Radiochemistry Practical Training
for chemists
Freie Universität Berlin
Institute for chemistry and biochemistry
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Contents
Rules for working in radioactive laboratories
Comparison of various detectors and nuclides
Decontamination of radioactively contaminated surfaces
Calculation of the γ-radiation dose rate
γ-spectrometry; recording of γ-spectra
β-spectroscopy (liquid-scintillation counter)
Backscattering of β--radiation
Self-absorption of β--radiation
The Thorium Generator – Determination of the half-lives of different
members of the natural thorium decay series
Responsible for content
Prof. Dr. U. Abram (abram@chemie.fu-berlin.de)
Dr. Adelheid Hagenbach (hagenb@zedat.fu-berlin.de)
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Rules for working in radiochemical laboratories
Only the bare necessities (writing material, calculator) are allowed into a radioactive laboratory. Jackets as well as book-bags are to be stored outside of the radioactive laboratory in available lockers.
Individuals, handling non-confined radioactive materials are obliged to wear suitable protective clothing (laboratory jacket, overshoes, disposable gloves) and equipment (dosimeter).
When handling radioactive substances, eating, drinking, smoking, and applying make-up are
prohibited in the laboratory. (§70-Handling of non-confined radioactive substances –
StrlSchV). Special attention should be paid to open wounds.
When working with non-confined radioactive substances, procedures which avoid the incorporation of radioactive substances and the contamination of participants are used. The occurrence of contamination immediately must be reported to an assistant.
Operation of equipment with the mouth is never allowed (application of piston-driven pipettes
or peleus balls).
Special attention is required when handling dust-developing, radioactive solids (danger due to
inhalation).
According to §4 of StrlSchG, anyone who carries out a task must adhere to the following obligations.
Any needless radioactive exposure or contamination of people, property or the environment must be avoided.
Taking into account the state of science and technology and bearing in mind all circumstances of a particular case, any radioactive exposure or contamination of people, property
or the environment must be kept as low as possible and below the limit defined in §8principles in the protection against radiation – StrlSchG.
When handling non-confined radioactive substances, disposable gloves must be worn at all
times. Disposable gloves are always regarded as being contaminated.
All work with non-confined radioactive substances is performed in designated plastic trays
(according to the “principle of double containment”).
When transporting radioactive substances, sealable transport containers must always be used
(“principle of double containment”). Non-confined substances are never to be touched by the
hands (contamination danger). Tweezers must always be used. Storage containers are to be
properly labelled and may only be used for the storage of radioactive substances (§91labelling obligation-StrlSchV).
Solid radioactive wastes are stored in available, specially designated containers. Combustible
and non-combustible wastes are collected separately.
Liquid radioactive wastes are poured into available, specially designated containers – never
down the drain (§99-Protection of air, water, and soil –StrlSchV).
Prior to and after completion of the - Use the gloves-monitor
experiment:
- Wash your hands
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Chapter 1
Comparison of Various Detectors and Nuclides
Task
Investigate which detectors are best suited for the detection of α-, β-, and γ-radiation! The
nuclides 241Am, 14C, 204Tl and 60Co are used.
Fundamentals
1. Production of nuclear radiation
Nuclear radiation is produced by the decay of unstable atomic nuclei. These nuclei are stabilized by emitting particles (α- and β-radiation) or electromagnetic waves (γ-radiation). The
emission of alpha and beta radiation is always accompanied by a nuclear transformation,
meaning that a different chemical element is formed. The emission of γ-radiation itself does
not lead to the formation of a different element; the same chemical nucleus simply goes from
a higher energy to a lower energy state (which may be the ground state). However, γ-radiation
often accompanies α- or β-decay of atomic nuclei. Other types of nuclear radiation are: electron capture reactions (“capture” of an electron near the nucleus by a proton), proton radiation, neutron radiation, or spontaneous fission of the nucleus.
1.1. α-Radiation
During α-decay, helium nuclei are emitted. The mass number of the nucleus is reduced by
four units and its atomic number by two.
X
Y
A →
X −4
Y −2
B + 24He
Typical elements experiencing α-decay are the
heavier elements toward the bottom of the periodic table of elements. α-Decay becomes possible
when the emission of the (extremely stable) αparticle is energetically favorable for the entire
system:
A → B + 24 He + ΔE
ΔE = (MA - MB - MHe) · c2
Fig. 1: Formation of α-radiation
(241Am → 237Np + He2+)
The emitted α-particles possess a defined energy, which depends on the particular nuclear
decay. α-Spectra are line spectra, which can be used to characterize nuclides. Besides αspectra with just one line, there are also examples of α-spectra with multiple energetic transitions. This can be explained by the favorable formation of decay products in different excited
states.
1
The half lives of α-emitters vary widely. Besides very short-lived nuclei (for example 212Po,
t1/2 = 3 x 10-7 s), there also exists very long-lived α-emitters (for example 232Th, t1/2 = 1.4 x
1010 a).
1.2. ß-Radiation
β-Radiation can be described as the conversion
of a neutron to a proton (β--radiation) or as the
conversion of a proton to a neutron (β+radiation). The result of these conversions is an
energetically more favorable situation in the
nucleus, achieved through a “balanced” ratio of
protons and neutrons. The number of neutrons
that are needed to stabilize an atomic nucleus
Fig. 2:
Formation of β--radiation
depends on the number of protons. The stable
(14C → 14N + e-)
nuclei of lighter elements have a proton/neutron ratio of about 1:1, while the stabilization of the heavier elements requires more
neutrons than protons. A line drawn along the stable nuclei of the nuclide chart is called the
“β-stability line.” Nuclides with an excess of neutrons (or a deficit of protons) approach this
line through β--decay. Nuclei with a neutron deficit (or proton excess) sustain β+-decay. As a
result of these nuclear transformations, negatively charged β-particles (electrons) or β+-particles
(positrons) are emitted. At the same
time, either an antineutrino (in the
case of β--decay) or a neutrino (β+decay) is also emitted (conservation of momentum). The energy
that is released by β-decay is distributed among the decay products
in such a manner that only a statistical energy distribution with typical parameters Emax (maximum
energy when all of the energy is
imparted to the β-particle) and
Fig. 3: Statistical energy distribution of β--particles
Emean (average value of β-energy,
approximately one-third of Emax)
can be assigned (Fig. 3).
During
β--decay,
the
atomic
mass
number
of
unchanged. However, the atomic number is increased by one unit.
X
Y
A →
X
Y +1
the
atom
remains
B + −10ß − + ν
During β+-decay, the atomic mass number of the atom also remains unchanged. However, the
atomic number is reduced by one unit.
X
Y
A →
X
Y −1
B + +10ß + + ν
Positrons resulting from β+-decay react immediately with surrounding electrons to produce
two γ-quants (annihilation of a positron).
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1.3. γ-Radiation
γ-Radiation is created by the transition of an atomic nucleus from an excited state to the
ground state (or to another excited state of lower energy). The energy difference between the
states is released in the form of an energy-rich electromagnetic radiation (γ-radiation). γRadiation is coupled to α- or β-decay, when these nuclear transformations do not result directly in the ground state of the product nucleus (Fig. 4).
γ-Radiation without accompanying particle
radiation is obtained from so-called meta-stable
nuclear isomers. This refers to nuclides, in
which the emission of γ-radiation occurs long
enough after their creation that they can be separated from the “mother isotope.” Such metastable nuclear isomers are distinguished with
the symbol “m”; for instance 99mTc (t1/2 = 6 h)
or 137mBa (t1/2 = 3 min).
γ-Radiation of one individual disintegration is
monoenergetic and γ-spectra, which are line
spectra, are characteristic of the individual de- Fig. 4: Formation of γ-radiation
cay processes. The emitted energy corresponds
(60Co → 60Ni ( + e- + ν ) + γ-quant)
almost exactly (recoil effects) to the energy
difference between the excited and ground state.
2. Interaction of radiation with matter
The high energy radiation emitted during a nuclear transformation cannot be directly observed. Its existence can be seen only through its interaction with matter. In this process, the
kinetic energy of the radiation is distributed to its surroundings. This energy transfer occurs
predominately through scattering processes at the electron shells and the atomic nuclei of the
irradiated material and is manifested in
production of ions.
α- radiation
The detection and quantitative measurement of nuclear radiation are primarily realized through its ionizing
β-- radiation
character. The ionizing character of
nuclear radiation depends on the energy and specific type of radiation (Fig.
γ- radiation
5). This is mainly due to the different
ionization mechanisms of α-, β- and γFig. 5: Ionizing effect of nuclear radiation in air
radiation.
2.1. Absorption of α-radiation
α-Particles emitted during a distinct nuclear decay are monoenergetic. They usually possess
energies of 2-10 MeV. Primarily, inelastic collisions with the outer shell electrons are observed. In these processes, 0.1-1 keV of energy is transferred to the electrons. The average
energy for the production of an ion in air is 32.5 eV. Collisions with atomic nuclei are relatively rare. Since the mass of α-particles are much larger than that of the particles with which
they normally collide, the path of the α-particle undergoes only a slight change in direction.
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This results in a short and nearly straight
flight path, which depends on the energy of
the α-particle and the nature of the absorber.
α-Radiation is a strongly ionizing form of
radiation, which in air forms (assuming an
average energy) approx. 10,000 ion pairs per
cm traversed (Fig. 6).
Fig. 6: Ionizing effect of α-radiation in air
2.2. Absorption of β-radiation
The energy released by β-radiation can be distributed to the electron and the antineutrino in
different ways. A continuous energy spectrum is obtained for β--radiation, with a characteristic maximum β--energy for each β--emitter. The energy of β--particles is in the range between
a few keV and approximately 10 MeV. When β--particles collide with shell electrons, much
larger changes of direction are expected than when α-particles collide with shell electrons
since the masses of the two collisions partners (β--particles and shell electrons) are roughly
equal. Because of these large changes of direction and since β--particles are not monoenergetic, no defined range is observed (see Chapter 7: Backscattering of β--radiation).
The ionization density of β−-radiation is
substantially smaller than that of αradiation. The formation of approx. 100 ion Fig. 7:
Ionizing effect of β-radiation in air
pairs per cm air can be assumed for βradiation of average energy (Fig. 7).
2.3. Absorption of γ-radiation
γ-Radiation is electromagnetic radiation (photons), which possesses neither resting mass nor
charge. Because of this, absorption is only possible through scattering processes at shell electrons and/or atomic nuclei. The most important
interaction mechanisms are photoreactions,
Compton reactions and pair formation processes. A more detailed description can be found in
Chapter 4 “γ-Spectrometry, Recording of γ- Fig. 8: Ionizing effect of γ-radiation in air
Spectra“. With the formation of around 10 ion pairs per cm traveled in air, γ-radiation has a
relatively small ionization potential (Fig. 8).
The relatively small interaction of γ-radiation with matter presents a problem in regard to radiation protection since the radiation has a much larger range than α- or β-radiation and can,
therefore, have a direct effect on the body. Since the interactions are proportional to the atomic number of the matter, lead absorbers and lead-glass windows are used while working with
γ-radiation.
2.4. Range of nuclear radiation
Due to their different interaction mechanisms, α-, β-, and γ-radiation have very different
ranges.
Since emitted α- and β-particles engage in intense interactions with matter, their energy is
completely transferred to their surroundings within a short distance. On one hand, this leads to
a large number of ionization events, but, on the other hand, also leads to a small range for
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these types of radiation. Because of this, they can be shielded quite effectively. However,
there exists a large potential danger from the incorporation of α-emitters in particular since a
cascade of ionization processes may take place directly in biological tissue,
leading to the introduction of a large Radiation Air Aluminium Muscle (water)
amount of energy. This causes direct damage to cell material, whereas β- and espe0.32
0.0002
0.0004
α
cially γ-radiation are expected to have
ß394
0.206
0.438
comparatively smaller interactions.
Since γ-radiation concerns electromagnetic Table 1: Range (in cm) of α- and β−-radiation
(1 MeV)
radiation, specification according to
“range” is senseless. Because of this, values of “half-value thickness” for numerous materials
are given. These values correspond to the distance in which the intensity of the penetrating
radiation is reduced by one-half.
Examples of ranges of α- and β-radiation
Water
Concrete
Iron
Lead
and half-value thicknesses of γ-radiation
are arranged in the Tables 1 and 2. These
9.85
4.66
1.47
0.893
values and knowledge of the different
ionizing mechanisms play a crucial role in Table 2: Half-value thicknesses (in cm) of
γ-radiation (1 MeV)
the selection of suitable shielding materials.
3. Detectors for the measurement of ionizing radiation
For the measurement of radiation, the following reactions during the irradiation of matter are
predominantly utilized:
Ionization in gases (Ionization chamber, proportional flow counter, release counter)
Scintillation in solids and liquids (α, β- and γ-scintillation counters)
Ionization in solids (semiconductor-detectors)
3.1. Ionization of gases (gas-filled detectors)
Gas-filled detectors all possess the same fundamental construction. They consist of a gasfilled chamber with a built-in,
charged capacitor. When ionizing radiation enters this chamber,
ions are produced, which lead to
the (partial) discharge of the capacitor. Measurement of the so
produced electric flow makes it
possible to record the ionization
processes (Fig. 9). The level of
the detected signals depends on
the ionization properties of the
respective types of radiation. The
actual design of the radiation Fig. 9: General measurement setup for a gas ionization detecmeasuring instruments can vary
tor
widely from this general description and, therefore, influences the measurement geometry and
applications.
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The strength of the registered signals can be greatly influenced by the applied voltage since
only at relatively low applied voltage it is possible to exclusively measure the ionization processes generated by radiation. At higher voltages, secondary ionizations caused by the strongly accelerated charged particles
play a larger role. On the one
hand, this leads to a considerable increase in the strength of
the signal, but, on the other
hand, reduces the ability to distinguish between the individual
types of radiation (consider the
different ionization properties
of α-, β-, and γ-radiation).
For this reason, three different
measurement systems based on
gas-filled detectors can be used.
These measurement systems are
mainly distinguished by the
applied voltage and the resulting number of ionization processes (Fig. 10).
In an electric field, the charged Fig. 10: Relationship between applied voltage and signal
particles, which are formed by
strength in a gas-filled detector
the influence of radiation, are
accelerated according to the applied field strength. The primary and secondary processes,
which are observed as a result of this phenomenon, vary according to the field strength,
whereby distinct areas are recognizable.
U1:
U1-U2:
U2-U3:
>U3:
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Minimum voltage is required to create an impulse.
All produced ion pairs contribute to the impulse without further secondary effects. This is called the “saturation area” because within this area an increase of
the field strength does not lead to an increase in the impulse level.
Above the field strength corresponding to U2, all created ion pairs contribute to
the voltage impulse. A further increase of the field strength above this point causes secondary effects, which enlarge the signal. That means that the ions which are
produced in the primary process are accelerated at such a rate that they produce
other ions along their path. In this case, the amplification of the signal depends on
the type and energy of the radiation. This part of the field strength is called the
“proportional area” due to the proportionality between the primary and secondary effects.
The acceleration of the electrons is so great that not only the electronic outer
shells, but also inner shells of the counter gas are ionized. Upon filling the inner
shells, X-rays, which also produce ion pairs, are emitted. In this way, the ionization can encompass the entire counting volume, resulting in a signal that is independent of the strength of the primary ionization. This part of the field strength is
referred to as the “Geiger-Müller-Area” or the “release area.” The addition of a
suitable quench gas can limit the time-span of these discharges (dead time).
In this area of the voltage, it is impossible to distinguish between the types of
radiation. An impulse causes the total ionization of the entire counting volume.
Through the selection of gases, gas pressure, applied voltage and casing materials of the detector, the response to specific types of radiation can be optimized.
Because the specific ionization and the consequent signal produced by an α-particle are considerably greater than those of a β--particle, the voltage for the detection of β--particles must
be raised accordingly.
3.1.1. Geiger-Müller Counter (working in the release area)
Various designs of GeigerMüller counters exist. For the
measurement of solid specimens,
an end window counter is used
(Fig.11). Its primary components
are:
The counting chamber: It is
usually made of metal (e.g. Cu,
Fe, Al, Pb, Mg) or glass with a
metal coating. The wall of the
counter functions as the cathode.
The counting wire: It is the central electrode and functions as
the anode. Diameter approx. Fig. 11: Experimental setup of a Geiger-Müller counter
0.05 mm, material: e. g. W, Mo,
Fe
The end window: This is generally made of mica (a layered silicate), nylon or similar materials with a surface weight between 1.5 and 3 mg/cm2.
The fill gas: This is generally Ar or He. By adding methane, the reinforcing factor of the gas
by secondary electron formation near the counting wire becomes less dependent on the voltage of the counter.
In extreme cases, a type of constant discharge develops, which can be avoided by
quenching. With self-quenching counters,
alcohols or halogens are used as quench
gases. The positively charged gas ions
(G+), which are formed by the ionizing
radiation, transfer their charge to the
quench gas molecules (L):
G+ + L → G + L+
The positively charged quench molecules Fig. 12: Ionization effects in a Geiger-Müller counter
(L+) dissociate on their way to the negatively charged wall of the counter. At this
point, the fragments are no longer able to produce secondary electrons and the discharge process is interrupted. By using a sufficiently thin end window, α-radiation can still be measured, but with a small, poorly reproducible counting yield. The same is true for weak β−radiation. Geiger-Müller counters are most suitable for measuring high-energy β−-radiation
(Emax > 1 MeV). γ-Rays are registered with a low counting yield since only approx. 1 % of the
γ-quanta which pass through the detector trigger an ionization. The registered signals are due
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to photo and Compton processes of the γ-quanta at the atoms of the counting chamber. (See
Chapter 5 “γ-Absorption”)
3.1.2. Proportional Flow Counter
Fig. 13: Typical setup of a proportional counter (methane flow counter)
The counter operates in the proportional area with methane or argon/methane employed as the
counting gas.
The advantage of this detector is
that the measurement can be
carried out directly without an
end window. The sample is
brought into the measurement
chamber with the help of a rotating disk, such that the radiating
particles are brought into direct
contact with the counting gas.
The voltage used with the proportional counter depends on the
type and pressure of the counting
gas. This voltage is between Fig. 14: Ionization processes in a methane flow counter
2000 and 4000 V.
Proportional flow counters are especially suited for measuring α- and β−-emitters.
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3.2. Solid Scintillation Detectors
Primary components are the single crystal, the photo cathode and a secondary electron multiplier (PSEV).
Fig. 15: Experimental setup of a solid scintillation counter
Solid scintillation detectors are used mostly for the measurement of γ-radiation. In order to
absorb the largest possible amount of the γ-radiation (to make interactions between the radiation and the crystal possible), thick crystals with large densities are used; preferably, sodium
iodide crystals doped with thallium (ρ=3.67 g/cm3).
During the interaction of the γ-radiation with the thallium ions, electrons are produced, which
release scintillations. This results in the emission of bursts of light in the range of 400800 nm. These bursts of light hit the photocathode, from which electrons are consequently
emitted. The number of electrons is multiplied over the dynodes in the secondary electron
multiplier (from one electron, approximately 109 electrons are produced giving rise to an electric signal).
The number of light bursts is dependent on the energy of the γ-rays, and each γ-quant is
measured individually. The more light bursts are produced, the higher is the resulting signal.
This is because the scintillations are produced so rapidly consecutively that they lead to one
impulse. The dead time of such a measurement instrument is very short.
By using organic crystals like anthracene, trans-stilbene or p-terphenyl, β−-radiation can be
detected. For the detection of β−-radiation, organic scintillators have an advantage due to the
low atomic number and, therefore, scarce occurrence of problematic bremsstrahlung (see
Chapter 7 “Backscattering of β−-radiation”). The appropriate thickness of the scintillators depends on the measurement task. Often, it is necessary to select a thickness which is similar to
the range of the hardest β−-radiation to be detected. Thin organic scintillators have only a
small response to γ-quanta and are, thus, suitable for measuring β−-radiation in the presence of
otherwise interfering γ-radiation. For the detection of α-radiation, only scintillators without
protection layer or with very thin one can be used. For this purpose, ZnS(Ag)-scintillators are
often used. For the detection of mixed radiation fields, scintillation detectors coated with
ZnS(Ag) and a organic counting material are nowadays commonly used. With such counting
tubes, both α- and β−-radiations can be detected with a high sensitivity.
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4. The relationship between the activity and the counting rate
A detector of a given geometry only registers a portion of the radiation emitted from a sample.
The net counting rate (RN) is proportional to the activity (A) of the sample and the average
number of the emitted quanta per decay (ν):
RN = f⋅A⋅ν
(1)
with f = fG ⋅ fA ⋅ fS ⋅ fR ⋅ fi (2)
fG:
Geometric factor
Absorption factor /
extinction coefficient
f A:
fS:
fR:
fi:
Self absorption factor
Backscattering factor
Response probability
The geometric factor fG takes into account
the distance of the sample from the detector
and the diameter of the counting chamber
window. Assuming a point source of radiation, only a certain angle will be detected. In
other words, a cone-shaped volume is measured and converted into a spherical volume,


a


f G = 0.5 1 −

2
2
(a + r ) 

(3)
where a is the distance of the sample from
the detector and 2r is the diameter of the
counting chamber window.
The absorption factor fA, or rather the extinction coefficient, takes into account the
attenuation of the radiation in the layer of air
between the sample and the detector. The
attenuation of the radiation by the counting
chamber window is not accounted for by this
factor.
f A = e − µ ⋅d = e − µ ⋅d
*
µ:
d:
Where
*
(4)
µ*:
Mass extinction coefficient
Mass per unit area of the absorber d*:
d = ρ ⋅ d*
and
Linear extinction coefficient
Thickness of the absorber
µ* = µ ⋅ ρ
(ρ = density of the absorber)
The mass extinction coefficient µ is a function of the energy of the radiation and the atomic
number of the absorber material and depends directly on the interaction of the different types
of radiation with matter.
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Because of the small distance between the detector and the sample, the absorption factor can
be disregarded with γ- and strong β−-radiation (fA = 1). However, with weak β--radiation
(14C!) and α-radiation, the absorption factor must be considered. In the case of α-radiation, the
energy changes with the distance, while the particle number remains almost constant up to the
maximum range.
The self-absorption factor fS describes the absorption of the radiation within the sample. However, by
assuming an infinitely thin layer of the sample, this
factor can, in approximation, be set to 1.
The factor fS is strongly dependent on the preparation
of the sample and must be determined empirically
(see Chapter 8 “Self-Absorption of β−-radiation”).
Since the used samples are very thin, the selfabsorption factor can be disregarded.
The backscattering factor fR increases with increasing mass per unit area and increasing atomic number
of the material supporting the sample. By using the
same material of adequate thickness under the samples, this factor can be assumed to be constant. From
equations (1) and (2) we obtain:
RN = fG ⋅ fA⋅ fS⋅ fR ⋅ fi ⋅ A⋅ ν
(5)
If the geometric factor, the absorption factor, the self absorption factor and the backscattering
factor are all held constant during the measurement, the net counting rate, given the activity of
the emitter, is directly proportional to the response probability fi of the detector. However, this
is only the case with one sample measured with different detectors of the same type. For this
reason, when comparing different detectors and different nuclides, fG, fA, fS and fR must all be
considered in the calculation.
If it can be assumed that the backscattering factor fR is identical for all experimental set ups,
the product fR.fi, known as the relative response probability frel, can be used to compare detectors. From equation (5), we get for frel:
f rel = f R ⋅ f i ⋅ 100 =
11
R N ⋅ 100
%
f G ⋅ f A ⋅ f S ⋅ A ⋅ν
(6)
Experimental
1. Safety regulations
There is no danger due to external irradiation resulting from the samples, since they are all of
low activity. Never touch the samples with your finger! Always use tweezers!
2. Nuclides to be used
14C
Pure β−-emitter with a maximum energy of 0.156 MeV.
One particle is emitted per decay (νβ- = 1).
Half-life: 5730 years (t1/2 = 5730 a)
On 17 Feb., 1986, the sample had an activity (A) of 51,800 Bq.
204Tl
β−-emitter (97.9 %) with a maximum energy of 0.765 MeV, which
also emits conversion electrons (EC, 2.1 %).
One particle is emitted per decay (νβ- = 1).
Half-life: 3.78 years (t1/2 = 3.78 a)
On 24 Nov., 2021, the sample had an activity (A) of 3 kBq.
60Co
γ-emitter which emits two γ-quanta per decay (νγ= 2) with energies
of 1.173 MeV (100 %) and 1.332 MeV (100 %), respectively.
The γ-radiation is coupled to a preceding β−-decay (0.313 MeV at
100 %; νβ− = 1; (see Chapter 1 Fundamentals, 1.3 γ-Radiation),
which is not detected in the experiment due to the use of an Alabsorber.
Half-life: 5.26 years (t1/2 = 5.26 a)
On 24 Nov., 2021, the sample had an activity (A) of 3 kBq.
241Am
α-emitter (100 %) with an energy of 5.4 MeV.
One particle is emitted per α−decay (να= 1)
Half-life: 432.2 years (t1/2 = 432.2 a)
On 17 Feb., 1986, the sample had an activity (A) of 3,700 Bq.
Here, only types of radiation which occur at more than 1 % are specified.
1 Bq (Bequerel) = 1 decay/s
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3. Procedure
1.
Familiarize yourself with the operation of the different detectors (interactive experiment)!
2.
Determine the reference rates of the different detectors (Geiger-Müller counters with
and without cap, proportional flow counter, α-scintillation detector, β-scintillation
detector, γ-scintillation detector)!
Measurement time: 5 min each!
3.
Determine the gross counting rates for the samples (14C, 204Tl, 60Co and 241Am) with
each of the different detectors! In the case of Geiger-Müller counter without cap, αscintillation detector, β-scintillation detector, all measurements should be done for two
distances between the detector and the sample: 2 cm and 3 cm. In the case of γscintillation detector and Geiger-Müller counter with cap, it is enough to do the measurement at 2 cm distance. While using the proportional flow counter it is not possible
to change the distance, although you should measure the gross counting for the whole
samples. Please do not forget to write down the unit for each gross counting rate.
Measurement time: 1 min each
Evaluation
1.
Calculate the net counting rates for the individual measurements as well as the standard deviations (2σR = ± 2 ⋅ N0.5 ⋅ t-1) of the net counting rates
2.
Calculate the present-day activity of each of the samples and enter your results in Table 4!
A = A0⋅e-λ⋅t ; λ = ln 2 ⋅ t1/2-1
3.
Calculate the relative proportional response probability frel for all of the detectors and
all of the nuclides and enter your results in Table 5!
The geometric factors and the absorption factor (for 14C only) are shown in Table 3.
The average number of emitted quanta per decay can be found in the “summary of
nuclides to be used.” Please note that 60Co can only be treated as a pure γ-emitter
when an aluminum absorber is used. (=> ν = 2)!
4.
Discuss the response potential of the detectors for the different types of radiation!
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Detector
a
[cm]
r
[cm]
fG
GM counter
2
2.25
0.17
0.74
GM counter
3
2.25
0.10
0.41
α-scintillation detector
2
2.5
0.19
0.74
α-scintillation detector
3
2.5
0.12
0.41
β--scintillation detector
2
2.5
0.19
0.74
β--scintillation detector
3
2.5
0.12
0.41
2
1.75
0.12
0.74
2
2.0
0.14
0.74
-
-
0.5
1
γ-scintillation detector
1·1 Inch
γ-scintillation detector
2·2 Inch
proportional flow counter
Table 3: geometric factor fG; absorption factor fA for 14C only
Isotpe
Co
241
Am
204
Tl
14
C
60
Activity [Bq]
Table 4: Calculated activity today
14
(for
fA
14C only)
Isotope
GM counter
2 cm
distance
GM counter
3 cm
distance
GM counter
with cap
2 cm
distance
α-scint.
2 cm
distance
α-scint.
3 cm
distance
β-scint.
2 cm
distance
β-scint.
3 cm
distance
γ-scint.
1·1 Inch
2 cm
distance
γ-scint.
2·2 Inch
2 cm
distance
proport.
α
Co
60
Am
241
Tl
204
C
14
Isotop
proport.
α-β
Co
60
Am
241
Tl
204
C
14
Table 5:
15
Relative proportional response probability of the various detectors for different types of
radiation
Data Sheet: Detector Comparison
R0:
N:
t
reference rate of the measurement instrument
number of measured signals
time
Detector
R0
[Unit]
Nuclide
60
GM counter
2 cm
distance
GM counter
3 cm
distance
GM counter
with cap
2 cm
distance
α-scint.
2 cm
distance
α-scint.
3 cm
distance
β-scint.
2 cm
distance
β-scint.
3 cm
distance
16
Co
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
241
N
[Unit]
RM = N ∙ t-1
RN = RM – R0
σR = N0.5⋅t-1
t
[Unit]
RM
[Unit]
gross counting rate
net counting rate
measurement rate error
RN
[Unit]
2 σR
[Unit]
Detector
γ-scint.
1·1 Inch
2 cm
distance
γ-scint.
2·2 Inch
2 cm
distance
proportional
flow counter
α-Plateau
proportional
flow counter
α+β-Plateau
17
R0
[Unit]
Nuclide
60
Co
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
60
Co
241
Am
204
Tl
14
C
241
N
[Unit]
t
[Unit]
RM
[Unit]
RN
[Unit]
2 σR
[Unit]
Chapter 2
Decontamination of radioactively contaminated
surfaces
Task
1. Investigate the possibilities of decontaminating different surfaces. Employ different
decontaminants! Determine the percentage of activity remaining on the plates.
2. In a plastic tub, there are contaminations caused by 14C, 60Co and 90Sr. Localize the
contaminants with the use of suitable detectors and identify the emitters! Determine the
activity present on the surface!
Fundamentals
When handling unconfined radioactive materials, the danger of causing contamination by
spraying radioactive solutions or by the production of radioactive gases, steam or particles of
dust always exists. To prevent endangering the health of the individuals working with
radioactive materials, regular checks of instruments, workbenches, floors, etc. for possible
contamination are mandatory. When confronted with a contamination, the differentiation
between two cases is required:
● Adherent contamination: cannot be wiped clean by normal means
● Non-adherent contamination: Possible to be wiped clean
Non-adherent contaminations may cause danger due to outer irradiation as well as
incorporation.
Two basic methods are applicable to check for the presence of contamination:
-
Direct measurement in the inspection of surfaces
Indirect measurement in the inspection of surfaces
1. Direct measurement in the inspection of surfaces
To define the contaminated area, the surface is scanned with a suitable radiation detector of
sufficient sensitivity. The type of radiation can be determined with absorption measurements.
Especially suitable for the quantitative assessment of surface activity are the so-called largearea proportional counters with a maximal counting area of 300 cm2. Calibration factors for
individual common nuclides are prepared and available by the manufacturers of such detectors.
This is the only possibility to approximate the surface-related activity (as well to convert
impulses to the unit Bq/cm2).
Advantage: The detection of adherent contamination is possible.
Disadvantage: γ-Radiation and strong β--radiation obstruct the detection of further
contamination by α- and weak β--emitters, which may also be present.
1
2. Indirect measurement in the inspection of surfaces
This procedure is mainly used when the expected contamination is so small that it could be
overlooked or its direct measurement could be disturbed by strong β--radiation or γ-radiation
when scanned with a radiation detector.
With this method, a filter paper is wiped on the investigated surface spiralling from the outside
proceeding inward to avoid a spread of contamination. Gloves must be worn when performing
these so-called wipe tests. Afterwards, the filter paper is measured with a suitable detector (e.g.
methane flow counter for α- and β--contaminations.)
Advantage: No interference from nearby emitters is encountered. Small contaminations can be
better detected.
Disadvantage: Adherent contaminations cannot be ascertained.
If contamination is detected by either the direct or indirect method, it is to be removed by proper
means. Contamination levels are to be kept as low as possible and, unconditionally, below the
limits defined by the German law (Strahlenschutzverordnung § 57).
Surface decontamination can be performed with various methods. In each case, the used method
depends on the nuclide causing the contamination, the surface and the chemical form of the
contamination. When dealing with radioactivity, surfaces which are difficult to decontaminate
should be avoided.
Commonly used decontamination agents are e.g., water, soap solution, complex forming agents,
diluted acids and bases or alcohols.
Care must be taken that the employed decontamination agent does not attack the surface itself.
Furthermore, the application of organic solvents is not advisable since they commonly diffuse
together with the contamination into the surface. As a result, the decontamination becomes
firmly fixed into the material and can no longer be removed.
3. Determination of surface-related activity
To assess the degree of contamination, the activity related to the surface must be determined.
The approach to determining this activity depends on whether direct or indirect methods are
used in the measurement.
3.1. Determination with direct measurements
In the quantification of surface contamination, a calibrated detector must be used. If the
calibration factor of the used nuclides for the employed measuring device is not available, it
must be determined with a calibration emitter of known surface-related activity as shown in
equation 1.
Kn =
An
(R M − R 0 )
Kn: Calibration factor of nuclide n
(1)
An: Surface-related activity
RM: Gross counting rate of the substance
R0: Background of the detector
Attention: Such calibration factors depend on the distance between detector and sample. They
are valid only at the distances employed in the calibration.
2
3.2. Determination with indirect measurements
With the aid of wipe tests, surface-related activity can be assessed. First, a methane-flow
detector must be calibrated for this measurement. For this purpose, three samples of known
activity are gauged (14C, 60Co, and 90Sr). With the use of equation (2), the β-detection capacity
εβ of the detector can be calculated for each nuclide.
R − R0
εβ = M
A
(2)
RM: Gross counting rate of the substance
R0: Background rate of the detector
A: Activity of the substance
Between the counting rate RM produced by wipe tests of the samples and the surface-related
activity AF exists the following relationship:
AF =
R M − R0
s ⋅ F ⋅εβ
(3)
s: Extraction factor
F: Area of the inspected surface
The surface F inspected for contamination should have an area of 100 - 300 cm2. With a wipe
sample, the extraction factor gives the contained amount of activity. This extraction factor
depends on the chemical composition of the nuclides responsible for the contamination and the
structure of the surface on which the contamination is found. The nuclide can adhere tightly to
the surface as a result of a chemical reaction, be bound due to adsorption or be incorporated
into the pores of the surface leaving only a small amount that can be removed with the filter
paper.
Therefore, the extraction factor must be experimentally determined. For example, verification
would be possible if the available contamination is assessed before and after the wipe test with
a suitable detector.
If no such verification of the extraction factor exists, it must be estimated. When in doubt, the
factor of the corresponding DIN-Norm (DIN ISO 7503 Part 1) is used. In this case, the
extraction factor amounts to 10%.
3
Experimental
1. Precautions
There is no danger due to external irradiation resulting from the samples, since all of them are
of low activity. All usual precautions are applied in order to avoid contaminations or
incorporations.
2. Procedure
2.1. Direct measurement of the contamination of surfaces
1. In a plastic tub, four different plates were prepared; a metal plate, a plastic plate, a
linoleum plate and a wooden plate. Each of the materials is contaminated with aqueous
99m
Tc solution.
With the help of a large-area proportional counter, ascertain the extent of the surface
contamination!
Which difficulties do arise during this procedure?
2. Examine the decontamination ability of the surfaces with water, ethanol and acetone!
Determine the percentage of remaining radioactivity on the surface with equation (4)!
R
FD = n 2 ⋅ 100
Rn1
(4)
FD = percentage of remaining radioactivity
Rn1 = Impulse rate before decontamination
Rn2 = Impulse rate after decontamination
2.2. Indirect measurement of the contamination of surfaces
1. With the aid of a Geiger-Müller counter, examine the nine fields in the plastic tub for
contaminations and note your results.
2. Assign each of the fields in which you have detected a contamination with the
responsible emitter (14C, 60Co, 90Sr). Various shielding materials are available at your
disposal.
3. A wipe-test of each field was prepared and measured in a methane flow counter. Copy
the count number and the measurement duration into the following table. Calculate the
results of the wipe tests for the individual fields of the plastic tub along with their
surface-related activity! The surface area of the wiped fields is 100 cm². The activities
of the standard samples for the calculation of the β-detection capacity εβ are found in
data sheet 2.
4
4. Discuss by means of the German law “Strahlenschutzverordnung“ (§ 57) required
consequences. For which nuclide is a decontamination required? Which dose limit is
useful? How to deal with contamination during daily lab work?
5
Data sheet 1: Decontamination of radioactively contaminated surfaces
Direct Measurement
Contamination with 99mTc
Decontamination agent:
Surface Contamination
Wood
Linoleum
Plastic
Metal
Before Decontamination
After 1st Decontamination
Percentage remaining
radioactivity after the 1st
Decontamination
After 2nd Decontamination
Percentage remaining
radioactivity after the 2nd
Decontamination
Indirect Measurement
Contamination with 14C, 60Co and 90Sr
Geiger-Müller counter
I
II
III
IV
V
VI
VII
VIII
IX
Data sheet 2: Decontamination of radioactively contaminated surfaces
Detector: Methane flow counter (α + β-Plateau)
6
High voltage: 2100 V
t
[min]
N
RM
[min-1]
RN
[min-1]
2 σR
[min-1]
Background
β--detection capacity of the detector εβ
3700 Bq
14
C
56486
5
3700 Bq
60
Co
250745
5
3700 Bq
90
Sr
921925
5
Results of the wipe tests
Field I
Field II
Field III
Field IV
Field V
Field VI
Field VII
Field VIII
Field IX
7
εβ
AF
[Bq⋅cm-2]
Chapter 3
Calculation of the γ-radiation dose rate
Task
1. Determine the dose rate of a γ-emitting sample (60Co) as a function of distance from the
source by using the measuring device GR-135.
2. Determine the dose rate of a γ-emitting sample (60Co) as a function of distance from the
source by using the measuring device Colibri VLD.
3. Determine the dose rate of a γ-emitting sample (60Co), upon the application of various
shielding materials, at distances of 5 cm and 10 cm.
4. Calculate the dose rates corresponding to distances of 5 and 10 cm from a 60Co source!
5. Calculate the dose rate corresponding to a distance of 5 cm from a 60Co source in the
presence of shielding with 8 mm thick lead and iron absorbers!
6. Calculate the effective body dose to which a person is exposed when working for a year
at a distance of between 5 and 10 cm from a 60Co source!
Fundamentals
1. Protection precautions when handling radioactive materials
When radiation strikes matter, ionization processes occur. Such occurrences in living organisms
can be seriously detrimental. In order to abate damages due to radiation in a given radiation
field, principles have been defined by law.
To minimize the radiation dose, the mnemonic aid DATS is applicable.
1.
D as in Distance
The Distance from the radiation source should be as large as possible (Fig. 1).
2.
A as in Activity
The Activities used in radiochemical experiments should be kept as low as possible.
3.
T as in Time
The experiment is to be planned in such a way that the Time spent working in the
radiation field is minimized.
4.
1
S as in Shielding
Always arrange suitable Shielding materials between the radiation source and
persons (Fig. 2).
Fig. 1: Effect of distance from a radiation source on the dose. Note, that an increase of
the distance by a factor 2, decreases the dose by a factor 4!
a.) without shielding
b.) with aluminium plate
c.) with lead plate
Fig. 2: Effect of shielding the radiation source on the dose
2. Dose definitions
For the description of the dose, which is transferred to a body by ionizing radiation, there are
different dose concepts to be used. They are practically very different and each of these concepts
has different uses. Table 1 gives an overview of some common dose magnitudes and their
relevance.
2
Types
Ion Dose I
Background
Describes the number of charges formed per mass of air.
Energy Dose D
Indicates how much energy is absorbed per mass. It is a J⋅kg-1 = Gy
physical quantity. It does not distinguish between the
types of radiation.
Considers the relative biological efficacy of the different J⋅kg-1 = Sv
types of radiation.
H = D · wR
Equivalent Dose H
Effective Dose
Unit
C⋅kg-1
for wR: α = 20; β = 1; γ = 1; n = 5 - 20
Is the sum of the tissue-weighted equivalent doses of the J⋅kg-1 = Sv
human body. In addition to the biological efficacy of the
different types of radiation, it also takes into account the
different radiation sensitivities of the organs. It is used as
a dose value fixed by the legislator as a maximum to
which a person may be exposed (Strahlenschutzverordnung).
E = ƩwT·HT
Table 1: Overview of relevant dose rates
3. Calculation of the energy dose rate for γ-radiation
The equation used to calculate the energy dose rate (absorbed dose rate) for γ-radiation is as
follows:
A: Activity of the substance [Bq]
 = A ⋅ I ⋅ r −2
D
γ
(1)
Iγ: γ-Radiation constant [Gy⋅m2⋅h-1⋅Bq-1]
r: Distance between the substance and
measuring device [m]
D : Energy dose rate [Gy⋅h-1]
For equation (1) to be valid, the radiation source must be defined as a point source and the
radiation source and the measuring device are only separated by air (no further shielding). The
interaction of the γ-quanta with air particles can be neglected.
The γ-radiation constant Iγ is a function of the energy of the different γ-quanta of the considered
nuclide: Iγ = f (ΣEγ)
A shield is located between the radiation source and the measuring device, where it
exponentially decreases γ-radiation as a function of the thickness and composition of the
absorber. The formula used to calculate the dose rate of γ-radiation with shielding is derived
from equation (1) with the multiplication of the attenuation term e-μd:
 = A ⋅ I ⋅ r −2 ⋅ e −µd
D
γ
3
(2)
d : Mass per unit area of the absorber [g⋅cm2
]
µ : Mass extinction coefficient [cm2⋅g-1]
d* : Thickness of the absorber [cm]
where d = d*⋅ρ
ρ : Density of the absorber material [g⋅cm-3]
µ* : Linear extinction coefficient [cm-1]
where µ = µ*⋅ρ-1
If the radiation source is partially enclosed,
a portion of the radiation is reflected by the
surrounding material (Fig. 3). This leads to
an increase in the measured γ-radiation. In
the calculation of the γ-dose rate, the
increased portion of γ-radiation is taken
into account by the so-called build-up
factor B. This factor can also be applied to
indicate the degree of γ-quanta dispersion.
The build-up factor is both a function of the
energy of the γ-quanta as well as the
reflecting material. This value must be
Fig. 3: Reflection of γ-radiation
specifically
determined
for
each
experimental assembly with the use of
calibrating substances.
The formula used for the calculation of the energy dose rate for γ-emitters with shielding and
build-up factor considerations is as follows:
 = A ⋅ I ⋅ r −2 ⋅ e −µd ⋅ B
D
γ
(3)
4. The γ-dose rate measuring devices
4.1 GR-135 (SIAC Exploranium)
The dose rate measuring device GR-135 can be used for photons in the energy range of 50 keV
to 3 MeV. This device is capable of measuring dose rates in the range of 0.01 μSv to 50 μSv.
A NaI-scintillator serves as the radiation detector. Every radiation measuring device undergoes
a warm-up phase after switching it on. A temperature difference of 60 ˚C leads to a variation in
value of about +/- 15%. Hence, it is important that the measuring device is calibrated prior to
actual use. This so-called stabilization of the measuring device is performed by the GR-135
with a 137Cs calibration emitter. During the stabilization, an internal energy calibration of the
device is conducted. The gain of the measuring device is checked and adjusted if necessary. If
the temperature fluctuates during the measurement, the device corrects the fluctuations
automatically.
4.2 Colibri VLD (Canberra)
The dose rate measuring device Colibri VLD can be used for photons in the energy range of 59
keV to 1.5 MeV. Compared to the measuring device GR-135, dose rates in the range of 0.01
4
μSv to 1 mSv can be detected. The radiation detector is a CsI (Tl) scintillation crystal. This
measuring device does not have to be calibrated, it can immediately be used.
5
1. Precautions
No direct danger of contamination exists. Substances are placed into the experimental set-up
by assistants.
2. Nuclides to be used
For calibration, a 137Cs sample is used (Eγ = 662 keV)
A(01.01.2005) = 9.25 kBq, t1/2 = 30.17 a
For the measurement, a 60Co sample is used.
A(30.12.2006) = 1.5 ⋅ 105 Bq, t1/2 = 5.272 a, Iγ = 3.41⋅10-13 Gy⋅m2⋅h-1⋅Bq-1
(Weighing factor: wγ = 1; Effective dose [Sv] = Energy dose [Gy])
3. Procedure
1. The two measuring devices are used in the rate-meter mode. The measured values,
related to the applied time constant, are continuously adapted to the measured result. To
read the measured values properly, please wait until the values are no longer
significantly changing.
2. Turn on the measuring device GR-135 with the joystick!
3. Select the menu item “Stabilize” and place the 137Cs source in the calibration opening.
Start the stabilization! Remove the 137Cs source after successful calibration! Return to
the main menu of the control panel!
4. Select the menu item “Search”! The device will now measure the corresponding dose
rate.
5. Measure the dose rate at the distances from the radiation source starting at 5 cm in 5 cm
intervals up to r = 100 cm! Make sure that the detector is directed towards the radiation
source!
6. Measure the dose rate at a distances of 5 cm and 10 cm from the source with each of the
8 mm thick absorbers of different materials (aluminum, lead, iron, plexiglass, PVC)
placed between the radiation source and the detection device!
7. Turn on the measuring device Colibri VLD.
8. Repeat steps 5 and 6 with this measuring device.
Evaluation
1.1 Calculate the current activity of the 60Co-containing sample to compare it later on with the
experimentally determined activities.
A = A0 · e-λ·t ; λ = ln 2 · t1/2-1
1.2 Plot the dose rates measured as a function of 1/r2! The slope m = ∆y/∆x of the resulting
line allows the determination of the γ-radiation constant when the activity is known or,
inversely, the activity when the γ-radiation constant is known as follows:
m = Iγ ⋅ A
Iγ = 3.41⋅10-3 µGy⋅cm2⋅h-1⋅Bq-1
Discuss, when a linear course of the order is to be expected!
1.3 Compare the calculated activity with the values obtained from the measurement!
6
2.
Calculate the dose rate of the 60Co-containing sample at distances of 5 and 10 cm from the
source without shielding! Use the activity, calculated in 1.1, for this task.
3.
Calculate the dose rate of the 60Co-containing sample at distances of 5 cm and 10 cm from
the source with the use of an 8 mm thick lead absorber as well as an 8 mm thick iron
absorber!
µPb =5.996⋅10-2 cm2⋅g-1
µFe =5.196⋅10-2 cm2⋅g-1
ρPb =11.34 g⋅cm-3
ρFe =7.86 g⋅cm-3
4.
Discuss the effectiveness of the absorber materials in regard to their reduction of the dose
rate! Discuss the influence of the shielding and the distance on the dose rate. What is the
effective protection against radiation damages?
5. In accordance with StrSchV (radiation protection regulation), a person of the category A
may absorb a maximum effective dose of 20 mSv per year. Assuming that the effective
body dose is equivalent to the local dose, how much of a dose is absorbed by a person who
works at a distance of 5 cm as well as a person who works at 10 cm from the source per
year? (The work time consists of 40 hours / week, 50 weeks / year). Is the legal annual dose
limit exceeded? Is a person allowed to work under such conditions?
Task
1.1
1.2
Magnitude
Result
Activity theoretically calculated
Activity determined by
meausurement
5 cm
2
3
5
7
Unit
D
D with 8 mm lead absorber
D with 8 mm iron absorber
Dose
10 cm
Data Sheet: Dose rate measurement
Equivalent dose rate [µSv⋅h-1]
Distance
[cm]
5
10
15
20
25
30
35
40
45
50
55
60
65
70
75
80
85
90
95
100
8
Dose rate
[µSv⋅h-1]
GR-135
Dose rate
[µSv⋅h-1]
Colibri VLD
Distance
[cm]
Absorber material
(8 mm)
Aluminium
Lead
5
Iron
Plexiglass
PVC
Aluminium
Lead
10
Iron
Plexiglass
PVC
9
Dose rate
[µSv⋅h-1]
GR-135
Dose rate
[µSv⋅h-1]
Colibri VLD
Chapter 4
Gammaspectrometry; recording of γ-spectra
Task
1. Perform an energy calibration of the NaI-scintillation detector by employing isotopes
60
Co and 137Cs!
2. Determine the γ-energies of 22Na! Discuss the γ-spectrum and identify the nuclide with
the help of nuclide libraries and the nuclide chart!
3. Determine the γ-energies of the unknown radioisotopes A and B! Discuss the γ-spectra
and identify the nuclides with the help of the nuclide libraries and the nuclide chart!
Fundamentals
γ-Spectrometry allows qualitative and quantitative analysis of unknown nuclides. Scintillation
sensors and semi-conductor detectors are employed as detecting devices.
1. γ-Radiation
γ-Radiation is emitted from the nucleus of an atom upon its transition from an excited state to
a state of lower energy.
The energy of γ-radiation lies between 10 keV and 104 MeV, which corresponds to an
approximate wavelength range of 10-1 to 10-7 nm. There are three types of interactions between
photons and matter: photo effect, Compton effect and pair formation.
2. Interactions of γ-quanta with matter
2.1. The photo effect
Fig. 1: Photo effect through the interaction of γ-quanta with matter
1
The photoelectric effect describes the interaction of a low energy photon with an inner-orbital
electron of an atom resulting in the complete transfer of the photon’s energy to the electron
(Fig. 1). This results in the ejection of the photoelectron, which at this point possesses an energy
equal to the energy difference between the initial γ-ray and the binding energy of the electron
to the atom.
The photo effect occurs predominately at the inner electron shell of the atom, where the binding
energy of such electrons is relatively large. Following the dissociation of the photoelectron from
the electron shell, a vacancy is available for a higher-energy electron of an outer shell to occupy.
The difference in binding energy between the higher and lower energy orbitals causes the
emission of X-rays. This resulting radiation is known as characteristic self-radiation or X-ray
fluorescence.
Due to the strong influence of the atomic number Z on photon absorption (pPhoto ≈ Z5), materials
with large Z (e.g. lead) are commonly used in the fabrication of radiation shields.
2.2. The Compton effect
The Compton effect occurs when, a photon of higher energy relative to the photo-effect strikes
an outer-orbital electron, conveying a portion of its energy. As a result of the encounter, the
flight-path of the γ-quant is altered (Compton scattering) and the electron is released from the
atom (Fig. 2). The resulting energy of the Compton electron is close to the energy transferred
from the photon to this electron, since the binding energy of the electron is small compared to
the γ-energy. This so-called Compton process occurs primarily at the outer-shell electrons of
the atom. As a result of the collision, the wavelength λ2 is greater than λ1. Depending on the
angle of scattering, quanta of different energies are obtained, which form the so-called
Compton-continuum of a γ-spectrum.
Fig. 2: Compton-effect through the interaction of γ-quanta with matter
2
2.3. Pair Formation
In the strong electric field near the atomic nucleus (Coulomb field), γ-quanta with an energy of
1.02 MeV or greater can be converted into an electron and a positron. In this process of
“materialization”, the electron-positron pair develops (Fig. 3). The sum of the resting masses
of the positron and the electron corresponds to an energy of 1.02 MeV. Therefore, the mass of
each of these particles is equivalent to an energy of 0.51 MeV. If the energy of the photon is
greater than 1.02 MeV, the excess energy appears in the form of kinetic energy of the positron
and negatron (electron). Thereby, the energy surplus is distributed equally to both particles
according to the law of momentum conservation. The generated positron has a short lifetime
before it combines with an electron in a process reciprocal of materialization, annihilation. As
a result of annihilation, the mass of the elementary particles are converted into radiation energy
as two photons are generated with an exact energy of 0.51 MeV each. These photons, produced
from pair production and the subsequent annihilation can proceed to interact with matter via
Compton or photoelectric processes.
The probability for the incidence of pair formation at the electrons of an atomic orbital is
proportional to the atomic number Z. The probability of pair formation at an atomic nucleus
(coulombic field) is proportional to Z2.
Fig. 3: Formation of an electron-positron pair through the interaction of γ-quanta with matter
3
3. Design of a NaI-scintillation detector
The fundamental components of a NaI-scintillation detector are the scintillation crystal, the
photo cathode and the photomultiplier (secondary electron multiplier). Further details are
described in Chapter 1 “Comparison of various detectors and nuclides”.
4. Single channel analyzer
Analysis of the counts is carried out via frequency discriminators (upper and lower data
thresholds are to be set) and an anti-coincidence level. Only such impulses in which the voltage
peak exceeds the lower threshold value and is less than the upper threshold value, are forwarded
to the counter.
The strength of the impulse is proportional to a defined energy. Therefore, this measurement
set-up permits the recording of the following spectra.
Integral spectrum
Differential spectrum
Fig. 4: Impulse rate as a function of energy
A single-channel analyzer enables the breakdown of the entire energy range into 1000
individual channels. Furthermore, the width of an energy range can be defined by the degree of
amplification. For example, a range from 0 to 1000 eV can be measured with 1000 channels (1
channel corresponding to 1 eV) (Fig. 5a), or a range from 0 to 3000 eV can be measured with
1000 channels (1 channel corresponding to 3 eV) (Fig. 5b).
Fig. 5: Impulse rate as a function of energy
4
Moreover, it is also possible to scan the content of a certain number of channels simultaneously.
At this point, one talks about the “window” which has been set. For this, a lower energy level
LL and an upper energy level UL are defined. Therefore, the measured data between two
channels are gathered.
5. Multi-channel analyzer
The recording of energy-resolved gamma spectra is today carried out practically and
exclusively with multi-channel analyzers. In order to simplify, a multi-channel analyzer can be
visualized as a large number of single channel analyzers connected in parallel. Whereas only
the impulses of certain energy intervals are counted with the single channel spectrometer, the
multi-channel spectrometer analyzes and stores almost every occuring impulse in the channel
of the corresponding energy. The measurements of these memory channels can be accessed in
succession and the content may be printed out or displayed on a screen.
An advantage, which multi-channel spectrometers have over single channels spectrometers is
that they immediately provide the entire spectrum distributed to a certain number of channels.
Calibration with γ-radiation of a known energy allows a comparatively exact reading of the
energy values (error approx. 1%).
Fig. 6: Scheme of a γ-radiation measuring station
Fig. 6 shows a scheme of a γ-radiation measuring station. It is mainly consisting of three
components: the detector, the electronic measurement system and the evaluation unit. As
detectors, semiconductor and NaI detectors can be used. The electronic consists of high-voltage
supply, preamplifier, pulse height amplifier, analogue-digital-converter (ADC) and
multichannel analyzer (MCA). The charge pulses, generated in the detector by the photons, are
first collected in the preamplifier and then multiplied. The signals are then passed to the pulse
height amplifier. There, the detector pulses are amplified and converted. Thereby, the pulse
height is proportional to the photon energy. The following multi-channel analyzer (MCA) is
the centerpiece of the system. It is composed basically of an analogue-digital-converter (ADC),
multichannel memory, control logic and display. It detects the incoming pulses, then measures
the voltage value and translates it into digital information. The contents of the MCA memory
are then copied into the memory of the PC, by using the software, and displayed as a spectrum
on the screen. Modern gamma-spectrometry measuring systems are controlled entirely by the
software of the computer. Here the necessary settings such as high voltage, amplification and
the evaluation can be carried out. For the nuclide identification various nuclide libraries are
5
available, which are helpful in the identification of the radiators. The activity can be determined
by a known activity of a calibration radiator.
6. Set-up of γ-spectra
The monoenergetic gamma radiation of a radionuclide leads by the interaction in the detector
(photo effect, Compton effect and pair formation) to a complex pulse height spectrum.
A characteristic feature of γ-spectra are “photopeaks”. Only with the photoelectric effect the
total
energy
of
the
γ-quanta
is
released in a single process. The photopeaks enable the identification of individual nuclides
with the use of appropriate tables or computer programs according to the “fingerprint” method.
Problematic for analysis, however, are coincidental dispersions in the background and the
overlapping of peaks.
For each photon emitted during the decay, a photopeak is to be observed in the spectrum.
Thereby, two photopeaks are expected in the γ-spectrum of 60Co according to the decay scheme
(Fig. 7). When evaluating the spectra, it is important to ensure that all peaks, which are observed
in the spectrum, have to be assigned. Otherwise, errors may occur in the interpretation of the
results.
Fig. 7: The decay scheme of 60 Co
The “Compton continuum”, which is observed at lower energies, is based on the Compton effect
and is not suitable for the identification of individual nuclides. Its position depends on the
maximum energy of the penetrating γ-quanta.
In the case of overlapping spectra, the contents of the channels in the range of the Compton
continuum are simultaneously summed. The lower the energy of incoming quanta, the higher
are the effects of the background radiation. Therefore, an evaluation of the low energy range of
the spectrum is difficult or not possible.
6
7. Comparison between semiconductor and NaI detectors
Fig. 8: γ-Spectrum of 60Co recorded with NaI and Ge(Li) detectors
Fig. 8 clearly shows that considerably smaller linewidths can be obtained with a semiconductor
detector. Improved resolution of the sprectra is especially important when the energies of γquanta lie close to each other. In this case, semiconductor detectors are especially well suited
for qualitative analysis. In general, a lower number of counts are yielded with a semiconductor
detector than with a NaI-scintillation detector.
The high response probability of a NaI-scintillation counter is advantagous in quantitative
analysis. NaI-scintillation detectors have a better response capacity to radiation of lower energy
than higher energy radiation. However, one must keep in mind that in the range of very low
energies the linearity pertaining to the energy of the penetrating radiation is not always
guaranteed.
8. Energy calibration
The width of the pulse height spectrum, and thus also the resolution and the position of the
individual energies, depend on different settings in the detector. The high voltage, with which
the detector is operated, as well as the setting of the amplification both cause changes of the
spectrum. For this reason the assignment of the channels of the corresponding energies have to
be first done for every detector, this can be achieved by using gamma emitters of known energy.
If the energy calibration is carried out, this is only valid as long as no changes are made to the
settings of the high voltage and the coarse and fine gain.
7
Experimental
1. Precautions
Because substances of low activities are used, no external danger due to radiation fields are
present. The substances are not to be touched with the hands. Tweezers must be used.
2. Procedure
Energy calibration of NaI-scintillation detectors
1.
2.
3.
4.
5.
6.
7.
8.
9.
Start the computer and open the "Prospect" measurement program, which you will find on
the desktop.
Select "Technican" as the user. The password is "welcome".
Click "Connect to device" and select the detector "Gammaspek1b.10.0.1.4". Click "take
over".
In the "Detector" submenu you can now select the high voltage of the detector. Set on the
value to 700 V. As soon as you click "on", the high voltage is switched on.
The gain of the detector can be set in the "MCA" submenu. Set the number of channels to
be measured on 1024. With "coarse gain" and "fine gain" you change the amplification.
Select x1 for this.
In the "Acquisition" submenu you can now set the mode and the time of the measurement.
Select PHA (= Pulse Height Analysis) as the mode. Select "live time" and 300 s as the
measuring time.
Place the calibration radiators 137Cs and 60Co on the detector!
Click "Start" to start the measurement. You should now see a spectrum with three signals.
Test the influence of the high voltage and the gain of the detector on the position and
resolution of the γ-lines. Repeat the measurement with the settings in Table 1. From your
results, select the settings, where the photopeaks of 60Co are in the second half of the
spectrum. Discuss the settings with the assistant.
These settings remain unchanged until the end of your experiment.
Spectrum 1
Spectrum 2
Spectrum 3
Spectrum 4
Table 1: Settings
8
High voltage
Coarse gain
Fine gain
High voltage
Coarse gain
Fine gain
High voltage
Coarse gain
Fine gain
High voltage
Coarse gain
Fine gain
700V
x1
x1
700V
x8
x5
600V
x1
x1
600V
x8
x5
10. Measure an impulse height spectrum of the 60Co and 137Cs radiators with the selected
settings.
11. Now perform an energy calibration for the selected settings. Select the "calibration"
submenu. You can now select the first peak height point with the left mouse button. This
should now be shown in green. Select the peak with the right mouse button "use this
ROI". Specify the appropriate power 137Cs (γ) - 662 keV. Proceed in the same way for
the signals for 60Co (γ1) - 1173 keV and 60Co (γ2) - 1332 keV!
12. Select "close calibration view and return to spectrum view" in the upper right corner. The
detector is now calibrated.
Determination of the energies of 22Na
1. The calibration emitters are exchanged by 22Na.
2. Click "Clear".
3. Click "Start" to start the new measurement.
4. Now determine the γ-energies of 22Na.
5. Compare your result with the nuclide chart and the interactive nuclide libraries. Click on
"Analysis" and select a suitable nuclide library!
Determination of the energies of the unknown radioisotope A
1. 22Na is exchanged by the γ-radiator A.
2. Click "Clear".
3. Click "Start" to start the new measurement.
4. Now determine the γ-energies of (A).
5. Compare your result with the interactive nuclide libraries for the identification of
γ-radiator A.
Determination of the energies of unknown radioisotope B
1. The γ-radiator A is exchanged by the γ-radiator B.
2. Determine the nature of γ-radiator B.
Evaluation
1.
2.
3.
4.
5.
6.
7.
9
Discuss why you need to calibrate the detector.
Calibrate the detector by taking an impulse height spectrum of 137Cs and 60Co. To do this,
test the settings for the high voltage and the gain of the detector and select the most suitable
one.
Determine the γ-energies of 22Na.
Compare your results with the interactive nuclide libraries. Which one is most suitable?
What is the influence of the selection of the nuclide library on the evaluation of the γspectrum?
Compare your result for 22Na with the nuclide card. Look for the γ-energies for 137Cs and
60
Co. What did you notice? Discuss your observations with your assistant.
Determine the γ-energies of the unknown emitter A. Which radiator could it be? Analyze
your results critically. What are the difficulties in identifying the radiator? Which nuclide
library do you choose?
Determine the γ-energies of the unknown emitter B. Analyze your results critically. What
are the difficulties in identifying the radiator? Which nuclide library do you choose?
Measurement sheet: Gamma spectrometry
Coarse gain
High voltage [V]
Radiator
137Cs
60Co
22Na
Fine gain
Emitter A
Emitter B
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Energy γ (keV)
Radiator
22Na
Nuclide library
Isotope
10
22
Na
Probe A
Probe B
1
Chapter 5
Betaspectrometry (Liquid Scintillation Counter)
Task
1. Determine the response probability of the liquid scintillation detector for the nuclides
3
H and 14C.
2. Record β–-spectra of the nuclides 3H and 14C and plot the counting rates of the standard
substances as a function of the channel.
3. Calculate the contributions of 3H and 14C to the β-radiation of an unknown sample with
the help of the net counting rate. Compare the calculated values to an automatic determination.
Fundamentals
1. β--Spectrum
The energy of the fission of a neutron to a proton (n → p+ + e– + ν ) is divided between the
electron and the antineutrino following the linear momentum and energy conservation law.
Therefore, when a great number of nuclei undergo β–-fission a continuous energy spectrum of
β– particles is obtained.
Fig. 1: Schematic distribution of energy of β– particles after the 14C resp. 3H decay
The maximum energy of a β– particle is constant for the decay of a specific nuclide. The average
energy is about a third of the maximum energy. Often a logarithmic scale of the energy is used
when plotting the β– spectra.
1
2
2. Fundamentals of liquid scintillation spectrometry
The exact measurement of the β–-radiations of 3H and 14C is of particular interest for the solution
of biological and biochemical problems. Both nuclides emit low-energy β– particles, which require special measuring techniques, in particular when the obtained counting rates are low.
The most suitable method for this is liquid scintillation counting. The principle of a liquid scintillation counter is based on the interaction of the radiation of a dissolved β–-emitting radioactive nuclide with a scintillation substance in the same liquid phase. For this, a so-called “scintillation cocktail” is used.
2.1. The scintillation cocktail
The cocktail is composed of a solvent, a primary scintillator and a secondary scintillator.
2.1.1. Solvent
Common solvents, such as water, acetone or alcohols, are unsuitable as solvents for the radioactive substance, since the radioactivity can cause chemical processes with the solvent molecules (radiolytic processes). Therefore, more inert solvents like benzene, toluene, xylene or
methoxybenzene are used. They bear characteristic π-electron systems which can readily be
excited by β–-particles and show a high quantum yield. Detergents are formulated to provide a
good solubitily of polar components in the unpolar solvents.
2.1.2. Primary scintillator
Primary scintillators are fluorescent dyes (e.g. PPO: 2,5-Diphenyloxazole, Diphenylbutadiene).
These complex aromatic molecules absorb radiation with short wave lengths and emit the energy in form of a radiation of a longer wavelength. They are used in a concentration range of
10–2 to 10–3 mol·l–1.
2.1.3. Secondary scintillator (“Frequency changer”):
The secondary scintillators shift the wave lengths of the emitted light (between 340 nm and 480
nm) to slightly longer wave lengths to meet the range of the highest sensitivity of the photocathode (between 400 nm and 500 nm). Examples for such substances are POPOP (1,4-Bis(5phenyloxazol-2-yl)-benzene) or Dimethyl-POPOP. The wavelengths obtained are between 400
and 550 nm.
PPO
Dimethyl-POPOP
Bis-MSB
2.2. The mechanism
The mechanisms of the energy absorption and the light developing process in a liquid scintillator system are extremely complex. The following description is strongly simplified and summarizes the steps from the interaction of a charged particle with the solvent in the cocktail up
to the light developing process:
2
3
Fig. 2: Interaction of the β--particles with the aromatic solvent and the emission of
fluorescent radiation
β--particle
e–
o
Solvent molecule in the initial state
•
Solvent molecule in the excited state
2.2.1. Collision of a β--particle with a solvent molecule
The β--particle transfers a part of its energy to the π-electron system of the solvent molecule
(L), which changes into an excited state (L*). The β--particle collides with many different solvent molecules, because only a few eV per collision are transferred; the energy of the β--particle
is generally several hundred keV. 105 to 106 collisions are expected before the β--particle has
released all of its energy and is captured by a molecule.
L + β– → L* + β–
2.2.2. Energy transfer between the solvent molecules / phosphorescence
The excited solvent molecules transfer their energy to neighboring solvent molecules.
L* + L’ → L + L’*
However, it is also possible that its excitation energy is emitted as light (e.g. in the range of
260-340 nm; phosphorescence).
L* → L + h·ν (260 - 340 nm)
Direct measurement of phosphorescence is difficult and not very accurate. It’s better to try to
change the wavelength by chemical means.
2.2.3. Absorption of the phosphorescence radiation by the primary scintillator
The radiation emitted by the solvent molecules is absorbed by a primary scintillator (P), which
changes into an activated state (P*).
3
4
P + h·ν (260 - 340 nm) → P*
In contrast to the solvent, the primary scintillator cannot transfer its energy to further molecules.
It releases energy exclusively in the form of fluorescence (e.g. within an energy range of 340480 nm).
P* → P + h·ν (340 – 480 nm)
2.2.4. Absorption of the fluorescence radiation by a secondary scintillator
When the fluorescence spectrum of the primary scintillator does not correspond with the absorption spectrum of the photo cathode, a secondary scintillator (S) is added to the cocktail.
This secondary scintillator works in the same way as the primary scintillator.
S + h ⋅ν 1 → S *
S * → S + h ⋅ν 2
It absorbs radiation within a distinct range (e.g. 340 nm to 480 nm) and emits radiation with
higher wavelengths compared to the primary scintillator (e.g. 400 nm and 550 nm).
The quantity of light emitted by the scintillator is extremely small. Highly sensitive detectors
are needed for the registration. Secondary electron amplifiers are used for these purposes.
2.3. Registration of a β--spectrum
The amount of light emitted from the scintillator is very little and therefore, a detector with a
very high sensitivity is needed. The photomultiplier transforms a light signal into an amplified
amount of electrons that are easier to detect.
The photocathode is coated with a light sensitive material that emits electrons when photons of
the fluorescence from the scintillators are absorbed. These electrons are accelerated by an electric field (100-200 V) towards a positive charged dynode where each single electron is amplified to 3-5 electrons. Negativly charged focus rings ensure that every electron is amplified. A
set of 12 dynodes further increase the amount of electrons by a factor of about 106.
The electric impulse is then analyzed by a discriminator. Different settings of the discriminator
enable to detect e. g. specific signal intensities (that are proportional to the energy of the β–
particle). The discriminator only forwards very specific signal intensities and can be divided
into 1000 different intervals (channels). An electric impuls counter is connected to the discriminator to analyze the counts.
2.4. Influence of the energy of the β- particle
After a β– decay the β– particle collides with a to its energy proportional amount of solvent
molecules. During each collision a similar amount of energy is transferred. The solvent molecules transfer the energy as described above. The overall process takes place in a time interval
of about 10–6 seconds. The signal intensity that is measured at the photocathode within this time
interval (<10–6 s) is proportional to the energy of the β– particle. The concentration of the sample
must be chosen that less then one decay is happening during that time slot. In case that more
than one decay appears within 10–6 seconds then the discriminator can not distinguish between
the two dacays and counts them as one.
4
5
2.5. Influences on the counting efficiency
The counting efficiency is influenced by different side effects. Contributions for the measured
background rates beside the dark current are cosmic radiation and naturally occuring radioactivity from the measuring device (e.g. 40K-radiation of the glass walls). Note! Sometimes the
radioactivity of the container material is higher than the radioactivity of the sample.
For standard substances frequently counting yields of less than 100 % are observed. This is
caused by:
2.5.1 Signal loss in the scintillation counter:
This contribution is small and mainly appears as loss in the photo cathode. It is usually either
caused when photons only have a low energy to not be able to cause the emission of an electron.
Secondly, it can happen that the amount of photo electrons from the photocathode is so little,
that it is less than the lower limit of the discriminator and, therefore, is not detected. This loss
can easily be accounted for by measuring of calibration standards of a known activity and then
be corrected. The response probability η (in %) is given by:
η=
R N ⋅100
A
2.5.2. Losses due to quench effects
More important is quenching, which influences the fluorescent dyes and decreases the counting
yields significantly. These quench effects shift the whole β– spectrum. Therefore, the maximum
energy can be shifted to lower energies as well as the integral counting efficiency can be reduced. Figure 3 shows a non-quenched curce “C” and two quenched curves “B” and “A” where
the amount of quenching increases from “B” to “A”.
5
6
Fig. 3: Non-quenched β– spectrum “C” and two quenched spectra “B” and “A”
2.5.3. Loss by primary or chemical quenching
Chemical quenching is caused, for example, by strong acids or bases. They chemically destroy
the organic molecules of the scintillator and, thus, the concentration of the scintillator substance
decreases. Similar conditions must always be applied for preparation of different samples to
being able to compare different results.
2.5.4. Loss by secondary or color quenching
Parts of the radiation are absorbed during the scintillation process. Thus, they are not registered
by the photo cathode. Any substance with absorptions between 200 nm and 500 nm (yellow
color!!) causes color quenching. Therefore, the result of colored samples must always be treated
with caution.
2.6. Corrections of deleted counts
The counting yield and the counting rate of a sample strongly depend on the quantity and the
concentration of quenching substances. Internal and external standardization are used to maintain comparable results for solutions containing various concentrations of compounds. Refer to
the literature for more information!
2.6.1. Internal standardization
First, the sample is analyzed in the scintillation counter. Second, an exactly known amount of
the same radionuclide is added to the sample with a volume as low as possible (volume error!)
and the sample is measured again. The added amount of radioactivity must be substantially
larger than the amount of the sample (20x larger is usually suitable). After measurement of this
“standardized” sample the net counts that arize from the added activity are easily calculated by
the difference (before and after addition of the activity). As the added activity is exactly known
the amount of quenching can be calculated. The exact activity of the sample then is calculated
from the response probability and the quenching rate. Even though, this method is very exact it
has some disadvantages:
very small amounts of solvents have to be pipetted
every sample must be processed and measured twice (time-consuming)
- radioactive standard solutions are very expensive: There are additional radioactive
wastes whose disposal is also expensive.
6
7
2.6.2. External standardization
An external standard is a long-life γ-emitter that is placed in close proximity of the sample. The
γ-radiation generates Compton electrons in the scintillation solution (compare to “gamma-spectrometry”) that are approximately quenched similarly to the electrons of the β–emmiter of interest. A calibration curve is prepared for the correction of the amount of quenching. Herefore,
samples with different concentrations of the quenching substance in the scintillation cocktail
are recorded and the logarithm of the counting rate is plotted towards to different concentrations. Usually slightly bent calibrations curves are obtained from which the amount of quenching can be extracted when the concentration of the quenching substance is known. Usually 137Cs
or 133Ba are used as external standard that are supplied by the manufacturer of modern liquid
scintillation counters.
2.6.3. Channel proportion method
Herein, the change of the β– spectrum when being quenched is used (compare to figure 3). Two
different channel intervals are set that should have similar counting rates in the not quenched
state, which are ideally right and left of the maximum of the β– spectrum (most abundant energy). For an increasing quenching rate both, the maximum energy and the most abundant energy, move to lower energies. The quotient of both channel intervals is formed:
𝑅𝑅𝑐𝑐ℎ𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎 𝐵𝐵
= 𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞𝑞
𝑅𝑅𝑐𝑐ℎ𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎𝑎 𝐴𝐴
For similar impulse rates (non-quenched sample) this quotient is around one. Whereas for increased quenching the β– spectrum is shifted to lower energies, the impulse rate for channel A
(lower energy) increases and the impulse rate for channel B (higher energy) decreases strongly.
Therefore, the quotient decreases for quenched samples. A plot of the quotient towards the
concentration of the quenching substance gives a similar calibration curve as achieved with the
external standardization.
This method has the disadvantage that for samples with strong quenching the counting rate of
channel B (higher energy) is very low and often requires either a very long time to be measured
exactly or a high activity is needed. The channel intervals A and B must be selected that this
only occurs for higher quenching rates.
2.6.4. Channel proportion method with external standard
This is the combination of the two methods as it has been described before. The channel proportional method is used for the determination of the quenching rate in the Compton spectrum.
Compton electrons are emitted when a long-life γ-emitter is placed in close proximity to the
scintillation cocktail. Different concentrations of quenching substances show the same shift as
a β– spectrum. Therefore, the plot of the quotient towards the concentration of quenching substance results in a similar calibration curve as achieved by the external standardization or the
channel proportion method.
2.7. Registration of two β--emitters in one cocktail
An overlap of different β--spectra is observed when mixtures containing more than one β-emitter are studied. This causes considerable problems in quantitative measurements. For the
determination of the individual contributions of 3H and 14C in an isotope mixture, energy windows are selected in such a way that the registered impulses are exclusively (or at least predominately) from one of the nuclides of interest. The experimental net counting rate is recorded for
7
8
each isotope standard (3H or 14C) of known activity in the defined windows. Herewith, the
concentration of the corresponding nuclides can be calculated.
Fig. 4: β–-spectra of 3H and 14C with two possible windows
8
9
Experimental
1. Precautions
There is no danger due to radiation fields due to the small activities. The glass bottles with the
cocktails should not be exposed to light in order to prevent undesired fluctuations of the measuring rates.
2. Nuclides to be used
3
H:
C:
14
A(11.04.1974) = 2450 Bq
A(11.04.1974) = 855 Bq
t1/2 = 12.3 a
t1/2 = 5730 a
Emax = 18 keV
Emax = 158 keV
3. Procedure
Part I:
Determination of the response probability of the liquid scintillation detector
for the nuclides 3H and 14C.
1. Load rack No. 1 with the samples (background, 3H standard, 14C standard, unknown sample) in the positions mentioned in the table on data sheet 1.
2. Place the rack in the measuring device on the right hand side and close the device.
3. Wait 5 min and familiarize yourself with the use of the software at the computer during
that time (see sheet with software explanation). The software must be set to ready-torecieve before starting the measurement.
4. Start the first measurement after the 5 min as soon as the software is ready to receive data.
Click on “Count Single Rack” and “Select User Program” and choose the measureing
program (see data sheet 1). Start with the program “4 SPEKTRUM”.
5. Confirm the program with “SELECT” and start the measurement with the “START” button.
6. Each sample is now measured for 1 min and the spectra saved on the computer. The overall
counting rates are also saved to a different file. Measurements are performed in the
“WIDE” mode, meaning that it is measured over the whole channel region (1-1000).
7. Calculate the response probability for both isotopes.
Part II:
Plotting of the β–-spectra in Excel and determination of the windows
1. Import the spectrum data of the 3H and 14C standards to one MS Excel sheet. Use the corresponding files “SNMMDDNN” as described in the section “At the computer”.
Example:
use the files SN011502 and SN011503 for the 2nd and 3rd measurement
on January 15 (0115).
2. Normalize the spectra on the maximum of each spectrum.
3. Plot the curve as a line without points in a xy-type graph for the normalized 3H and 14C
standards.
4. Determine intervals for the windows A and B for the measurement of the unknown sample.
Choose window A as large as possible but take care that only 14C counts are measured in
this window and none of 3H. Choose the interval for window B so that it contains as many
9
10
5.
counts of 3H as possible with having as little of 14C as possible at the same time. The maximum of the 14C spectrum must be in window A and the maximum of 3H in window B. The
windows may not overlap.
Discuss the choosen windows with your lab assistant.
Part III:
1.
2.
3.
4.
5.
6.
For measureing in the range of the windows choose “Review and Edit User Program”,
choose the first program “1 PRAKTIKUM” and confirm with “SELECT”. Change the
window range settings at “Isotope 1” and “Isotope 2” according to your selection.
To set the window range change the values for “lower level, upper level” (e.g. “0,500” for
choosing the channel range between channel 0 and channel 500 for window A). Save your
settings each time with “SELECT”.
After changing the settings save the program by pressing “Previous Menu” until “User
Program Is Being Stored …” is displayed for a short period of time.
Click on “Main Menu” and start the measurement as described before by going on “Count
Single Rack” and choosing the program “1 PRAKTIKUM”.
The counting rates are now measured in the windows. “WIND1” is corresponding to window B and “WIND2” to window A.
Perform part III of the Analysis.
Part IV:
1.
2.
3.
4.
10
Measurement in the choosen windows
Automatic measurement of the nuclides 3H and 14C
Start a measurement for the automatic determination of the activities of 3H and 14C.
Click on “Count Single Rack” and choose the user program “3 AUTOISOTOP”.
You obtain automatically determined activities for both nuclides of all samples.
Compare these values with the values you calculated yourself. Do they differ? If yes, what
could be the reason?
11
At the computer: Software to receive data from the liquid scintillation counter
1.
2.
3.
Start the software “LS Data Capture”. Start the capture mode under “Capture” and “Activate”.
The program should now show “Ready to receive data …” as well as “WAITING FOR
SESSION” or “WAITING FOR NEXT SESSION” (see screenshot). The program must
now remain untouched until all measurements are finished. The data are automatically
transferred after each measurement and saved as a BSF file (.BSF, including all details)
and as a text file (.ASC, only counting rates).
Please take care to not use any other software as an internet explorer because they often
cause the failure of the data receiving.
Figure 5. Receiving data at the computer. Screenshot during activated receiving mode. The program must run
in this mode during all measurements.
Data are transferred after each measurement automatically.
4.
The files are saved in a folder on the desktop. All spectra are saved in the subfolder
“USER00”. The file names are given according to the following scheme:
5.
A collection of all overall counting rates of each rack are saved to the subfolder
“SNGLRACK” after the whole rack is completely measured.
Important: Please note the sample number “NN” for each sample that you measure
(“NN” used for saving each spectrum in the subfolder USER00) as well as each rack
number (different (!) “NN” used in the subfolder SNGLRACK) to be able to know
which file corresponds to which measurement.
6.
11
12
Evaluation
Part I:
1.
2.
Calculate the activities of the standard cocktails of 3H and 14C for today.
Calculate the percentage of the response probability η of the liquid scintillation counter for
both nuclides 3H and 14C (η = RN ⋅ 100 ⋅ A–1).
Part II:
3.
4.
6.
7.
8.
Plotting of the β–-spectra in Excel
Plot the normalized spectra of 3H and 14C in one Excel plot and choose suitable window
settings.
Discuss your choice with your assistant.
Part III:
5.
Response probabilities
Determination of the molar account of 3H and 14C of an unkonwn sample
Calculate the percentage of the net counting rate (subtracted background) for 3H and 14C in
the windows A and B.
Determine starting from the percentage of the net counting rate the theoretical counting
rate and, herewith, the activity of 14C in the unknown sample.
Calculate the overall counting rate of 14C and 3H in the unknown sample (see chapter 2.7).
Calculate the molar amount of 3H and 14C in the unknown sample.
Part IV:
Validation of the results and comparison with the automatic measurement
9. Compare the manually calculated results with the automatic measurement of 3H and 14C.
10. Add the normalized spectrum of the unknown sample to your Excel plot of 3H and 14C.
11. Are there any differences? If yes, how could they possibly be explained? Can you think of
a method to improve your results?
12
13
Data Sheet 1: Sample Overview
3 AUTOISOTOP
1 PRAKTIKUM
4 SPEKTRUM
channel _____ – _____
window B
channel _____ – _____
sample
window
LL – UL
t
N
–1
[min ] [min]
1
background 1
0 – 1000
1
2
3
H standard
0 – 1000
1
3
14
C standard
0 – 1000
1
4
unkn. sample
0 – 1000
1
A
1
1
background 1
B
1
3
A
1
2
B
1
A
1
B
1
A
1
B
1
spectrum ID*
window A
position
rack ID*
program
Channel interval for the windows:
2 σR
[min–1]
H standard
3
14
4
unkn. sample
C standard
1
background 1
2
empty
3
3
4
5
RN
[min–1]
auto
auto
1
H standard
3
auto
1
14
14
C
auto
1
3
H
auto
1
C standard
unkn. sample
H
14
C
auto
1
*The first file of the day will be numbered 01 and then each file with increasing number. This
is important for the identification of the correct file at the computer. Please note that your first
measurement can have a different number than 01 if you are not the first group of the day.
13
14
Data Sheet 2: Tables for Analysis
Part I:
Response probabilities
3
14
H
C
calculated activity today
response probability
Part II:
Window settings
window A
window B
% of 3H
signals
RN in the
window
window interval
% of 14C
signals
–
–
Calculation of the activities and the molar amount
100% counting rate [min–1]
activity [min–1]
molar amount [mol]
14
unknown sample
unknown sample
automatically
3
3
H
14
C
H
14
C
Chapter 6
Backscattering of β--radiation
Task
1. Determine the energy of the β-emitter 204Tl by measuring the net counting rate RN as a function of the weight per unit area of an aluminum absorber by determining the maximum range
of the radiation and by using the graph: “β-energy as a function of the maximal range
[mg⋅cm-2]”!
2. Determine the "atomic numbers" of plexiglass and brass from the dependence of the net
counting rate RN on the atomic number of different backscattering materials for 204Tl!
Fundamentals
Interactions of β-particles with matter are mainly dominated by two types of processes:
1. Ionisation and activation processes
2. Scattering processes at atomic nuclei and orbital electrons
1. Ionisationen
For the relationship of the loss of energy of β-radiation through scattering process to the loss of
energy through ionization, the following equation roughly applies:
∆E bremsstrahlung
∆E ionization
= E ⋅ Z / 800
(1)
(E = energy in MeV; Z = atomic number)
2. Scattering processes
2.1. Scattering processes at electrons
Inelastic scattering is observed at the orbital electrons. The probability that an electron is scattered with a distinct angle δ is proportional to the number of electrons per volume of the scattering material and, therefore, proportional to the atomic number Z.
The scattering of β--particles at orbital electrons is very rare and of less importance compared
with the scattering at atomic nuclei.
2.2. Scattering processes at atomic nuclei
Coulomb scattering is observed at the atomic nucleus. It occurs most frequently with highenergy β-radiation. In the electrical field of the atomic nucleus the high-energy electrons lose
energy and send out a continuous spectrum of X-rays (bremsstrahlung). The loss of energy is
proportional to the atomic number of the atomic nucleus and to the energy of the β-radiation.
The probability that an electron is scattered with a distinct angle δ is directly proportional to
the thickness of the absorber surface, to the number of scattering centers, as well as directly
1
proportional to the square of the atomic number of the absorber substance, but indirectly proportional to the square of the energy of the β-radiation.
This probability is described by the following formula:
n ( δ) F ⋅ N ⋅ d
e4
(Z − A F )2
=
⋅
⋅
p ( δ) =
n0
4 ⋅ r 2 (m e v e2 ) 2 sin 4 ( δ2 )
n(δ)
n0
N
F
d
F⋅d
4⋅r2
e
meve2
Z
AF
sin4(δ/2)
(2)
number of electrons in a distance „r“. The angle „δ“ is defined as the angle between the directions of the arriving and the scattered electrons
number of scattered electrons at the surface
number of atomic nuclei per volume of the scattering substance (in practice corresponding to the density of the material).
area of the scattering surface
"thickness" of the scattering surface
scattering volume; the product „F⋅N⋅d“ represents the total number of scattering
centers
distance and the geometry of the detector to the scattering surface
elementary charge
kinetic energy of the β-particle
atomic number of the scattering material
atomic form factor, describing the geometry of the atomic nucleus: when a βparticle hits a nucleus, the nucleus deviates from its spherical form
describes the angle dependence of the scattering of the β-particles
The values which determine the scattering angle δ of β-particles favor small angles. A change
of the direction of a β-particle by about 180° due to unique coulomb scattering at an atomic
nucleus is very rare. It is almost exclusively caused by multiple scatterings with smaller scattering angles δ, as is shown in figure 1.
Fig. 1: (a) max. range of an electron in a substance; (b) backscattering of an electron due to
repeated Rutherford scattering (Scattering at atomic nuclei)
Backscattering is observed when a β-particle leaves the substance into which it has penetrated
with an overall scattering angle of about 180°. The backscattering increases the measuring rate
as a function of the material causing the scattering, that is, the material behind the sample.
2
The backscattering factor ϕR is defined as:
ϕR =
R ( R)
Intensity of the backscattering
= N
Primary intensity of the radiation
RN
(3)
R (NR ) = Counting rate with backscattering; RN = Counting rate without backscattering
Fig. 2: Dependence of the backscattering factor ϕR on the area weight and the atomic number
of the absorbers and on the energy of the β-radiation
Figure 2 shows that the backscattering factor increases at first with increasing thickness of the
scattering substance and reaches a maximum value ϕSR when a certain saturation thickness has
been reached. It can be assumed that the β-particles, after they have entered the material, interact with the atomic nuclei on their way out of the material. The further the β-particles have
penetrated the material, the smaller the chance that they leave it again. This saturation thickness
(dS [mg⋅cm-2]) depends solely on the energy of the β-radiation. This thickness could be determined theoretically; practically it corresponds to 1/5 of the maximum range rmax of the β-particles, which penetrated into the lower layers of the material.
3
3. Application of backscattering / Absorption of radiation
Attenuation of β-radiation:
N = N0 · e–µ·d
N = number of not-scattered electrons behind the scattering shield
N0 = number of electrons before the scattering shield
µ = linear attenuation coefficient [cm–1]
d = thickness of the scattering shield
The absorption and scattering of radiation of radionuclides is often used in industry for thickness measurements or material testing (see Figure 3). This method can also be used for determination of the thickness of metal coatings or of liquid filling levels.
Fig. 3: Thickness measurements by use of radioactive radiation
Experimental
1. Precautions
Work carefully and pay particular attention to avoid incorporation. Avoid contamination of the
the measuring instruments and chamber.
2. Experimental set-up
Insert a plexiglass slider into the lowest position of the counting chamber (to avoid
backscattering from the chamber material).
Insert a second slider with holders for the Al
backscatterers, and finally prepare a third
slider to carry the radioactive substance. In
this setup only the β-particles that are
backscattered through the hole in the aluminum carrier are measured (Figure 4).
Fig. 4: Experimental setup
4
3. Procedure
1.
2.
3.
4.
5.
6.
Determine the background radiation in the chamber with all sliders but without any substances (for checking later the chamber for contamination). Gate time: 5 min.
Use tweezers while transferring the substance from the desiccator and insert it into the hole
slider with the radioactive side down. The radioactive nuclide 204Tl is located in the notch
of the aluminum carrier.
Put the aluminum plug into the hole and determine the background again. Here, the bremsstrahlung and the γ-radiation are considered. Gate time: 5 min.
Remove the aluminum plug. Measure the aluminum absorbers of different weights per unit
area. The appropriate weight per unit area specification can be found inside the cover of
the box. Gate time: 60 sec for each measurement.
Note: Keep the distance from the absorber surface to the substance and to the window of
the Geiger-Mueller counter constant (that means, the surfaces must "grow downwards").
For this reason the numbers of the aluminum absorbers must be put upside down into the
hole slider.
Ask the assistant how many aluminum absorbers should be used for the measurements (up
to a maximum weight per unit area of 14000 g⋅cm-2).
Keep the hole slider which contains the 204Tl material in the highest position.
a) Install the hole slider with the backscattering disks in the second position from the top.
b) Leave the plexiglass slider in the lowest position in the counting chamber. Gate time: 60
sec for each measurement.
The thickness of the samples of each material is larger than the saturation thickness for
the β-radiation in the given material. The following materials are used:
Material:
Z:
7.
C
6
Al
13
Fe
26
Cu
29
Ag
47
Sn
50
Pb
82
Plexiglas
?
Brass
?
Remove the 204Tl containing substance from the counting chamber. Determine the background in accordance with 1.
Evaluation
1.
Use the Program Excel for your evaluation, draw the net counting rates (experimental
counting rates minus background with aluminum plug in accordance with 3) as a function
of the weight per unit area of the backscatter (with errors). Determine the thickness of the
saturation from the curves.
Fig. 5: Plot of the counting rates as a function of the weight per unit area of one absorber
material (d: Weight per unit area; RN: Net counting rate; (*): Maximum backscattering density,
saturation region)
5
Select two scalings for the axis: 0 – 700 g⋅m-2 and 0 – 7000 g⋅m-2 (see above). Determine the
β-energy of 204Tl with the help of the determined saturation thickness (consider the experimental error) and figure 6.
Fig. 6: β-energy as a function of the maximum range [mg·cm–2]
Note the different units used in the two diagrams and the relation: rmax = 5⋅dmax
2.
Use the Program Excel for your evaluation,
draw the net counting rates (experimental
counting rates minus background with aluminum plug in accordance with 3) as a function
of the atomic number of the used backscatterer
with errors. Determine the “atomic number“ of
plexiglass and brass with the help of the curve.
Fig. 7: Dependency of the
backscattering as a function of
the atomic number.
6
Data sheet 1: Backscattering of β–-radiation
Geiger-Müller Counter
Background (including bremsstrahlung)
Nuclide: 204 Tl
Weight per
unit area of the
absorber
[g⋅m–2]
7
N
R01 =
R02 =
t
[min]
RM
[min–1]
[min–1]
[min–1]
RN
[min–1]
2 σR
[min–1]
Data sheet 2: Backscattering of β–-radiation
Geiger-Müller Counter
Background (including bremsstrahlung)
Nuclide: 204 Tl
8
Backscattering
material
Atomic
number
C
Al
Fe
Cu
Ag
Sn
Pb
Plexiglass
Brass
6
13
26
29
47
50
82
?
?
N
R01 =
R02 =
t
[min]
RM
[min–1]
[min–1]
[min–1]
RN
[min–1]
2 σR
[min–1]
Chapter 7
Self Absorption of β-Radiation
Task
1. Prepare eight CaCO3 samples with different thicknesses.
2. Determine the saturation layer thickness of CaCO3 for the β-radiation of 45Ca. Do this by
determining the net counting rate as a function of the area weight of the precipitate.
Fundamentals
Ionizing radiation can be partly absorbed by the radioactive substance itself. This part of the
absorption is called self-absorption. Self-absorption depends on the thickness of the substance
and on the type and the energy of the radiation. In the case of γ-radiation self-absorption can be
neglected because of the strong penetrating power of this radiation. The extent of self-absorption becomes important with α-emitters and with low-energy β-emitters.
One way to eliminate error due to self absorption is to work with samples that are so thick that
the radiation from the bottom layer of the sample is completely absorbed and never leaves the
sample. This saturation layer thickness must be determined experimentally. The second possibility is to correct the measurement rates to a thickness of zero.
Fig. 1: Saturation layer thickness
Experimental
1. Precautions
There is no danger due to radiation fields because of the low energy of the β-radiation of 45Ca
(Eβ = 0.3 MeV; t½ = 165 d). But be careful with incorporation: Danger exists when measuring
the dry substances (formation of radioactive dust)! The samples may only be transported in the
desiccator.
1
2. Equipment
1 Hahn’s suction filter with witt’s suction bottle, glass cylinder, specialized filter paper, 1
beaker 400 ml, 1 beaker 800 ml, 8 beakers 50 ml, 1 glass rod, 2 tweezers, 8 aluminum holders
for the filter cakes, 2 piston pipettes (2-5 ml) 1x active and 1x inactive, 1 waste bottle for liquid
45
Ca waste, 1 waste bucket for solid 45Ca waste, 1 Geiger Mueller counter in a measurement
chamber, 1 computer
3. Chemicals
CaCl2 solution I (0.25 mol/l; A = 33 300 Bq⋅ml–1)
CaCl2 solution II (0.025 mol/l; A = 3 330 Bq⋅ml–1)
Aqueous ammonia solution (6 mol/l)
Na2CO3 solution (1 mol/l)
4. Experimental procedure
Preparation of the precipitates
1. Install the suction bottle in the working area for radioactive materials!
2. Prepare the CaCl2 solution II: 5 ml of the solution I is given into a 50 ml measuring flasks
and is filled up to the calibration line with distilled water!
3. Prepare the CaCO3 precipitates!
a) The non-radioactive solutions are added to the beakers no. 1 to no. 8 according to
the amounts in table.
No.
Beaker
[ml]
CaCl2 solution I
[ml]
CaCl2 solution II
[ml]
Distilled
water
[ml]
NH3
[ml]
Na2CO3
[ml]
Expected
amount of
CaCO3 [mg]
1
2
3
4
5
6
7
8
50
50
50
50
50
50
50
50
1
1
2
3
4
1
4
6
5
-
5
5
10
10
5
10
10
25
1
1
1
1
2
2
2
2
1
1
1
1
2
2
2
2
2.5
10
15
25
37.5
50
75
100
b) A membrane filter is put on the Hahn suction filter and the equipment is checked for
tightness! For this the glass cylinder is filled up to ⅔ of its overall height with distilled
water. If no liquid runs between the glass cylinder and the frit, the water is sucked
off in a way that a small quantity of liquid remains on the filter paper.
4. Filtrate and wash the precipitates:
Suction filtration is performed with the precipitates in beakers no. 1 to no. 8. Be
careful: Whirl up the precipitates with a glass rod first and then transfer the solids
2
into the glass cylinder of the suction filter with the use of a glass rod to avoid splashing of the radioactive mixture. Please take care that the glass cylinder is never sucked
empty (regulation of the vacuum in the Witt’s pot). The solids from the beakers must
be transfered quantitatively. The precipitates are washed twice with water and dried
with a stream of air. Please avoid a cracking of the solid layers. The sucking device
is checked each time for tightness! Additionally, make sure that the 400 ml beaker in
the Witt’s pot does not overflow. Empty the liquid into the 800 ml waste container.
5. Prepare the samples!
The glass cylinder is carefully removed with a slight vacuum, so that the filter paper
remains sticking to the frit. It is then carefully removed with tweezers, whereby the
filter cake should not be damaged. Some drops of a glue-acetone mixture are distributed into a small aluminum plate, which carries the number of the filter cake on the
reverse side. The filter paper is put into the center of the small plate. Subsequently,
the substance is put into a desiccator for drying over night.
Decontamination
All contaminated devices should be carefully decontaminated subsequently with half concentrated HCl and distilled water. The radioactive waste solutions are collected in the 800 ml
beaker. Special attention should be paid to the frit. The glass containers must be checked with
a detector after the decontamination by a wipe test, since the weak β-radiation of the 45Ca does
not penetrate the glass walls. After successful decontamination, the contents of the 800 ml
beaker are emptied into the bottle with the radioactive 45Ca waste. Liquids for the decontamination should be used as sparingly as possible – however, use as much as necessary. Operate
economically! The treatment of radioactive waste solutions is very expensive.
Measurement
Determine the background: Gate time 5 min
Each substance is measured for 60 sec. in a slider with a high ring in position 4 (from the top)
in the measurement chamber. Subsequently, the background is determined again (5 min); check
the measurement chamber, the slider as well as the ring for contamination. After measurement,
put the samples in the active waste bottle for solid 45Ca.
Evaluation
The inner diameter of the glass cylinder is 1.6 cm, the surface is 2.0 cm2 according to the formula: Area = π⋅r2.
Using the amounts of the CaCO3 precipates, summarized in the table, the weights per unit area
as well as the thickness of the solid layers are calculated for each sample.
ρ CaCO3 = 2.93 g ⋅ cm −3
Use the program Excel for the evaluation, the net counting rates as a function of the weight per
unit area. Then determine the thickness of the saturation layer for 45CaCO3 precipitates.
3
Data sheet: Self-absorption of β-radiation
Geiger Müller Counter
Nuclide: 45Ca
No.
N
t
[min]
RM
[min–1]
1
2
3
4
5
6
7
8
Saturation layer thickness [cm]:
4
RN
[min–1]
R0 =
[min–1]
2σR
[min–1]
Thickness of
the layer
[mg⋅cm–2]
[cm]
Weight
per unit
area
saturation
layer
Chapter 8
The “Thorium Generator” – Determination of the
half-lives of different members of the natural thorium
decay series
Task
1. Separate the daughter nuclides 212Pb, 212Bi, 208Tl and 208Pb from the naturally occurring
isotope mixture of the 232Th decay series.
2. Perform an energy calibration of the NaI scintillation counter by using 137Cs
3. Separate the nuclides 212Pb, 212Bi and 208Tl by chemical methods.
4. Determine the half-life of 212Bi indirectly by means of the decay curve of 208Tl.
Fundamentals
1. The thorium decay series
In nature, the element thorium is found almost exclusively as isotope 232Th. This radioactive
isotope is the first member of the 4n-decay series:
232
10
−
−
a)
,7 a )
,13 h)
→228 Ra β(5
→228 Ac β( 6
→228 Th
Th α(10
In geological periods, the short-lived daughter 228Th is formed from the first member, 232Th, of
the thorium decay series as shown in the scheme above.
228
(1,9 a)
( 3, 64 d )
Th α

→224 Ra α
→220 Rn
As 228Th continues to decay, 220Rn is formed following two α-transformations. This radioactive
noble gas is constantly released from soils containing thorium. Radon, in turn, decays to 212Bi
as shown in the following scheme:
220
−
h)
( 0.15 s )
( 55.6 s )
Rn α
.6
→212 Pb β(10
→212 Bi
→216 Po α
Decay of 212Bi proceeds in two different paths. The first possibility (36.2 %) occurs by α-transformation to produce 208Tl; the second (63.8 %) results in 212Po following a β--transformation.
Ultimately, both of the paths result in the formation of 208Pb.
1
−
212
Bi α(60.6
 min)
→ 208 Tl β(3.1min)
→ 208 Pb
212
min)
μs)
Bi β(9
→ 212 Po α(0.3


→ 208 Pb
−
2. The radiochemical equilbrium
The so-called “radiochemical equilibrium” should not be compared to thermodynamic or kinetic equilibriums.
The “radiochemical equilibrium” is not reversible i.e. it cannot be attained from both side of
the equation. Furthermore, it can never be considered as a stationary state. A “radiochemical
equilibrium” is reached when the speeds of decay of the mother and daughter nuclides are
matching thus leading to:
NM ⋅ λM = ND ⋅ λD
Which translates to: AD = AM
(1)
(2)
The activity of the mother and daughter nuclides equalize when the “radiochemical equilibrium” is reached.
Three cases can be described:
The secular equilibrium: T1/2 (Mother) >> T1/2(Daughter):
The activity of the daughter nuclide approaches asymptotically the activity of the mother nuclide.
The transient equilibrium T1/2 (Mother) > T1/2(Daughter):
The activity of the daughter nuclide decreases along the activity of the mother nuclide.
The absent equilibrium
T1/2 (Mother) ≤ T1/2(Daugther):
The daughter´s activity continuously increases. An equilibrium cannot be attained.
Die Zeit tGGW, die für die Einstellung des „Radiochemischen Gleichgewichts“ benötigt wird,
kann nach Gleichung 3 ermittelt werden. Beim säkularen Gleichgewicht kann λMutter vernachlässigt werden.
The time tEq required for the establishment of the “radiochemical equilibrium” can be determined using equation (3). λMother can be ignored in the case of a secular equilibrium.
tEq = (𝜆𝜆
ln 0,01
𝐷𝐷𝐷𝐷𝐷𝐷𝐷𝐷ℎ𝑒𝑒𝑒𝑒 −𝜆𝜆𝑀𝑀𝑀𝑀𝑀𝑀ℎ𝑒𝑒𝑒𝑒 )
(3)
Systems with long-lived mother nuclides and comparatively short-lived daughter nuclides are
often used as radionuclide generators for medical or technical applications.
The mother nuclide is brought to chemical form suited for a separation from the daughter nuclide. The latter can then be separated or “milked”. Such setups are also referred as “Cow”.
Those systems allow the production of radioisotopes without a fission reactor. The half-life of
the mother isotope determines the useful life of such a generator.
2
3. The thorium cow
Since 220Rn is a gas and is easily separated from its “parents”, its decay products can be collected in a carrier-free form. 216Po daughter nuclei, which develop from the α-decay of 220Rn,
as well as 212Pb daughter nuclei, which are subsequently formed from 216Po, are affected by a
strong recoil and are positively ionised. Therefore, these daughter nuclei can be deposited on a
negatively charged tantalum metal sheet (cathode).
A device which makes such separation possible is known as a “thorium generator” or “thorium
cow.” This device consists of a sealed cylindrical container with the approximate height of
10 cm, in which a 228Th-containing substance is found. Also, within the device is a tantalum
metal plate with a negative potential gradient relative to the chamber wall. The 228Th activity
of the substance used in this trial amounts to approximately 37 kBq. Once the thorium generator
is closed, a radiochemical equilibrium will be established between 228Th and 220Rn and their
decay products.
228
Th: t½ = 1,9 a or. t½ = 5,99 ⋅ 107 s, λ1 = 1,16 ⋅ 10-8 s-1
220
Rn: t½ = 55,6 s, λ2 = 1,25 ⋅ 10-2 s-1
The separation of the daughter nuclides onto the tantalum cathode occurs when a voltage is
applied for 24 hours to the device. To remove the plate, the voltage is turned off and the reaction
vessel is shortly opened.
4. Separation of 212Bi from a mixture of 212Bi, 212Pb, 208Tl and 208Pb.
The daughter nuclides are removed from the tantalum plates using concentrated nitric acid thus
oxidizing the metallic deposits according to equation (4) to cationic species:
3 Pb + 8 HNO3 → 3 Pb(NO3 )2 + 4 H2 O + 2NO
(4a)
3 Tl + 4 HNO3 → 3 TlNO3 + 2 H2 O + NO
(4b)
Bi + 4 HNO3 → Bi(NO3 )3 + 2 H2 O + NO
(4c)
The separation of the dissolved nuclides is possible thanks to the different basicity of the metal
hydroxides. The hydroxides of the daughter nuclides are formed are at different pH values and
precipitate from the aqueous solution (pH: Bi(OH)3<Pb(OH)2<Tl(OH)). EDTA is used to form
a soluble lead complex to a avoid a co-precipitation with the other elements. Equation (5) shows
the formation of this complex.
3
O
2O O
N
Pb2+
+
EDTA4-
O
Pb
O
N
O
O
O
(5)
Because lead is masked as EDTA-Lead complex and remains in solution and thallium only
precipitates as a hydroxide not before very high pH values are reached, it is possible to selectively isolate bismuth at a pH of 7 according to equation (6).
Bi(NO3 )3 + 3 NaOH → Bi(OH)3 ∙ + 3 NaNO3
(6)
The formed precipitate is filtered and dissolved in hydrochloric acid leading to chlorido complex of bismuth(III) e.g. [BiCl4]-, [Bi2Cl10]2- or [BiCl5]2-.
5. Indirect determination of the half-life of 212Bi with interval separation from 208Tl
208
Tl is rapidly formed by the decay of 212Bi in the solution obtained above.
The half-life of 212Bi is determined indirectly by separating its daughter nuclide 208Tl at known
intervals. The bismuth solution is therefore loaded on a strongly basic anion exchange column.
The negatively charged bismuth chlorido complexes are retained by the resin while the cationic
thallium can be eluted with diluted hydrochloric acid (0.5 M).
The decay curve of the obtained 208Tl can be measured precisely and then back extrapolated to
the time of the elution. A radiochemical equilibrium was established at the moment of the elution meaning that the decay constant of bismuth and thallium were equal.
By repeating the separation at given time intervals- separated at least 30 min. or 10 half-lives
of 208Tl to re-establish a radiochemical equilibrium- a series of extrapolated activity at elution
is obtained. The decay curve of 212Bi is obtained by linking those points.
6. Carrier free conditions
Carrier free conditions are met when a radionuclide is isotopically pure meaning that is not
diluted by stable isotope of the same element.
The newly formed isotopes are separated under carrier free conditions during the first part of
the “Thorium-cow” experiment. 220Rn leaves the 228Th sample and its daughter isotopes are
deposited on the cathode. The quantities of daughter nuclides obtained are extremely limited.
As such it would be pointless to attempt to precipitate those elements as their solubility products
could never be reached. The addition of a non-radioactive “carrier isotope” is thus required for
the subsequent work-up. The added compounds should ideally exist in the same chemical form
as the radionuclides. Lead acetate is used as a “lead carrier” and bismuth nitrate as a “bismuth
carrier”. The solubility product of bismuth hydroxide can only be reached by adding bismuth
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nitrate thus allowing the isolation of the radioactive bismuth. On the other hand, the addition of
non-radioactive lead allows the radioactive lead to remain in solution. The radioactive leadwhich is extremely diluted might co-precipitate with bismuth without such a “retention carrier”.
Experimental part
1. Precaution measures
The tantalum metal sheets are located in a glass beaker covered with cellulose and are brought
to the workspace by an assistant. All isotopes of the decay series prior to 212Pb are almost
completely decayed prior to the start of the experiment.
Please follow the usual precautions while working with radioactive materials to avoid incorporation and contamination.
2. Chemicals
The following solutions are needed for this experiment:
HNO3 conc.
0.1 M EDTA
6.0 M NaOH
2.0 M NaOH
2.0 M HCl
0.5 M HCl
0.59 M Bi(NO3)3
0.59 M Pb(CH3COO)2
Thymolphthalein indicator
Dissolution
Titriplex-III for the complexation of lead
Hydroxide precipitation
Washing solution
Dissolution of the hydroxide
Eluant
Bismuth carrier
Lead carrier
Acidic: colorless, basic: blue
These solutions have to be prepared:
Glass I and Glass II for the dissolution of activity
1.5 ml HNO3 conc.
1.5 ml Bi(NO3)3 solution
1 ml Pb(CH3COO)2 solution
Glass A:
Glass B:
Glass C:
Glass D:
Glass E:
Glass F:
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5 ml Titriplex for the complexation of lead
Thymolphthalein indicator (3 drops)
8 ml 6.0 M NaOH for the hydroxide precipitation
8 ml Titriplex + 2 ml 2.0 M NaOH as washing solution
5 ml 2.0 M HCl for the dissolution of the hydroxide
5 ml distilled water to wash the filter
10 ml 0.5 M HCl as eluant (3x, since there are three elutions)
3. Procedure
3.1 Separation of the isotopes of the Th decay chain the Th cow.
A thin 228Th sample is deposited on the bottom of cylindrical pot which is tightly sealed to
prevent the diffusion of 220Rn. Several tantalum plates are attached on a metal hook. A cathodic
current then flows for 24h though the plates.
A “radiochemical equilibrium” establishes itself between the decay products of the Th decay
chain. Radon diffuses continuously from the Th probe. Its daughter elements deposit on the
cathode as they are ionized by a strong recoil of their alpha decay.
The nuclides listed below accumulate on the electrode
212Pb
t½ = 10.6 h
β-: 0.35 MeV
γ: 238.9 keV (47%)
212Bi
t½ = 60.6 min
α: 6.05 MeV
β-: 2.27 MeV
γ: 727 keV (7.1%)
212Po
t½ = 0.3 µs
α: 8.795 MeV
γ: 2610 keV (2.6%)
208Tl
t½ = 3.1 min
β-: 1.79 MeV
γ1: 583 keV (86%)
γ2: 2615 keV (100%)
208Pb
stable
3.2. Setting of the measurement device
The measurement is performed using a sodium iodide scintillator. 208Tl has two gamma lines at
583 keV and 2615 keV. A multi-channel analyzer is used for the purpose of this experiment.
An energy calibration of the measuring device muss be performed prior use. The calibration is
made as such to allow the measurement of the gamma line at 583 keV. The decay lines are
measured without resolving the energy. The measurement of the decay curves using a finite
time window is thus possible.
Please prepare the detector for the measurement. Open the program ProSpect at the computer.
Connect the multichannel analyzer and the computer, please choose the menu point “Connect
to Device” and choose the detector “Thoriumkuh”.
Optimize the high voltage (maximum 700 V) and the gain of the detector in a way that the
Cs-Photopeak (E = 662 kV) at around 1/2 of the energy scale appears.
Perform an energy calibration, choose the acquisition mode PHA (also see experiment gamma
spectrometry).
137
Change the acquisition mode to MCS and use the following settings:
Acquisition Mode: MCS
MCS Conv Gain: 256
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Dwell: measuring time
Disc Mode: ROI
Start Channel: 400
End Channel: 800
Measure the background for 5 min. The measurements last 1min each.
For the evaluation, you need to stop the measurement and save the data. You can save the data
with de Butten “EXPORT DATA”. You receive a file with the ending *.csv where the data are
separated by comma. This file can be opened by a program such as Excel. Use the option “characters like commas separating fields” while importing the data.
3.3. Preparation of the chemical separation
Working with radioactive compounds, each step should be carefully planned and required items
should be prepared sensibly. All inactive substances required for the procedure are to be prepared beforehand. When working with short-living nuclides like 208Tl, all needed items must
be readily available.
Make sure that the separation columns are ready to use (ask an assistant). Turn on the water
bath and prepare the suction device (tightness test).
3.4. Production of the radioactive parent solution
The tantalum plates are brought by the assistant to experimental workspace. To separate the
formed radioactive thorium, the metal sheet is placed in nitric acid (Glass I) and warmed for 5
minutes in a water bath. Beaker number I can then be removed from the water bath and placed
in the plastic working chamber. Next, the metal sheet is transferred with tweezers to beaker
number II and heated for additional 5 minutes in the water bath. Subsequently, the metal sheet
is removed from solution II, rinsed with a few drops of water and placed back in the transport
container. The two solutions are then pooled into beaker I.
3.5. Separation of bismuth
Bismuth is first separated from the radioactive solution, which contains 212Pb, 212Bi, 208Tl and
Pb. This separation is successful due to the different basicities of the hydroxides of these
metals, which decreases in the order Tl > Pb > Bi. Bismuth hydroxide precipitates at pH 7,
whereas thallium remains in solution. In order to completely avoid the co-precipitation of lead,
EDTA is used as a lead-complexing agent.
208
For this separation, solution A is added to beaker I, in order to mask the lead which is contained
in the solution. Subsequently, a small amount of indicator is added to beaker I and the prepared
NaOH (solution B) is then also added until the solution is getting blue. A white precipitate of
Bi(OH)3 begins to form and the mixture is again placed in the water bath for 3 minutes. After
cooling, the precipitate is filtered with a membrane filter and washed several times with solution
C. After the solution from the last washing step is removed by sucking air through the filter,
the Bi(OH)3 precipitate is dissolved in HCl. In doing so, the filter paper is carefully transferred
using tweezers into beaker D, which is then placed in the warm water bath for a short period of
time. Subsequently, this solution is then diluted with 5 ml distilled water (beaker E) by rinsing
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off the filter paper held above Glass D. The filter paper is then wrapped in cellulose and placed
in the radioactive thorium generator waste.
3.6 Separation of bismuth and thallium
As a result of radioactive decay, 208Tl is quickly reformed in the freshly produced 212Bi solution.
In order to separate 208Tl from 212Bi, Cl- loaded anion exchange resin (Dowex) is employed.
Using a glass rod, the HCl-bismuth solution is carefully loaded to the column. This solution
contains bismuth as tetrachloro complexes [BiCl4]-, which bind in place of the chloride ions on
the anion exchange resin. The solution is allowed to run slowly through the column until the
meniscus of the liquid reaches the surface of the exchanger material. In order to avoid separation
failure, the column should not be too long. The work is performed carrier-free, therefore, only
traces of thallium are present.
Once the column is loaded with bismuth, time is required to establish the “radiochemical equilibrium”. 30 minutes after the column is loaded, it can be assumed that the “radiochemical
equilibrium” between 212Bi and 208Tl has been established and the separation of 208Tl can begin.
This waiting period can be used to prepare the counting device. For the separation of thallium,
10 ml of 0.5 M HCl (beaker F) are added to the column. This solution is allowed to run quickly
though the column until its meniscus reaches the surface of the resin. The stopwatch is started
once the stopcock is turned and runs throughout the experiment (≈ 1.5 h). The eluate is collected
in a suitable container and measured immediately. Work should be performed as efficiently as
possible.
3.7. Measurement of the decay curve of 208Tl
Choose a measurement time of 1 min before you start the first separation. After the separation
put the elution beaker with the solution on the detector, start the measurement at a full or half
minute. You need the exact measurement time for the evaluation. Please write down the starting
time. Stop the measurement 27 min after the start of the elution, save the data-file. The last
value is used as background for the measurement just done.
Clean the elution beaker for the second elution that should start exactly 30 min after the first
one. All in all, three elution’s are carried out with 10 ml of 0.5 ml of HCl.
Evaluation
The evaluation is performed using a spreadsheet software e.g. MS Excel. The net count rates of
Tl are logarithmically plotted relative to the time elapsed after each chromatographic separation from 212Bi. The logarithmic values can be calculated prior plotting, or an automatic scaling of the Y axis can be selected in the spreadsheet software. Three linear regressions of the
decay of 208Tl (in the form y=mx+n)in total can be obtained. The half-life can be obtained from
each slope m. The theoretical values at the time of elution of the count rates are then obtained
by extrapolating the slopes. The three obtained linear equations are solved according to their
respective time of elution (x value of the elution 30/60/90 min). The obtained y values a.k.a. as
head points are proportional to both the activity of 208Tl and 212Bi.
208
The three obtained pairs of values (count rates at a specific elution time) are plotted logarithmically-as done with thallium previously. A linear regression of those three points leads to the
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linear decay slope of 212Bi. As the activity of the bismuth declined throughout the experiment,
it can be calculated from the obtained decay slope as done previously for the thallium.
Measurement results:
Half-life 208Tl from the 1st separation:
Half-life 208Tl from the 2st separation:
Half-life 208Tl from the 3st separation:
Average half-life 208Tl:
Half-life 212Bi:
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