosimeter, namely DIS-1, and finally to the new Instadose devices. The third chapter shows an analysis of the results of the tests performed on the Instadose 2 dosimeter. The fourth shows the results of the study on the calibration factor that was performed in order to improve the performance of the Instadose 2 dosimeters.
4
S
UMMARY
ABSTRACT ... 3
PART 1: INTRODUCTION ... 8
1. DEFINITIONS ... 9
1.1 Technical terms and definitions ... 9
1.1.1 Accuracy ... 9
1.1.2 Calibration ... 9
1.1.3 Calibration factor ... 9
1.1.4 Coefficient of variation ... 9
1.1.5 Confidence interval for the mean value ... 10
1.1.6 Confidence interval for a combined quantity ... 10
1.1.7 Conventional true value ... 10
1.1.8 Coverage factor ... 10 1.1.9 Detector... 10 1.1.10 Dosimeter ... 10 1.1.11 Dosimetry system ... 11 1.1.12 Expanded uncertainty ... 11 1.1.13 Influence quantity ... 11
1.1.14 Influence quantity of type F ... 11
1.1.15 Influence quantity of type S ... 11
1.1.16 Mandatory range ... 11 1.1.17 Measurand ... 11 1.1.18 Point of test ... 11 1.1.19 Rated range ... 12 1.1.20 Reader ... 12 1.1.21 Readout ... 12 1.1.22 Reference conditions ... 12 1.1.23 Reference direction ... 12 1.1.24 Reference orientation ... 12
1.1.25 Reference point of a dosimeter ... 12
1.1.26 Reference measurement ... 12
1.1.27 Relative expanded uncertainty ... 12
1.1.28 Standard conditions ... 12 1.1.29 Standard deviation ... 13 1.1.30 Standard uncertainty ... 13 1.1.31 Student’s t-value ... 13 1.1.32 Type test... 14 1.1.33 Uncertainty ... 14 2. PERSONAL DOSIMETRY ... 15 2.1 Physical quantities ... 15 2.1.1 Absorbed Dose ... 15 2.1.2 Exposure ... 15 2.1.3 Fluence ... 16 2.1.4 Kerma ... 16
2.1.5 Linear energy transfer ... 16
2.2 Protection quantities... 16
2.2.1 Mean absorbed dose ... 17
2.2.2 Equivalent dose and radiation weighting factors... 17
2.2.2.1 Photons, electrons and muons ... 18
2.2.2.2 Protons and pions ... 18
5
2.2.2.4 Alpha particles ... 19
2.2.2.5 Fission fragments and heavy ions ... 19
2.2.3 Effective dose and tissue weighting factors ... 19
2.3 Operational quantities ... 20
3. EXPOSURE TO IONIZING RADIATION ... 22
3.1 Types of exposure ... 22
3.1.1 Internal irradiation ... 22
3.1.2 External contamination ... 23
3.1.3 External irradiation ... 23
3.2 Sources of ionizing radiation ... 23
3.2.1 Natural sources ... 23
3.2.1.1 Cosmic radiation ... 24
3.2.1.2 Natural radionuclides ... 24
3.2.1.3 Radon... 25
3.2.2 Artificial sources ... 25
3.2.2.1 Nuclear military test explosions and weapons production ... 25
3.2.2.2 Medical diagnosis ... 26
3.2.2.3 Nuclear fuel cycle ... 27
3.2.2.4 Occupational exposure ... 28
3.2.2.5 Incidents ... 29
4. HEALTH EFFECTS OF IONIZING RADIATIONS ... 30
4.1 Damaging process ... 30
4.2 Factors affecting cellular damaging process ... 32
4.2.1 LET ... 32 4.2.2 Dose rate ... 35 4.2.3 Dose fractionation ... 35 4.3 Health effects ... 36 4.3.1 Deterministic effects ... 36 4.3.2 Stochastic effects ... 39
5. THE NEED FOR PERSONAL DOSIMETRY SYSTEMS ... 40
5.1.1 The goals of radioprotection ... 40
5.1.2 Dose response model ... 40
5.1.3 Dose limits ... 41
5.1.4 Performance requirements for a personal dosimetry system ... 41
PART 2: EVOLUTION OF PERSONAL DOSIMETRY SYSTEMS ... 44
6. LUMINESCENT DETECTORS... 45 6.1 Luminescence ... 45 6.2 Thermoluminescence ... 48 6.2.1 Heating process ... 50 6.2.2 Aging effect ... 50 6.2.3 Linearity ... 50 6.2.4 Fading effect ... 50 6.2.5 Thermoluminescent (TLD) materials ... 52
6.2.6 LiF – Lithium Fluoride ... 53
6.3 OSL - optically stimulated luminescence ... 53
6.4 Albedo luminescent neutron detectors ... 54
7. DIS-1 ... 55
7.1 The construction ... 56
6
7.1.2 DIS element ... 57
7.1.3 DIS Memory Cell ... 57
7.2 The operation ... 58
7.2.1 Resetting the dose ... 58
7.2.2 Reading the dose ... 59
7.3 Laboratory tests ... 59
7.3.1 Repeatability ... 59
7.3.2 Energy dependence ... 60
7.3.3 Angular dependence... 61
7.3.4 Temperature dependence ... 62
7.3.5 Dose rate dependence ... 62
7.3.6 Hard reset dependence ... 62
7.4 Practical tests ... 63
7.4.1 SCK•CEN nuclear research center ... 64
7.4.2 Hospital ... 65 8. INSTADOSE ... 68 8.1 Instadose 1 ... 68 8.1.1 Dose readout ... 68 8.1.2 The construction ... 69 8.2 Instadose 2 ... 71 8.2.1 Elements layout ... 72 8.2.2 The operation ... 72
8.2.3 Resetting the dose ... 72
8.2.4 Dose readout ... 73
8.2.4.1 Instadose App ... 74
8.2.4.2 InstaLink USB ... 74
8.2.4.3 InstaLink HotSpot Stations ... 74
PART 3: TESTS ON INSTADOSE 2 ... 76
9. EXPERIMENTAL SETUP ... 77 10. REFERENCE IRRADIATIONS ... 84 10.1 Hp(10) reference irradiation ... 84 10.2 Hp(0.07) reference irradiation ... 85 11. REPEATABILITY TEST... 87 11.1 Hp(10) results analysis ... 87 11.2 Hp(0.07) results analysis ... 89
11.3 IEC performance requirements ... 92
11.3.1 Coefficient of variation (IEC 62387, par. 11.2) ... 92
11.3.2 Non-linearity (IEC 62387, par. 11.3) ... 93
12. ENERGY RESPONSE TEST ... 104
12.1 Hp(10) results analysis ... 104
12.2 Hp(0.07) results analysis ... 105
12.3 Radiation performance requirements ... 107
12.3.1 Radiation energy (IEC 62387, par. 11.5 and 11.7) ... 107
13. ANGULAR RESPONSE TEST ... 117
7
13.2 Hp(0.07) results analysis ... 118
13.2.1 Cs-137, dosimeters in reference position ... 118
13.2.2 N60, dosimeters in reference position ... 118
13.2.3 N60, dosimeters rotated 90° from reference position ... 119
13.2.4 N30, dosimeters in reference position ... 119
13.3 IEC performance requirements ... 122
13.3.1 Radiation energy and angle of incidence (IEC 62387, par. 11.5 and 11.7) ... 122
14. TEMPERATURE RESPONSE TEST ... 134
14.1 Results analysis ... 134
14.2 IEC performance requirements ... 134
14.2.1 Ambient temperature (IEC 62387, par. 13.2) ... 134
15. BACKWARDS IRRADIATION TEST... 144
15.1 Hp(10) results analysis ... 144
15.2 Hp(0.07) results analysis ... 145
PART 4: OPTIMIZING THE CALIBRATION ... 148
16. DOSIMETERS RE-CALIBRATION ... 149
16.1 Energy response test ... 149
16.2 Angular response test ... 149
CONCLUSIONS ... 155
8
P
ART
1:
9
1. D
EFINITIONS
For the purpose of this work, the following terms and definition apply, as defined in the following documents:
Radiation Protection NO 160 – Technical Recommendations for Monitoring Individuals Occupationally Exposed to External Radiation
ISO 4037 – X and gamma reference radiation for calibrating dosemeters and dose rate meters and for determining their response as a function of photon energy
IEC 62387 – Radiation protection instrumentation – Passive integrating dosimetry systems for personal and environmental monitoring of photon and beta radiation
ICRP 116 – Conversion Coefficients for Radiological Protection Quantities for External Radiation Exposure
ICRP 103 – The 2007 Recommendations of the International Commission on Radiological Protection
1 . 1 T
E C H N I C A L T E R M S A N D D E F I N I T I O N S1.1.1 Accuracy
Accuracy of a measurement is defined as the closeness of agreement between a measured quantity value and a true quantity value of the measurand[1].
1.1.2 Calibration
Quantitative determination, under a controlled set of standard test conditions, of the reading given by a dosimeter as a function of the quantity to be measured[2].
1.1.3 Calibration factor
N0, quotient of the conventional true value of a quantity and the indicated value under
reference conditions[3].
𝑁0 = 𝐶𝑟,0
𝐺𝑟,0 (1.1)
1.1.4 Coefficient of variation
v, ratio of the standard deviation to the arithmetic mean of a set of n values[3].
𝑣 = 𝑠
10
1.1.5 Confidence interval for the mean value
The confidence interval for the mean value 𝑥 is:
(𝑥 − 𝑈𝑖 ; 𝑥 + 𝑈𝑖) (1.3)
where Ui is the half-width of the confidence interval of the mean value. When calculating 𝑥
from n measurements the half-width of the confidence interval at a confidence level of 95% is given by:
𝑈𝑖 =
𝑡𝑛−1
√𝑛 𝑠 (1.4)
where s is the standard deviation and 𝑡𝑛−1 is the student’s t-value for a double sided
confidence level of 95%[3].
1.1.6 Confidence interval for a combined quantity
Suppose the mean value of w quantities 𝑥𝑖 and the half-width of the corresponding confidence interval Ui, then the half-width of the combined quantity is given by the equation 1.5[3].
𝑈𝑐𝑜𝑚 ≈ √∑ (𝜕𝑥 𝜕𝑥𝑖 𝑈𝑖) 2 𝑤 𝑖=1 (1.5)
1.1.7 Conventional true value
C, value attributed to a particular quantity and accepted, sometimes by convention, as having an uncertainty appropriate for a given purpose[3].
1.1.8 Coverage factor
k, numerical factor used as multiplier of the combined standard uncertainty in order to obtain an expanded uncertainty. A coverage factor is typically in the range 2 to 3[3].
1.1.9 Detector
Apparatus or substance used to convert incident ionizing radiation energy into a signal suitable for indicatin or measurement. A dosimeter comprises at least one detector[3].
1.1.10 Dosimeter
11
1.1.11 Dosimetry system
Dosimeter, reader and all the associated equipment and procedures used for assessing the indicated value[3].
1.1.12 Expanded uncertainty
U, quantity defining an interval about the result of a measurement that may be expected to encompass a large fraction of the distribution values that could reasonably be attributed to the measurand. It is obtained by multiplying the standard uncertainty by a coverage factor[3].
1.1.13 Influence quantity
Quantity that is not the measurand but that affects the results of the measurement, for instance radiation energy, angle of the incident radiation, ambient temperature and relative humidity[3].
1.1.14 Influence quantity of type F
Influence quantity whose effect on the indicated value is a change in response. F stands for factor because the indication due to radiation is multiplied by a factor due to the influence quantity. The radiation energy and the angle of the incident radiation are examples of influence quantity of type F[3].
1.1.15 Influence quantity of type S
Influence quantity whose effect on the indicated value is a deviation independent of the indicated value. S stands for sum because the indication is the sum of the indication due to radiation and due to the disturbance. The electromagnetic disturbance is an example of influence quantity of type S[3].
1.1.16 Mandatory range
Smallest range specified for an influence quantity or instrument parameter over which the dosimetry system must operate to be in compliance with the IEC International Standards[3].
1.1.17 Measurand
It is the quantity intended to be measured[1].
1.1.18 Point of test
Point in the radiation field at which the conventional true value of the quantity to be measured is known[3].
12
1.1.19 Rated range
Specified range of values which an influence quantity can assume without causing a deviation or variation of the response exceeding specified limits[3].
1.1.20 Reader
Instrument used to read one or more detectors in a dosimeter[3].
1.1.21 Readout
Process of measuring the stored dose information of a detector in a reader[3].
1.1.22 Reference conditions
Set of specified values and/or range of values under which the uncertainties admissible for a dosimetry system are the smallest[3].
1.1.23 Reference direction
Direction, in the coordinate system of a dosimeter, with respect to which the angle to the direction of the radiation incidence is measured in unidirectional field[3].
1.1.24 Reference orientation
Orientation for which the direction of the incident radiation coincides with the reference direction of the dosimeter[3].
1.1.25 Reference point of a dosimeter
Physical mark or marks on the outside of the dosimeters, possibly described in the manual, to be used in order to position it with respect to the point of test[3].
1.1.26 Reference measurement
M0, measurement done under reference conditions[3].
1.1.27 Relative expanded uncertainty
Urel, expanded uncertainty divided by the measurement result[3].
1.1.28 Standard conditions
Range of values of a set of influence quantities under which a calibration or a determination of response is carried out. Ideally, calibrations should be carried out under reference conditions. As this is not always achievable or convenient, a small interval around the
13 reference values may be used. During type tests, all values of influence quantities which are not the subject of the test are fixed within the interval of the standard condition[3].
1.1.29 Standard deviation
s, for a series of n measurement of the same measurand, this quantity characterizes the dispersion of the results.
𝑠 = √ 1 𝑛 − 1 ∑(𝐺𝑖 − 𝐺) 2 𝑛 𝑖=0 (1.6)
where Gi is the result of the i-th measurement and 𝐺 is the arithmetic mean of the n results[3].
1.1.30 Standard uncertainty
u, uncertainty of the result of a measure expressed as standard deviation[3].
1.1.31 Student’s t-value
The student’s t-values as a function of the number of measurements is given in Table 1.1[3].
n 𝒕𝒏−𝟏 𝒕𝒏−𝟏 √𝒏 2 12.71 8.98 3 4.30 2.48 4 3.18 1.59 5 2.78 1.24 6 2.57 1.05 7 2.45 0.925 8 2.36 0.836 9 2.31 0.769 10 2.26 0.715 15 2.14 0.554 20 2.09 0.468 25 2.06 0.413 30 2.05 0.373 40 2.02 0.320 60 2 0.258 120 1.98 0.181 > 120 1.96 1.96 / √𝑛
14
1.1.32 Type test
Conformity test made on one or more items representative of the production[3].
1.1.33 Uncertainty
Uncertainty of measurement is the parameter associated with the result of a measurement, that characterizes the dispersion of the values that could reasonably be attributed to the measurand[1].
15
2. P
ERSONAL DOSIMETRY
One of the fundamental requirements in radioprotection is the measurement and the assessment of radiation doses. The process of assessment of doses in individual and area monitoring takes the name of dosimetry and it is necessary to demonstrate compliance with dose limits and meet regulatory requirements. In the course of the discussion on dosimetry and dosimetry systems included in this and the following chapters, different quantities related to radiological protection will be needed.
Protection quantities have the purpose of specifying exposure dose limits to ensure that the occurrence of health effects is kept below unacceptable levels. Operational quantities were defined for practical measurements, both for area and individual monitoring, for the purpose of demonstrating agreement with those limits. These two sets of quantities are related to physical quantities as shown in the next sections.
2 . 1 P
H Y S I C A L Q U A N T I T I E S2.1.1 Absorbed Dose
The absorbed dose D represents the mean energy imparted to matter per unit mass by ionizing radiation. It is the basic physical dose quantity and it is used for all types of ionizing radiation and for any irradiation geometry. The absorbed dose D is defined by the quotient of 𝑑𝜀 by dm, where 𝑑𝜀 is the mean energy imparted by ionizing radiation to the matter of mass dm.
𝐷 = 𝑑𝜀
𝑑𝑚 (2.1)
The SI unit is joules per kilogram and its special name is gray (Gy)[4].
2.1.2 Exposure
The exposure X is defined as the ratio of the absolute value of the total charge dQ of the ions of one sign produced in air when all the secondary particles liberated or created by uncharged particles in air of mass dm are stopped.
𝑋 = 𝑑𝑄
𝑑𝑚 (2.2)
16
2.1.3 Fluence
The fluence 𝜙 is defined as the number of particles incident per unit of surface. It can be expressed as the ration of the number of particles dN incident on a sphere of cross-sectional area dA, so:
𝜙 = 𝑑𝑁
𝑑𝐴 (2.3)
The unit of fluence is m-2. Fluence is independent of the directional distribution of the particles[4].
2.1.4 Kerma
The transfer of energy from indirectly ionizing uncharged particles (such as photons and neutrons) to matter happens through the liberation and the slowing of secondary particles in the matter. This phenomenon leads to the definition of the quantity kinetic energy released in matter, kerma. Kerma is defined, for uncharged particles, as the ratio of the mean sum of the initial kinetic energies dEtr of all the charged secondary particles liberated in a mass dm of
material by the uncharged particles incident on dm:
𝐾 = 𝑑𝐸𝑡𝑟
𝑑𝑚 (2.4)
The unit of kerma is J/kg, and its special name is gray (Gy)[4].
2.1.5 Linear energy transfer
The linear energy transfer or LET is defined as the quotient of dE by dl, where dE is the mean energy lost by the charged particle due to electronic interactions in traversing a distance dl, thus:
𝐿𝐸𝑇 = 𝑑𝐸
𝑑𝑙 (2.5)
The unit of linear energy transfer is joule per meter (J/m), but often it is given in keV/µm[4].
2 . 2 P
R O T E C T I O N Q U A N T I T I E SFor the purpose of limiting exposure to ionizing radiation, the absorbed dose is averaged over specific organs and tissue, weighted by the biological effectiveness of different types of radiation to obtain the equivalent dose to an organ or tissue, and compounded into the quantity effective dose, taking into account the radiosensitivity of different organs and tissues.
17
2.2.1 Mean absorbed dose
The quantity absorbed dose D is defined to give a specific value at any point in matter, but for practical applications absorbed dose is often averaged over larger tissue volume. The mean absorbed dose in a region of an organ or tissue T is defined by:
𝐷𝑇 = 1 𝑚𝑇 ∫ 𝐷 𝑑𝑚 𝑚𝑇 (2.6)
where mT is the mass of the organ of tissue and D is the absorbed dose in the mass element
dm.
The SI unit is joules per kilogram and its special name is gray (Gy)[4].
2.2.2 Equivalent dose and radiation weighting factors
The equivalent dose HT is a physical quantity that quantifies the biological effects of the
absorption of ionizing radiation. It is based on the physical quantity absorbed dose, but takes into account the biological effectiveness of the radiation, which is dependent on the radiation type and energy.
Equivalent dose is calculated using the mean absorbed dose deposited in body tissue or organ T, multiplied by the radiation weighting factor WR which is dependent on the type and energy
of the incident radiation R.
𝐻𝑇 = ∑ 𝑊𝑅 𝐷𝑇,𝑅 𝑅
(2.7)
where HT is the equivalent dose absorbed by tissue T, DT,R is the absorbed dose in tissue T by
radiation type R. The SI unit is joules per kilogram and its special name is sievert (Sv).
The radiation weighting factor represents the relative biological effectiveness of the radiation and modifies the absorbed dose to take account of the different biological effectiveness of various types of radiation.
The ICRP has assigned radiation weighting factors to specified radiation types dependent on their relative biological effectiveness. These weighting factor for the different types of radiation and energies are shown in Table 2.1[4].
18
Radiation Weighting Factor
Photons 1
Electrons and muons 1
Protons and charged pions 2
Neutrons A continuous function of neutron energy Alpha particles, fission fragments and heavy ions 20
Table 2.1 – Weighting factors for different types of radiation and energies[5]
2.2.2.1 Photons, electrons and muons
Electrons, muons and secondary particles generated by photons are radiation with LET values generally less than 10 keV/µm. These radiation types have always been given a radiation weighting of 1[4].
2.2.2.2 Protons and pions
For radiological protection purposes, a single weighting factor is given with value of 2 for protons of all energies. Pions are negatively or positively charged or neutral particles resulting from interaction of primary cosmic rays with nuclei. These particles contribute to exposures in space, aircrafts and radiation fields behind shielding of high-energy accelerators. A single value of 2 is recommended as weighting factor for pions of all energies[4].
2.2.2.3 Neutrons
The biological effectiveness of neutrons incident on the human body depends strongly on neutron energy. For this reason, the weighting factor for neutrons is defined as function of their energy as shown in equation 2.8 and Figure 2.1.
Figure 2.1 - Radiation weighting factor for neutrons as function of neutron energy[5] 𝑤𝑅= { 2.5 + 18.2 exp (−[ln(𝐸𝑛)] 2 6 ) 5 + 17 exp (−[ln(2𝐸𝑛)] 2 6 ) 2.5 + 3.25 exp (−[ln(0.04𝐸𝑛)] 2 6 ) (2.8)
19 2.2.2.4 Alpha particles
A single weighting factor of 20 is recommended for alpha particles. Exposure from alpha particles are usually more relevant in the case of internal emitters (for instance due to inhalation of radon progeny, ingestion of radionuclides such plutonium, polonium, radium, thorium and uranium). In the case of external exposure, alpha particles are less importance because of their short range in air and in tissue[4].
2.2.2.5 Fission fragments and heavy ions
As in the case of alpha particle, the doses from fission fragments are relevant in radiological protection mainly in the case of in internal emitters. For external exposure, the situation regarding the weighting factor is similar to that for alpha particles: a value of 20 is recommended for all types energies of heavy charged particles as conservative estimate. For application in space, where heavy charged particles contribute in a relevant way to the dose in the human body, a different approach should be used in order to assess the radiobiological effects[12,13].
2.2.3 Effective dose and tissue weighting factors
E, the effective dose is the tissue-weighted sum of the equivalent doses in all specified tissues and organs of the body. It quantifies the stochastic health risk of incident ionizing radiation delivered to those body parts. It takes into account the type of radiation, in the equivalent dose, and the nature of each organ or tissue being irradiated.
To calculate the effective dose E, the absorbed organ dose DT is first corrected for the
radiation type using factor WR and the result is further corrected for the tissues or organs being
irradiated using factor WT, to produce the effective dose quantity E.
𝐸 = ∑ 𝑊𝑇 𝐻𝑇
𝑇
(2.9)
where HT is the equivalent dose absorbed by tissue T and WT is the weighting factor for the
tissue T. The SI unit for effective dose is the sievert (Sv) which is one joule/kilogram (J/kg). The ICRP tissue weighting factors are given in the Table 2.2, and the equations used to calculate from either absorbed dose or equivalent dose are also given[5]. If only part of the body is irradiated, then only those regions are used to calculate the effective dose.
20 The tissue weighting factors sum to 1.0, so that if an entire body is irradiated with uniformly penetrating external radiation, the effective dose for the entire body is equal to the equivalent dose for the entire body. Some tissues like bone marrow are particularly sensitive to radiation, so they are given a weighting factor that is disproportionally large relative to the fraction of body mass they represent. Other tissues like the hard bone surface are particularly insensitive to radiation and are assigned a disproportionally low weighting factor[12,13].
Organs Weighting factor
Red bone marrow 0.12
Colon 0.12 Lung 0.12 Stomach 0.12 Breasts 0.12 Gonads 0.08 Bladder 0.04 Liver 0.04 Esophagus 0.04 Thyroid 0.04 Skin 0.01 Bone surface 0.01 Salivary glands 0.01 Brain 0.01
Remainder of the body 0.12
Total 1
Table 2.2 – Weighting factors for different organs[5]
2 . 3 O
P E R A T I O N A L Q U A N T I T I E SBecause the radiation protection quantities cannot be measured, operational quantities have been defined for practical measurements in area and individual monitoring. The operational quantities are defined at an appropriate point (differently from protection quantities which are defined as averages over an extended mass) and are aimed at providing an estimate or upper limit for the value of the protection quantities related to an exposure of persons under most irradiation conditions. They can be calculated using the knowledge of the fluence at the point of interest, allowing radiation monitoring devices (for instance personal dosimeters) to be calibrated in terms of these quantities. For these reasons, they are normally used in regulation. In the case of individual monitoring, the recommended quantity is the personal dose equivalent.
21 With Hp(d) is indicated dose equivalent in soft tissue at an appropriate depth d below a
specific point on the body. The recommended depths are 0.07 mm for superficial radiation, 3 mm to monitor the eye lens dose, and 10 mm for penetrating radiation. These three quantities are used to assess in a conservative way the effects of ionizing radiation. Hp(0.07), Hp(3) and
Hp(10) are used to estimate in a conservative way relatively the skin equivalent dose (in order
to assess the deterministic effects on the skin), the eye lens equivalent dose (to assess the deterministic effects on eye lens) and the effective dose (to assess the stochastic effects on whole body).
The dose equivalent is defined at a point as the product of the absorbed dose D and the quality factor Q:
𝐻 = D Q (2.10)
The SI unit for dose equivalent is the sievert (Sv) which is one joule/kilogram (J/kg)[3,4]. The quality factor Q is defined to take into account the difference in effectiveness of different types of radiation in producing biological effects and it is specified as a function of the unrestricted linear energy transfer L of the charged particles in water as:
𝑄(𝐿) = {
1 𝑓𝑜𝑟 𝐿 ≤ 10
0.32 𝐿 − 2.2 𝑓𝑜𝑟 10 < 𝐿 < 100
300/√𝐿 𝑓𝑜𝑟 𝐿 ≥ 100
(2.11)
22
3. E
XPOSURE TO IONIZING RADIATION
People are exposed to natural radiation on a daily basis. The exposure is due to the natural background radiation, medical diagnostic and therapeutic procedures, nuclear weapons testing, generation of electrical energy by nuclear power plants, accidents in the nuclear fuel cycle and security inspection[6,7].
3 . 1 T
Y P E S O F E X P O S U R ERadiation exposure may be classified through the different exposure pathways.
3.1.1 Internal irradiation
Internal radiation exposure occurs when a radionuclide is inhaled, ingested or enters in the bloodstream (this could happen trough open wounds or injection). This type of exposure ends when the radioactive elements are eliminated from the body, both in a spontaneous way and as results of treatments[7].
The assessment of doses for intakes of radionuclides relies on the calculation of their retention inside the body. The intake can be estimated either from direct measurements (external monitoring of specific tissues or of the whole body) or from indirect measurements (for instance through a blood, urine or faeces analysis). The effective dose is then calculated using dose coefficients recommended by the ICRP for the specific radionuclides. The radioactive elements incorporated in the human body irradiate the tissue over time periods determined by their biological retention within the body and their half-life, so it is possible that these radionuclides give doses to the tissues for months or years. The need to regulate internal exposures and the accumulation of dose over extended period has led to the definition of committed equivalent dose.
The committed dose from an incorporated radionuclide is the total dose expected to be delivered within a specified time period. The committed equivalent dose 𝐻𝑇(𝜏) in the tissue or organ T is defined by:
𝐻𝑇 = ∫ 𝐻̇𝑇(𝑡)𝑑𝑡
𝑡0+𝜏
𝑡0
(3.1)
23 The quantity committed effective dose 𝐸(𝜏) is then given by:
𝐸(𝜏) = ∑ 𝑤𝑇𝐻𝑇(𝜏)
𝑇
(3.2)
For workers the committed dose is normally evaluated for a period of 50 years after the intake. In the case of the population, for adults the commitment recommended period is 50 years, on the other hand for infants and children the recommended value is 70 years[5].
3.1.2 External contamination
External contamination exposure occurs when airborne radioactive elements are deposited on skin or on the clothes. The radionuclides may be in the form of dust, aerosol or liquid. This type of exposure ends when the radioactive elements are removed from the body, often simply through a washing procedure[7].
3.1.3 External irradiation
External irradiation exposure occurs when all or part of the body is exposed to penetrating radiation from an external source. During exposure this radiation can be absorbed by the body or it can pass completely through. This type of exposure ends when the radiation source is shielded or when the person moves out from the radiation field[7]. The assessment of doses from external irradiation is usually performed by individual monitoring using personal dosimeters. The operational quantities for individual monitoring are Hp(10), Hp(3) and
Hp(0.07)[5].
3 . 2 S
O U R C E S O F I O N I Z I N G R A D I A T I O NFor as long as mankind lives on Earth, men and women have been exposed to ionizing radiation from natural sources. In addition, new artificial sources of exposure have been developed over the past century. In Table 3.1 are shown the different contributions for both natural and artificial source of exposure[6].
3.2.1 Natural sources
For most individuals, exposure to natural background radiation is the largest component of their total radiation exposure. The main natural sources of exposure are cosmic radiation and natural radionuclides found in the soil and in rocks[6].
24
Source or mode Annual average dose (worldwide) [mSv]
Typical range of individual dose [mSv]
Natural Sources of exposure 2.4 0.2 – 10
Inhalation of radon 1.26 0.2 – 10
External terrestrial 0.48 0.3 – 1
Ingestion 0.29 0.2 – 1
Cosmic radiation 0.39 0.3 – 1
Artificial sources of exposure 0.6 0.6 – several tens
Medical diagnosis (not therapy) 0.6 0 – several tens
Atmospheric nuclear tests 0.005 Some higher doses around test sites still occurs
Occupational exposure 0.005 0 – 20
Chernobyl accident 0.002 See paragraph 3.2.2.5
Nuclear fuel cycle 0.0002
Doses are up to 0.02 mSv for critical groups at 1 km from some
nuclear reactor sites
Total 3
Table 3.1 – Annual average doses and ranges of individual doses of ionizing radiation by sources[6]
3.2.1.1 Cosmic radiation
Cosmic radiation is a relevant component of natural background radiation. At sea level, it contributes about 15% of the total dose from natural source, but at higher altitude and especially in outer space, it is the dominant radiation source. This is because cosmic radiation is significantly attenuated by the Earth’s atmosphere. At cruising altitude of commercial aircrafts, the average dose rates are about 3-8 Sv per hour, these values are two order of magnitude higher than at sea level[6].
3.2.1.2 Natural radionuclides
Everything in and on the Earth contains radioactive elements: the primordial radionuclides (K-40, Th-232 and U-238) together with some radionuclides into which they decay emitting radiations. External exposure due to natural radionuclides varies considerably from place to place and it can range up to hundred times the average value. These radionuclides are also present in food and drink and so become incorporated in the body and they are present in the environment in a highly variable way. In Figure 3.1 is shown, as example of this variability, the concentration of natural uranium observed in drinking water in different states[6].
25
Figure 3.1 - Variability of natural uranium concentration in drinking water[6]
3.2.1.3 Radon
One important radionuclide produced from the U-238 decay series is Rn-222. This gas is a normal constituent of soil gas and seeps into buildings. Radon level varies very dramatically depending on the local geology and other factors like the permeability of the soil, the local climate and the building layout. When radon is inhaled, some of its short-lived decay products are retained in the lungs and irradiate cells in the respiratory apparatus. Radon contributes for about half to the average dose provided from natural sources. Because of this, very extensive measurement programs have been conducted in order to implement procedures to reduce indoor radon concentrations[6].
3.2.2 Artificial sources
Artificial sources of ionizing radiation may be subdivided in two groups: the ones due to military activities (such as military bomb tests and the process of nuclear weapon production) and the ones related to peaceful activities (such as the medical use of radiation, the production of energy and the nuclear fuel cycle)[6].
3.2.2.1 Nuclear military test explosions and weapons production
Nuclear test explosions in the atmosphere were conducted, between 1945 and 1980, mostly in the northern hemisphere. Totally, 502 tests were conducted and the most active test periods were 1952-1958 and 1961-1962. As shown in Figure 3.2, the highest value of the estimate per caput effective dose of ionizing radiation was 0.11 mSv for the year 1963 and after it fell to its
26 present level of about 0.005 mSv. This source will decline very slowly in the future because most of it is now due to C-14 that has a long half-life of about 5700 years. It is clear that some people living near the test sites received very large dose due to the local nuclear fallout. Following the signature of the “Treaty of banning nuclear weapon tests in the atmosphere, in outer space and under water” in the 1963, tests were conducted underground and any radioactive debris were usually contained: the radioactive residue generated is located deep underground and essentially fused with the host rocks.
.
Figure 3.2 - Estimated annual per caput effective dose of ionizing radiation worldwide from atomic bomb tests[6]
3.2.2.2 Medical diagnosis
Concerning the peaceful use of radiation, medical exposures represent the dominant part accounting for about 98% of the contribution from all artificial sources and continuing to grow at a remarkable rate. Medical exposures are usually voluntary and provide direct benefits to the exposed individual. In countries with high levels of health care, exposure from medical uses is on average now equal to about 80% of that from natural sources (see Table 3.1 and Figure 3.3).
27 In Figure 3.3 is shown, for the period of 1997-2007, the annual average per caput effective dose due to diagnostic medical examinations by health care level (I the highest, IV the lowest – based on the number of physicians per population)[6].
Figure 3.3 - Annual average per caput effective dose due to diagnostic medical examinations by health care level for the period 1997-2007[6]
3.2.2.3 Nuclear fuel cycle
Despite the increase of the decommissioning of the older nuclear power plants, the generation of electrical energy by nuclear energy has grown steadily in the last decades as we can see in Figure 3.4.
The nuclear fuel cycle includes the mining and the milling of uranium ore, the fabrication of the fuel, the production of energy in the nuclear power plants, the reprocessing and the storage of the irradiated fuel and the storage and disposal of the radioactive wastes. The public exposure varies widely with the different types of installation but they are generally small and they decrease markedly with the distance from the facility. The annual per caput dose to representative local population around nuclear power plants is less than 0.0001 mSv, practically the equivalent to the dose received from cosmic radiation in a few minutes on a commercial aircraft. The dominant part of exposure related to the nuclear fuel cycle is the one concerning the mining and milling processes[6].
28
Figure 3.4 - Installed nuclear electricity-generating capacity worldwide[6]
3.2.2.4 Occupational exposure
Until the 1990s, the attention in the area of occupational exposure was focused only on the exposures related to artificial sources of radiation. On the other hand, now, it is realized that a very large amount of workers are exposed occupationally to natural sources of radiation. The total number of workers exposed to ionizing radiation is estimated to be around 23 million, of whom about 13 million are exposed to natural sources of radiation. Medical workers comprise the largest part, about 75%, of the 10 million artificial sources exposed workers. Radiation exposure of workers involved in military activities occurs during production and testing of weapons, operation of reactors for propulsion of naval vehicles and other activities similar to the civilian ones. The workers of mining sector represent a relevant part of the exposed ones: in this case, the main source of radiation exposure is radon. In Table 3.2 is shown the average annual effective dose in different working sectors[6].
Working sector Average annual effective dose [mSv]
Military 0.1 – 0.2
Coal mines 2.4
Other mines (excluding uranium mines) 3
Airline flight crew 2 – 3
Short space mission 2 - 30
Nuclear fuel cycle 1
Medical use of radiation 0.5
Industrial use of radiation 0.3
29 3.2.2.5 Incidents
A small number of accidents have occurred in association with the nuclear fuel cycle and have attracted widespread publicity. However, more than one hundreds accidents have occurred with medical and industrial sources and have caused injury to workers and the public. Accidents can also occur during the use of radiation emitting medical machines both due to human or machine errors. Large radiation sources are used in industry (industrial irradiation facilities or accelerators) and have been involved in accidents, usually attributable to operator’s errors. The most serious accident that caused radiation exposures of the general population was the Chernobyl one in 1986. The accident caused a large uncontrolled radioactive release into the environment during about 10 days after the initiator event. The radioactive cloud created by the accident dispersed over the entire northern hemisphere and deposited substantial amounts of radioactive material over larger areas of the ex-Soviet Union and over Europe, contaminating land, water and food. The collective dose from this accident was many times greater than the combined precedent ones: in 1986, the average dose to more than 300000 recovery workers was nearly 150 mSv and more than 350000 other individuals received doses greater than 10 mSv[6].
Since the Chernobyl accident, the international community has made unprecedented efforts to assess the magnitude and characteristics of radiation-related health effects. This was done through collaboration between United Nations Educational, Scientific and Cultural Organization (UNESCO), the World Health Organization (WHO), the International Atomic Energy Agency (IAEA), the ICRP (International Commission on Radiological Protection) and the European Commission[6].
30
4. H
EALTH EFFECTS OF IONIZING RADIATIONS
When ionizing radiation passes through matter, included living tissues, it deposits energy. This deposition produces ionization and excitation in the matter. The biological damage caused by ionizing radiation is related to the amount of energy deposited. The amount of energy deposited per unit of mass is called absorbed dose and it is measured in grays, (1 Gy = 1 J/kg).
Different kinds of radiation have different biological effects for the same amount of energy deposited. Furthermore, different tissues react in different ways to ionizing radiation due to their different radiosensitivity. To take into account these two phenomena, a weighted quantity called effective dose is used as indicator of the potential biological effects. The unit of measurement of the effective dose is the Sievert, 1 Sv = 1 J/kg, and it is a way to measure ionizing radiation in terms of their harmful potential[5,7].
4 . 1 D
A M A G I N G P R O C E S SIncident radiations can cause excitation or ionization in matter. The excitation generates heat but it is usually very low and negligible unless it is generated by a very high amount of radiation (a dose of over 4000 Gy is required to raise the average temperature of a soft tissue by 1°C and this elevation is overshadowed by normal metabolic fluctuations).
On the other hand, ionization is more dangerous because it could damage DNA. The critical volume of cells exposed to radiation is the one occupied by chromosomes. Chromosomes occupy about 1% of the total volume and the remainder is mostly occupied by water. For this reason, is more likely that the radiation interacts with water than with DNA.
So, as we can see in Figure 4.1, the DNA damaging process can be classified in two different typologies:
Directly damage – when the interaction happen directly on DNA and the ionization breaks the DNA strands.
Indirectly damage – when the interaction happens not directly with DNA, but the ionization creates free radicals in the proximity of the DNA strands and damage it. Free radicals, i.e. molecules or ions with unpaired electrons, are quite reactive atoms,. The 2/3 of the DNA damage is caused by indirect action[8].
31
Figure 4.1 – Direct and indirect radiation action on DNA strands[8]
Ionizing radiation can damage DNA breaking its strands. We called single strand break the damage of only one DNA strand and double strand break the two opposite DNA strand damage. Figure 4.2 shows the different possible DNA configurations after the interaction with ionizing radiation.
A – Two-dimensional representation of the normal DNA helix: the base pairs carrying the genetic code are complementary (guanine pairs with cytosine and adenine pairs with thymine).
B – A break in one strand occurred. It is not a relevant problem because the broken strand is repaired using the opposite strand as complementary template.
C – Breaks in both strands: if well separated, they are repaired as independent breaks.
D – Breaks in both strands, directly opposite. The repairing process is not uniquely determined. The cells try to repair themselves without having a template, and so it may lead to errors in the genetic code that could be transmitted with the mitosis process.
32
Figure 4.2 - Single and double DNA strand breaks[8]
Summarizing, double DNA strand breaks are more dangerous because they can lead to cell death or the creation of genetic errors that can be transmitted during cell reproduction. Furthermore, the repair time for double strand break is greater than for single strand breaks, relatively several hours instead of about 10 minutes[8].
4 . 2 F
A C T O R S A F F E C T I N G C E L L U L A R D A M A G I N G P R O C E S SThe damage of DNA is affected by the characteristics of the exposed cells (for instance, the size, the type and the cell cycle effects) and by the exposure (such as the type and the characteristic of the incident radiation)[8].
4.2.1 LET
There is a big difference between high LET (Linear Energy Transfer) and low LET radiation: the first type (for instance alpha particles) releases a high amount of energy per unit of length, on the other hand, low LET radiation (such as X-rays and photons) has a low interaction rate per unit of length.
Figure 4.3 shows the different types of interaction between a low LET particle (on the left) and a high LET particle (on the right) with DNA strands. As we can see, having a low LET means that the distance between two consecutive interactions is greater than the distance between the two DNA strands. So, in the case of low LET radiation particles the probability of double DNA strands breaks is lower than in the case of high LET radiation. In the case of low LET radiation, a dose of 1 Gy produces an average of 1000 DNA strands breaks per cell with a percentage of double breaks of about 4%. In the case of high LET radiation this values is greater and can reach up to 20%[8].
33
Figure 4.3 - Low and high LET radiation interaction with DNA strands[8]
Figure 4.4 shows the surviving fraction in cell culture as function of the given dose and the LET of the incident radiation. In the case of low LET particle (in green), the trends is characterized by an initial plateau due to the fact that for low doses the damage is repairable and the exponential decrease of the surviving cells takes place after a certain value of the dose. On the other hand, in the case of high LET particles (in red), the exponential decrease starts immediately because the cells are not able to repair the damage also for low values of the dose[8].
In addition to the cell culture analysis, epidemiological studies on populations exposed to radiation (for example atomic bomb survivors) were taken in the past. The Hiroshima bomb was different from the Nagasaki one, relatively composed by enriched U-235 and Pu-239. For that reason, the Hiroshima bomb was characterized by a greater amount of neutrons, which have a higher LET. As we can see in Figure 4.5, the health effects were different: in the case of Hiroshima bomb survivors, the number of leukemias per 100000 person-years was greater than in the case of Nagasaki bomb survivors. The received dose was estimated with retrospective dosimetry technique[8].
34
Sparsely ionizing – X-rays or gamma rays
Densely ionizing – Alpha particles
Figure 4.4 – Fraction of surviving cells as function of given dose and LET of the incident radiation[8]
35
4.2.2 Dose rate
In Figure 4.6 is shown the fraction of cells surviving as function of the given dose and the dose rate of the incident radiation. As we can see, increasing the dose rate the exponential decrease of surviving cells is faster. This is because, at higher dose rate, cells haven’t enough time to repair the damage[8].
Figure 4.6 - Fraction of surviving cells as a function of given dose and dose rate[8]
4.2.3 Dose fractionation
In Figure 4.7 is shown the fraction of surviving cells as function of the given dose and the dose fractionation of the incident radiation. The first curve, abc, is obtained by irradiate the cells in a single exposure. As we can see, giving the dose D, only about 1 on 1000 cells survived. The second curve, abd, is obtained fractionating the dose. Once arrived to point b, giving the dose Df, a recovery time of about 15 hours is allowed. Once started again the
irradiation and reached the dose D, the fraction of cells survived was 1 on 10 instead of 1 on 1000. This is because the cells have the time to repair the damage and so increase their probability to survive. As we can see in Figure 4.7, to reach the fraction of 1 on 1000 cells survived, we need to expose the cell culture to a higher value of dose than the previous value D. The dose fractionation technique is used in radiotherapy because, differently from normal
36 cells, the tumor ones, don’t have the capacity to recover the damage produced by ionizing radiations[8].
Figure 4.7 - Fraction of surviving cells as a function of given dose and fractionated treatment[8]
4 . 3 H
E A L T H E F F E C T SThe health effects due to exposures to ionization radiation can be subdivided in somatic effects, if the effects are seen in the irradiated individual, and inheritable effect, if the effects are seen in the offspring of the irradiated individual. In the case of inheritable effects, embryos and fetuses are more sensitive to radiation because their cells are in the phase of rapid subdivision. The somatic effects can be subdivided in deterministic and stochastic effects[8].
4.3.1 Deterministic effects
Beyond a certain threshold dose, radiation impair the functioning of tissues or organs and produce acute effects. These effects are more severe at higher dose and higher dose rates. For instance, the dose threshold for acute radiation syndrome is about 1 Sv[7]. These effects may be early (within hours or days) or chronic/late. In Table 4.1 is shown a summary of acute effects for a whole body exposure[8].
37
Acute radiation exposure, effective
dose, [Sv]
0-1 1-2 2-6 6-10 10-50 >50
Treatment required Reassurance
Reassurance and hematologic surveillance
Blood transfusion and antibiotics
Consider bone marrow transplant
Maintenance of
electrolyte balance Sedatives
Overall treatment plan None needed Observation Effective Therapy promising Palliative Palliative
Incidence of vomiting None 5% at 1 Sv
50% at 2 Sv 100% at Sv 100% 100% 100%
Delay time prior to
vomiting N/A 3 h 2 h 1 h 30 min 30 min
Leading organ affected None Blood forming tissue Blood forming tissue Blood forming tissue Gastro intestinal track Central nervous system
Characteristic signs None Mild leukopenia
Severe leukopenia, hemorrhage, hair loss
above 3 Sv Severe leukopenia, infections, erythema Diarrhea, fever, electrolyte imbalance Convulsion, tremor, ataxia, lethargy Critical period post
exposure N/A N/A 4-6 w 4-6 w 5-14 d 1-48 h
Prognosis Excellent Excellent Good Guarded Hopeless Hopeless
Incidence of death None None 0-80 % 80-100 % 90-100 % 90-100 %
Cause of death N/A N/A Hemorrhage and infection
Hemorrhage and
infection Circulatory collapse
Respiratory failure and brain edema
In Figure 4.8 is shown the trends of the percentage lethality within 30 days as a function of the given dose for a single whole body exposure. The value related to the probability of death of 50% within 30 days is called LD 50/30. As we can see from the graph, no deaths are expected below 1 Gy. In this particular graph, which represents human response to radiation exposure, LD 50/30 is reached at 3.5 Gy. The graph also demonstrates that at a dose of 6 Gy or above no one is expected to survive. In reality, survival is still possible with extensive medical intervention. The LD 50/30 is different for different animal species, as shown in Table 4.2. The sensitivity to radiation and so the lethal dose are determined mostly by the efficiency of the immune system[8].
39 Species LD 50/30 [Gy] Human 3.5 Monkey 4 - 4.75 Dog 3 Hamster 7 Rabbit 7.25 Rat 9 Turtle 15 Newt 30
Table 4.2 – Value of LD 50/30 for different animal species[8]
4.3.2 Stochastic effects
If the dose is low or delivered over a long period of time (low dose rate) there is greater probability that damaged cells repair themselves. However, long-term effects may still occur if the damaged cells are repaired but incorporates errors that could be transmitted during cells division. These transformations may lead to cancer after years or decades and there is no threshold below which it can be said with certainty that the effect will not occur. Then, effects of this type will not always occur, but their probability is proportional to the radiation dose. On the other hand, the severity of the effect in a single individual is unrelated to the magnitude of the dose. The risk is higher for children and adolescent, as they are more sensitive to radiation exposure than adults. Epidemiological studies on populations exposed to radiation (for example atomic bomb survivors or radiotherapy patients) showed a significant increase of cancer risk at dose above 0.1 Sv[7,8].
40
5. T
HE NEED FOR PERSONAL DOSIMETRY SYSTEMS
As we have seen in the previous sections, the interaction of ionizing radiations with the human body can lead to deterministic or stochastic effect in relation with the amount of received dose and the exposure time.
5.1.1 The goals of radioprotection
The goals of radiation protection are the prevention of the occurrence of deterministic effects (serious acute and chronic radiation-induced conditions) in exposed people and the reduction of the stochastic effects to a degree that is acceptable in relation to the benefits, to the individual and to the society, from the activities that generate such exposure[8].
5.1.2 Dose response model
Based on the hypothesis that genetic effects and some cancers may results from radiation induced damage, the risk of stochastic effects is assumed to be proportional to the dose without threshold. Furthermore, the probability of response is assumed to accumulate linearly with dose. Regarding deterministic effects, at higher dose, more complex (nonlinear) dose-risk relationship may apply for acute exposure.
In Figure 5.1 different dose response models for low doses are shown:
a, linear with no threshold – the one used for the determination of stochastic effect for radioprotection purposes
b, linear with threshold – in this case, such as for deterministic effect, it is suggested that below a certain value negative effects are negligible or even beneficent ones (hormesis phenomenon: low exposures help the survive stimulating the activation of repair mechanisms that protect against disease, that are not activated in absence of ionizing radiation)
c, nonlinear – the curve is characterized by a low response at lower doses
d, over-response – in this case the over-response is due to the radio-induced apoptosis: phenomenon in which cells exposed to radiation emit, while dying, a bio-chemical signal that indicates to other cells that there was damage and, in order to prevent further damages, all cells die[8].
41
Figure 5.1 - Different dose response models[8]
5.1.3 Dose limits
Ionizing radiation damage to the human body is cumulative, and it is related to the total dose received. Then it appears clear that it is very important to have a measure of the different doses received by the exposed people, and this is done using personal dosimetry system. Indeed, workers exposed to radiation are required to wear dosimeters so a record of occupational exposure can be made. In order to reduce the risk of biological effects related to the exposure to ionizing radiation, the ICRP recommended dose limits both for exposed workers and for the public[4]. Within a category of exposure, occupational or public, dose limits apply to the sum of exposures from sources related to practices that are already justified. The recommended dose limits are summarized in Table 5.1[4].
5.1.4 Performance requirements for a personal dosimetry system
In order to assess the quality of a determined personal dosimetry system it is necessary to perform on it type tests as shown in IEC International Standard 62387 “Radiation protection instrumentation – Passive integrating dosimetry systems for personal and environmental monitoring of photon and beta radiation”. A list of the most important tests with a brief description is given in Table 5.2.
42
Type of limit Annual dose limit [mSv]
Occupational Public
Effective dose 20a 1b
Equivalent dose
Lens of the eye 20c 15
Skind 500 50
Hands and feet 500 -
Table 5.1 – Recommended dose limits in planned exposure situations[4]
a - Averaged value over defined periods of 5 years with the further provision that the effective dose should not
exceed 50 mSv in any single year.
b - In special circumstances, a higher value of the effective dose could be allowed in a single year, but the average over 5 years does not exceed 1 mSv/year.
c - Averaged value over defined periods of 5 years with the further provision that the effective dose should not exceed 50 mSv in any single year.
43
Test subject Brief description
Coefficient of variation The statistical fluctuationsshall not exceed the given limit
Non linearity The variation of the response due to a change of the given dose shall not exceed the given limit Overload When the dosimeter is irradiated with high dose the
system shall display an overload message
After-effects
If a dosimeter irradiated to high dose values produces after-effects on any subsequent
measurement, suitable measures shall be taken to ensure that the requirements of the IEC 62387 standard are met in the subsequent measurements.
Reusability
If the dosimeters cannot be reused indefinitely or if usability depends on the history of the dosimeter, tests shall be done to ensure the quality of
the measurement after a determined number of readouts.
Energy
The variation of the relative response due to a change of the radiation energy
within the rated ranges shall not exceed the given limits
Angle of incidence
The variation of the relative response due to a change of the radiation angle of incidence
within the rated ranges shall not exceed the given limits
Over response to radiation incidence from the side
If the dosimeter is irradiated free in air from the side, the indicated value
shall not exceed 1,5 times (or 2 times) the indicated value resulting from an irradiation free in air with the same radiation quality from the front (0°)
for Hp(10) and Hp(3) dosimeters (for
Hp(0.07) dosimeters, the factor of 2 applies). This shall apply to all radiation energies within
the rated range of energy.
Response to mixed irradiation
The response to mixed irradiation must be a linear combination of the signal for the different radiation types detectors and fulfilled the given requirements.
Ambient temperature
The variation of the relative response due to a change of the ambient temperature
within the rated ranges shall not exceed the given limits
Relative humidity
The variation of the relative response due to a change of the relative humidity
within the rated ranges shall not exceed the given limits
Light exposure
The variation of the relative response due to a change of the light exposure
within the rated ranges shall not exceed the given limits
Dose build-up and fading
The relative response and the deviation due to dose build up and fading shall not exceed the
given limits
Electromagnetic
The absolute value of the deviation due to electromagnetic disturbance shall not exceed the given
limits
44
P
ART
2:
E
VOLUTION OF PERSONAL DOSIMETRY
SYSTEMS
45
6. L
UMINESCENT DETECTORS
TLDs are inorganic crystal class of detectors based on the principle of luminescence. Luminescence is the process of light emission from a material not involving heating[9].
6 . 1 L
U M I N E S C E N C EAccording to Stoke’s law, when radiation is incident on a material, some of its energy may be absorbed and re-emitted as light of a longer wavelength: this is the process of luminescence. The emitted light is characteristic of the luminescence substance and not of the incident radiation. Luminescence can be classified in order to take into account the type of phenomena used to excite the emission of light.
In the case of stimulation by an incident radiation, we can have:
Photoluminescence – when the excitation is made with optical or ultra-violet light
Radioluminescence – in the case of nuclear radiations like X-rays, gamma rays, beta particles, etc.
Cathodoluminescence – in the case of electrons beam
In addition to excitation by radiation, there are also other ways to stimulate the emission of light:
Chemiluminescence – by chemical energy
Triboluminescence – by mechanical energy
Electroluminescence – by electrical energy
Bioluminescence – by biochemical energy
Sonoluminescence – by sound waves
The emission of light takes place after a characteristic time 𝜏𝑐 after the absorption of radiation; this parameter allows us to sub classify the process of luminescence as follows:
Fluorescence – 𝜏𝑐 < 10−8 𝑠
Phosphorescence – 𝜏𝑐 > 10−8 𝑠
Phosphorescence itself may me conveniently subdivided in two main types:
Short-period phosphorescence – 𝜏𝑐 < 10−4 𝑠
46 In Figure 6.1 is shown the classification of luminescence on the basis of the delay time 𝜏𝑐
between the irradiation and the emission of light. The prefix to the term luminescence distinguishes between the modes of excitation.
Figure 6.1 - Classification of luminescence[10]
From a practical point of view, the only clear way to distinguish between phosphorescence and fluorescence is to study the effect of temperature. In the case of phosphorescence there is a strong temperature dependence, instead fluorescence is essentially independent of temperature. Figure 6.2 shows the relationship between radiation absorption and the emission of fluorescence, phosphorescence and thermoluminescence. T0 is the temperature at which
irradiation takes place, tmax is the time at which phosphorescence reaches the maximum light
emission intensity, tr is the time at which irradiation ends and the decay of phosphorescence
