Download Development of Constancy Control and Calibration Protocols for

Master Thesis
RF9300, 30 HP
VT 08
Development of Constancy Control and
Calibration Protocols for Radiation Monitor
Devices and Estimations of Surface Dose Rates
from Radioactive Waste Containers Used at
University of Gothenburg.
Dan Thorelli
Supervisors:
Mats Isaksson and Annhild Larsson
Department of Radiation Physics
University of Gothenburg 2008
Abstract
Radiation monitor detectors are the most important tool available for evaluating and
examining the workplace. It is important to only use the detector for measurements
in situations it is designed for and to have an accurate calibration.
The regulation SSI FS 2000:7 [12] states, in 9 § regarding quality assurance, that a
quality handbook should be available. In dealing with extensive laboratory work,
defined in 2 §, the handbook should contain routines for calibration and constancy
control of radiation monitoring devices. This thesis deals with a development of a
simple constancy control routine. As a result a correctly calibrated radiation monitor
or
should not give a reading that is lower than E. E is calculated
that measures
for a few sources of interest.
The maximum amount of radioactivity allowed to be disposed of, depending on the
nuclide, and the maximum surface dose rate allowed for waste containers is given
by the regulations in SSI FS 1983:7 [13]. This thesis deals with a few estimations of
surface dose rates from waste containers filled with gamma or beta emitting
nuclides.
The investigated instruments that measure the ambient dose equivalent could not
be said to give a reading that is lower than the calculated effective dose, taking
statistical factors in mind. This is an important result as it means that they will not
underestimate the effective dose that is related to the risk of the exposure.
Unfortunately none of the used sources are an ideal constancy control source. A
more appropriate constancy and control source could be a Cs-137 source with an
activity of 4 MBq. The effective dose from the new source could be calculated, and
the instruments could be re-measured with the new source by following the general
measurement steps.
Introduction
1
Working with radionuclides
1
Radiation protection
3
Units and concepts
4
Disposal of radionuclides
6
Microshield
7
Radiation Detectors
7
Detector efficiency
12
Energy response
14
Calibration of radiation monitoring instruments
14
Material and methods
16
Point dose estimations method
16
Volume source estimations method
18
General measurement steps
22
Constancy control and calibration method
23
Results
26
Dose levels around waste containers
26
Calibration and constancy control
29
Discussion
32
References
38
Appendix
Nuclides used at Gothenburg University
Volume source definitions in Microshield
Investigated detectors with spread sheet results
2
1. Introduction
When ionizing radiation is used in laboratory work, or other purposes, it is important to
make sure that the workplace is safe and to minimize the risk to the worker, as in any
other field of work. Radiation monitor detectors are the most important tool available
for evaluating and examining the workplace. A single type of detector that could
measure all form of radiation with the same accuracy would be convenient,
however no such radiation detector exists. Instead a number of detectors, all with
different properties, are used to measure the radiation from different forms, at
different intensities and from different nuclides. This makes it important to only use the
detector for measurements in situations it is designed for and to have an accurate
calibration. The regulation SSI FS 2000:7 [12] states, in 9 § regarding quality assurance,
that a quality handbook should be available. In dealing with extensive laboratory
work, defined in 2 §, the handbook should contain routines for calibration and
constancy control of radiation monitoring devices. This thesis deals, in part, with
development of a simple constancy control and calibration method for radiation
monitoring devices, with the main focus set on a simple constancy control. The
method needs to be simple not to be skipped or ignored.
At laboratories or institutions that deal with radioactivity, there will eventually be a
need for disposal of some radioactivity. If the disposal is solid waste and packaged in
a container, a measurement of the surface dose rate needs to be done. The
measurement is made to make sure the dose rate is below the legal limit, before the
container is sent away. The maximum amount of radioactivity allowed to be
disposed of, depending on the nuclide, and the maximum surface dose rate
allowed is given by the regulations in SSI FS 1983:7 [13]. This thesis deals with a few
estimations of surface dose rates from waste containers filled with gamma or beta
emitting nuclides.
2. Working with radiation
2.1 Radiation
An unstable nucleus can, in every moment, decay by emission of radiation with a
certain probability. The emitted radiation is nuclide specific, and can be gamma,
beta or alpha radiation. The probability for decay cannot be influenced in any
physical or chemical form. The decay modes of interest, in this thesis, are gamma
and beta decay. Alpha radiation will not be considered due to the measurement
setup requirements to get a satisfactory measurement.
2.1.1 Gamma radiation (γ)
Photons travelling in a material, with energies of interest in laboratory work, are
subject to an interaction by either photoelectric absorption, compton scattering or
pair production. The sum of the probability for each interaction is included in the
linear attenuation coefficient, µ.
1
If a shied is placed between an emission point and a measuring point , the number
of transmitted photons (N) relative to incoming number of photons ( ) without the
shield, is given by equation (1)
(1)[1]
Where µ is the linear attenuation coefficient, x is the thickness of the material, N is the number of
transmitted photons and
is the number of incoming photons.
Equation (1) show that the photons are subject to an exponential attenuation. Thus a
fully absorbing shield cannot be constructed. The shield has to be designed to
reduce the photon fluence to an acceptable level instead. When half of the number
of incoming photons has interacted in the material, a useful expression in radiation
protection can be derived. Equation (2) gives the half value layer (HVL).
(2)[2]
Where µ is the linear attenuation coefficient
This is the thickness of a material that will reduce the photon fluence by half of its
original value, and is a useful guide when deciding shield dimensions.
2.1.2 Beta radiation (β)
There are two types of beta decay;
-
, electron decay
-
, positron decay
decay occurs in nuclides with an abundance of neutrons and
decay occurs
in nuclides with an abundance of protons. The beta particle share the energy
released (in a decay) with a neutrino, that is created alongside the particle. The
energy distribution between the particles is governed by statistics. There are two
energies of interest for the beta particle;
• The maximum beta energy
A beta particle emanating with maximum energy occurs when it receives all of the
energy released in a decay. The maximum energy can be used to calculate the
particles maximum range. When the range is known it can be used to construct
appropriate shielding. If the dimensions of the shield exceed the maximum range of
the beta particle, in the material, no beta particles will emanate from the surface of
the shield. Equation (3) can be used to estimate the range for beta particles with
maximum energy between 0.01 to 2.5 MeV.
(3)[3]
Where R is the range in mg/
and E is the maximum beta energy in MeV
2
When the beta particle interacts in the shield there is a chance for a bremsstrahung
photon to be produced. The probability, P, for the emission can be approximated by
the expression;
(4)[1]
Where Z is the atomic number of the material and m is the mass of the particle.
Equation (4) show that the probability for emission increases with decreasing particle
mass. This makes the probability for emission higher for beta particles, compared with
other more massive charged particles. Equation (4) also shows that a shield should
be constructed in materials with low atomic numbers, to decrease the probability for
photon contribution.
• The mean beta energy
The mean beta energy is approximately a third of the maximum beta energy [11].
The mean beta energy is used in dose calculations.
2.2 Radiation protection
The recommendations made by the ICRP (International Commission on Radiation
Protection) have a profound influence on radiation protection all over the world.
One important presumption made by the ICRP, is that even small radiation doses
can cause harmful effects. The three main principles of radiation protection are;
•
Justification
To prohibit practices involving additional exposures unless they produce sufficient
societal benefits. The benefits should be weighed against the risks.
• Optimization
The optimization principle requires the radiation exposure, to the worker, to be as low
as reasonably achievable, or ALARA. Implementing ALARA in practice involves;
- Reducing the source
A way of eliminating the radiation source can be done by using ultrasound instead
of diagnostic x-ray wherever possible. Source reduction is also a reduction of the
dose rate, and can be done in several ways. One way is to properly ventilate areas
where airborne radioactivity is present.
- Source containment
To make sure that proper containment, ventilation and filtration is used.
- Time
The minimization of the time that radioactive materials are handled, less time leads
to a lower dose. This can be achieved by practice the procedures without activity
present. The work should be performed quickly, but without rushing.
-
Distance
3
Maximization of the distance from the source. For a gamma point source, the dose is
inversely proportional to the square of the distance. A significant dose reduction can
be achieved with increased distance, for example by using distance tools
- Shielding
Shielding should be used wherever it is necessary to reduce or eliminate the radiation
exposure to the worker. By placing a shield between the source and the worker, the
exposure can be reduced to an acceptable level. The type and thickness of the
material needed to reduce the dose to a safe level varies with the type and amount
of the nuclide.
• Dose limits
Limits of radiation exposure to individuals. SSI issues regulations regarding dose limits
in Sweden, where the most important are
-
SSI FS 1998:4
States that the dose limit for workers, and public, exposed to ionizing radiation
from the workplace is 50 mSv per year and a maximum of 100 mSv for five
consecutive years. The dose limits does not apply for patients subject to
medical radiation treatments, people helping patients during medical
radiation treatments (of their free will), voluntary test-subjects and in
emergency rescue situations.
-
SSI FS 1998:3
Regulations for categorizing the worker in two categories, A or B. If the risk is
not insignificant on an annular basis for: the effective dose to exceed 6 mSv,
the equivalent dose to one eye lens exceeding 45 mSv or the equivalent
dose to the hands, forearm or skin exceeding 150 mSv the worker should be
placed in category A. The effects of accidents that can lead to high
exposures that could justify the placement in category A has also been taken
into mind. Workers that are not placed into category A are placed into
category B
- SSI FS 1998:6
States that a worker classified as category A should be subject to regular
medical examinations and wear a radiation dosimeter. If the result from the
medical examinations is not satisfactory the worker can be limited or
forbidden to work with radiation as a category A worker.
2.3 Units and concepts
2.3.1 Units
4
The absorbed dose (D) is a measure of the energy absorbed per unit mass from
ionizing radiation in a medium. Different types of radiation can cause different
amount of damage. The absorbed dose does not take that into account. To
calculate the total biological effective dose for different types of radiation, the
differences must be considered.
The equivalent dose (H) is given when the absorbed dose is multiplied by a weighting
factor that reflects the radiations ability to cause damage, equation (5). The
weighting factor is dependent on the ionization density, for gamma radiation the
factor is set to unity, and other types of radiation are related to the value according
to their ionization densities.
(5)[4]
Where
is the weighting factor for the radiation type
and
is the absorbed dose in the organ
of
radiation type .
All organs and tissues do not have the same sensitivity when exposed to the
radiation. And in most cases the body is not uniformly irradiated. The effective dose
(E) takes this into account and is given as the equivalent dose multiplied by a organ
weighting factor, equation (6)
(6)[3]
Where
is the weighting factor for different types of organs.
One of the main advantages of using the effective dose is that the risk of a radiation
exposure to one specific organ can be compared with the risk of a whole body
exposure.
Ambient dose equivalent,
[Sv]
The ambient dose equivalent is the dose equivalent in a point in a radiation field that
corresponds to the dose in an expanded and parallel radiation field (see figure (1))
at the depth of 10 mm in the ICRU sphere [5]. Instruments that measure the ambient
dose equivalent should be directional independent.
Figure (1). An expanded and parallel radiation field
Personal dose equivalent,
[Sv]
5
is the dose at a depth d under the placement point of the meter. The readout
from the meter should be directional dependent. The depths most commonly used
are 0.07 mm and 10 mm, to simulate skin or organ exposure.
Relationships between units in radiation protection
Figure (2) show the quotient between the effective dose, (E), and the ambient dose
), for photon radiation in different geometries.
equivalent, (
Figure (2). The quotient between E and
plotted versus photon energy for several different
geometries. [5]. AP: Parallel radiation field, directed as figure (1); ISO: Isotropic radiation field.
Figure (2) show that
is larger than E in many different geometries. Similarly it has
been shown that
is larger than E in most geometries [5]. As a result a correctly
calibrated radiation monitor that measures
or
should not give a reading that is
lower than E.
2.3.2 Concepts
ALI stands for Annual Limit on Intake, and is defined, in ICRP 30 [14], as the annual
intake of a nuclide that would lead to an effective committed dose equivalent
below or equal to 50 mSv and an annual dose equivalent to any organ or tissue
below or equal to 500 mSv. The definition can be expressed as equation (7) and
equation (8)
Sv
Where
is the tissue weighting factor and
(7)
is the total committed dose equivalent in tissue T
Sv, for all T
Where
(8)
is the total committed dose equivalent in tissue T
6
Different values exist for inhalation and ingestion,
is the lowest of these values
for the nuclide of interest.
-
DAC
The ALI value only gives an intake limit for a specific nuclide. No considerations of the
intake rate or the atmospheric/environmental concentrations that leads to the
intake limit are made. For airborne contaminations of radioactivity the Derived Air
Concentration (DAC) takes these effects in mind. The DAC, for any radionuclide, is
that concentration in air (Bq/ ) which, if worked for a year, would result in the
ALI value for inhalation. The DAC can be calculated according to equation (9)
(9)[3]
Where ALI is the intake limit for the nuclide [13] and 2400 is the volume of air a standard person will
inhale during a year at work.
-
Radiotoxicity
For laboratory work with unsealed substances, the regulation SSI FS 2000:7 classifies
nuclides into four categories dependent on their radio toxicity (A to D). Where class
A is highly toxic and class D is the least toxic. The amount of activity that is allowed to
handle during extensive laboratory work is given in [13]
2.4 Disposal of radionuclides
A certain amount of radioactive waste can be disposed of at the local landfill or in
the sewer. The amount depends on the radioactive nuclide, and is defined by the
value. This is regulated by SSI FS 1983:7 [13]. The limitation does not
nuclides
apply to waste from patients subject to treatment or diagnostics with radioactive
substances.
2.4.1 Regulations for solid waste
The maximum amount of activity, per month, allowed to be disposed of at the local
.
landfill is 10
The maximum amount of activity placed in one waste container should not exceed
, and the maximum surface dose rate should not exceed 5 µGy/h. The waste
1
containers should not contain any sealed source with an activity above 50 kBq.
2.4.2 Regulation for liquid waste
The maximum amount of activity that can be disposed of in the sewer, per month, is
or 100
10 ALImin. The maximum activity per disposal should not exceed 1
MBq. After each disposal significant amount of water should be used to flush the
system.
7
2.5 Microshield
Microshield is a radiation shielding and dose assessment program for gamma
emitting nuclides. Microshield is written by Grove Software Inc, situated in the USA.
The version used in this thesis is 6.20.
Microshield is used to estimate the value of the effective dose for the gamma
emitting calibration and constancy control sources at the measurement points.
Microshield is also used to estimate the surface dose rate from the waste containers
filled with gamma emitting nuclides.
3 Radiation detectors
The human body lacks a sense for detecting radiation. Instead the detection is
based on physical and chemical effects produced by radiation in the exposed
material, other than man. Common effects produced by the radiation are ionization
in gas, ionization and excitation in solids, chemical changes and neutron activation.
A detector is usually constructed with one of these effects in mind.
3.1 Gas detectors
When the gas is exposed to ionizing radiation it produces ionization and excitation
effects in the detector gas. This is the principle on which the gas filled detectors are
based. If the detector is exposed to steady state irradiation, the created ion-pairs will
be kept constant. Commonly used gas-filled radiation detectors, like ion chambers,
proportional counters and GM-tubes, then makes use of the direct ionization created
by the radiation.
To be able to collect the ion pairs, an electric field is used. When the electric field is
present, the electrostatic force start to move the particles (electrons and ions) away
from the point of creation. With the applied electric field the moving particles
creates an electric circuit, and a current can be measured. Without the electric field
the net current is zero. The created ions can be lost due to either recombination or
diffusion out of the volume. With increasing electric field, less of the original charge is
lost to recombination effects. At a certain level the electric field is strong enough to
reduce the recombination to a minimal level and all of the created charge is
measured. Further increasing the strength of the electric field, up to a certain
strength, has no effect. This is due to the constant formation the ion pairs and an
efficient collection of the charge created. This plateau is called ion-saturation and is
the area where ion chambers are operated. See figure (3).
8
Figure (3). The pulse amplitude plotted versus the applied voltage of the electric field for two different
energies. [6]
One important application for the ion chambers is to measure the absorbed dose in
different materials. The absorbed dose can be measured using the Bragg-Gray
principle that states that the absorbed dose in a material can be given from the
ionization that is produced in a small gas filled cavity, the ion chamber, in the
material. Several conditions need to be met for the principle to hold. One important
condition is that the cavity, ion chamber, needs to be small compared with the
primary and secondary range of the radiation, to minimize the effect it poses on the
particle flux.
If the detector gas is air and the detector walls are air equivalent, the chamber can
be used to measure the absorbed dose in air. A measurement of the absorbed dose
in air is equivalent with a measurement of the gamma ray exposure, which is another
important application of ion chambers.
If the electric field is further increased, it will cause an effect called gas
multiplication. In the ion saturation region, the electrons and ions simply drift to their
collecting electrodes. On their way the particles frequently collide with neutral gas
molecules. The ions gain little average energy between the collisions due to their low
mobility. The electrons can easily be accelerated in the electric field, due to their low
mass, and have high kinetic energy when colliding with a neutral gas molecule. If
this energy is high enough to cause ionization in the neutral gas molecule, a second
ion pair can be formed. A threshold exists when this second ionization event can
take place, due to electrons gain increased kinetic energy with increased electric
field. The second electron created in the collision is accelerated in the electric field
and can also create new ion pairs; this process is called a Townsend avalanche. The
9
avalanche will terminate when all electrons are collected at the electrode. Using the
effects of the Townsend avalanche, given the right circumstances, the secondary
ionizations can be kept proportional to the primary ionization events, but the total
numbers if ionizations can be increased significantly. This is the region called the
proportional region where the proportional counters are used, see figure (3).
Proportional counters are more sensitive to impurities in the detector gas than
ionization chambers, which can cause problems due to the formation of many
excited molecules or atom states created during the avalanche. The excited
molecules decay by photon emission which can cause new ionizations by either
photoelectric interaction in the gas or releasing electrons when interacting in the
detector wall. In a proportional counter these effects will cause a loss of
proportionality, increased dead time and reduced spatial resolution for positioning
sensing detectors. By adding another gas, a fill gas, the effects can be reduced by
absorbing the photons in a way that does not cause new ionizations. The amplified
charge gained when using a proportional counter requires less external amplification
of the signal, which makes the proportional counter have an increased SNR (signal to
noise ratio) compared with an ion chamber.
Further increasing the electric field will generate a linear amplified response from the
detector in a certain region; this is the region of true proportionality. Increasing the
electric field beyond this region will cause a non-linear response. The most important
contribution to this effect is caused by the slowly moving positive ions. During the
time it takes to collect the electrons the ions has hardly moved at all. The slowly
drifting ions creates a cloud of positive charge in the detector, if the numbers of ions
is high enough they can alter the shape of the electric field. This will cause nonlinear
effects to occur; due to the gas multiplication dependence of the strength of the
electric field. This region is called limited proportional region and is not a desirable
region of operation for any detector. See figure (3).
If the electric field is increased above the limited proportional region, the space
charge created by positive ions will determine the output pulse from the detector.
The high electric field is used to intensify the avalanches. In ideal conditions one
avalanche can trigger another avalanche at another point in the detector. This
chain reaction leads to an exponential growth of avalanches in the detector, called
a Geiger discharge. The avalanche will continue to a point where the space charge
from the ions reduces the electric field below the limit for continued gas
multiplication, thus making the process self limiting. The same number of positive ions
will be formed to cause the electric field to drop below the threshold of gas
multiplication, independent of the number of primary ionization events in the
detector. The output pulse from the detector will then be of the same magnitude
and all properties of the radiation are lost. This makes the detector only function as a
counter. This is the GM-region, see figure (3).
In the GM tube the effects of the excited molecules and atoms, that caused
problems for proportional counters, are desirable. The propagation of the Geiger
10
discharge is made possible by the photon emissions. In proportional counters each
avalanche is formed in a position that corresponds with the original position of the
ionization event. In the GM tube the discharge grows due to random formations of
avalanches, caused by the emissions and interactions of the emitted photons by to
cover the entire collection wire. This produces a massive amount of positive ions that
need to be collected which leads to a high dead time in GM tubes, see figure (4).
The increase in signal strength leads to less requirements of external amplification.
Figure (4). The propagation of a Geiger discharge [6]
Special precautions must be taken in Geiger counters to avoid creating a
continuous output loop of pulses. When the positive ions arrive at the collecting
electrode they are neutralized when combining with electrons from the electrode; in
this process energy is released. If this energy exceeds the energy needed to extract
an electron from the electrode surface, it is possible that a new free electron can be
released and create another Geiger discharge. This effect would then produce a
continuous output of pulses from the GM tube. To avoid this effect either external or
internal quenching can be used.
Internal quenching is done by adding an additional gas, quench gas, to the
detector gas. The quench gas prevents the continuous output by using the effect of
change transfer collisions. The quench gas has a lower ionization potential and more
complex molecular structure than the detector gas. When the positive ions collide
with the quench gas and transfer the charge, the ions are neutralized and the
quench gas molecules start to drift to the collection electrode instead. If the
concentration of the quench gas is sufficiently high all of the positive ions arriving at
the collection electrode will be of the quench gas. When they are neutralized the
energy released may go to dislocating the more complex molecular structure
instead of releasing an electron from the surface.
External quenching can be done by reducing the high voltage for a specific time
after each pulse, below the value for the gas multiplication to take effect.
11
For almost any detector system there will be a minimum amount of time needed for
two separate events in the detector to be recorded as two separate pulses. In some
detectors the limits are set due to processes intrinsic to the detector itself, in others
the surrounding electronic sets the limit. The minimum time needed to separate the
two events is called dead time. There is always a probability that true events may be
lost due to the randomness of radiation, dead time losses. At high counting rates
these losses can become severe, and methods for compensation have to be used
to get any accuracy in the measurement. These dead time losses affect almost any
detector system, but especially the GM-tube due to its design. The methods of dead
time compensation depend on the behavior of the detector system. There are two
common models used, a paralyzable or nonparalyzable detector system. The
nonparalyzable detector system can give a reading in high intensity radiation fields,
the paraplyzable detector system can fail to give a reading at all. The models
represent two extreme behaviors of an idealized detector system, where one, or the
other, usually describes the true detector system adequately. The two models differ
greatly from each other at high dead time losses but predict the same amount of
losses at low levels. Measurements taken under conditions with high dead time losses
should be avoided, to avoid the increasing error in the correction. When the dead
time losses are at 30 to 40 % the uncertainty is high, and efforts should be made to
reduce the dead time losses [6]. This can be done by either changing the measuring
conditions or by changing the detector system.
3.2 Scintillation detectors
Scintillation detectors are based on the principle of induced luminescence that is
produced from the detector material when exposed to ionizing radiation. In
scintillation detectors composed of organic materials the molecules are excited
through the kinetic energy absorbed from electrons that are released from photon
interactions in the detector material. The excited molecules, in the detector, return to
their original state by photon emission, and these photons are then collected. In
scintillation detectors of inorganic materials the atoms are arranged in a crystal
structure, and the crystal is excited by the energy absorbed in the passage of
electrons.
The requirements for a good scintillation material are:
•
•
•
•
•
•
•
High probability for interaction with photons.
Proportionality between the light emitted and the energy deposited in the
detector
Efficient conversion from kinetic energy to emitted light.
Transparency for the light emitted in the material.
The decay time for the induced luminescence should be short.
A refraction index close to that of glass.
Good optical quality and be able to manufacture in practical detector sizes.
12
The detector material should also be able to respond quickly to radiation. Before the
next photon strikes the detector all light from the previous interaction should have
been converted. This is essential to get a correct measurement.
The light emitted from the material then strikes the surface of a photo multiplier tube
(PM-tube). The front part of the PM-tube, facing the scintillation material, is coated
with a light sensitive material. When the photons, produced in the detector material,
strike the surface electrons are released. The electrons are then accelerated through
the PM-tube by an electric field. During the acceleration the electrons collide with
plates, called dynodes, where each collision releases additional electrons. As a result
an amplified signal is produced that can be further processed and amplified.
3.2.1 NaI(Tl) scintillation detectors
The NaI(Tl) scintillation detector is used for detection of photon radiation. The
detector has a relatively good energy resolution, which makes it possible to
distinguish between photons with different energies. The photoelectric effect is the
dominating way of interaction for low energy gamma rays and the probability for
interaction can be approximated by the expression
(10)[1]
Where z is the atomic number of the absorption material and hf is the photon energy.
Iodine has a high atomic number, this makes interaction by photoelectric absorption
significant and gives a high probability that the total photon energy gets stored in
the detector. With decreasing energy the probability for interaction also increases.
This inherent property makes the NaI(Tl) scintillation detector useful in situations where
the GM-tube will fail to give a reading. To increase the probability that the emitted
photons lies in the visible spectrum, a small amount of thallium is added to the
detector. NaI is sensitive to moisture, thus an enclosure of metal is often used. The
added metal will attenuate some of the incoming radiation, and reduce the
detection efficiency.
3.2.2 Liquid scintillation detectors
The scintillation material in a liquid scintillation detector is composed of a liquid and
the sample is mixed with the scintillation liquid. The light emitted from the solution is
then registered by a PM-tube in the same way as the other scintillation detectors. The
main advantage with a liquid scintillation detector is that samples containing low
can be measured with an
energy beta and alpha particles can be measured.
efficiency of about 50% [7].
Mixing the sample with the scintillation liquid can be complicated, not all samples
can be mixed directly without preparations. Sample containing more than one beta
emitting nuclide, with similar energies, can cause problems in distinguishing between
the different nuclides. This can be solved with separation methods.
13
3.3 Detector efficiency
If the radiation striking the detector is alpha or beta radiation, and if the particle has
travelled a small range in the detectors active volume, it usually has created enough
ionization and excitation to be counted. This makes it possible to arrange the
detector so that is sees every particle that enters its active volume, which makes the
detector have an intrinsic counting efficiency of 100%.
Gamma photons must first undergo a significant interaction in the detector to be
counted, making the detector have a counting efficiency less than 100%. This makes
it important to have a value of the number of pulses counted relative to the incident
particles striking the detector, the efficiency.
Two classes of efficiency can be defined
1. Absolute efficiency is defined by equation (11). The absolute efficiency is not only
dependent on the detector properties but also the measuring geometry.
(11)[6]
2. Intrinsic efficiency is defined by equation (12). The intrinsic efficiency does not
include the solid angle seen by the detector from the source.
(12)[6]
For a point source, the two efficiencies are related by the expression
(13)[6]
Where
is the solid angle of the detector seen from the source.
Ω is given by the integration over the detector surface that faces the source
according to equation (14)
(14)[6]
Where r is the distance from the source to the surface element dA, and
is the angle between the
normal of the surface element and the source direction.
The intrinsic efficiency primarily depends on the detector material, thickness of
detector material and radiation energy. A slight dependence on source distance
remains, because the path length of the radiation, in the detector, change slightly
with distance.
The efficiency can also be defined by how the events are recorded.
The total efficiency assumes that all pulses from the detector are accepted, no
matter how low energetic the interaction was. In practice, however, any measuring
14
system requires the pulse to be higher than a certain pulse-height set to discriminate
against electronic noise.
The peak efficiency assumes that only full energy interactions in the detector are
counted. These full energy interactions are usually seen as the peak at the end of a
differential pulse height distribution.
The quotient, r, of the efficiencies are related by equation (15)
(15)[6]
The peak efficiencies are commonly tabulated, because the full energy events are
not as sensitive to scattered radiation and false pulses. The detector should be
specified according to both efficiency criteria. The most common efficiency
tabulated for a gamma ray detector is the intrinsic peak efficiency. If the detector
has a known efficiency it can be used to measure the activity of a radioactive
source, equation (16).
(16)[6]
Where S is the number of emitted particles from the source during the measuring time and N is the
number of recorded events.
Equation (16) can be rewritten by using equation (13) and using that the
source strength (point source) equals the activity multiplied by the sum of the
branching ratio (
. The result is seen in equation (17)
A=
Where A is the activity, N is the number of recorded events,
(17)
is the efficiency and fi is the probability
for decay for each emission.
Equation (17) is used for calibration or constancy control purposes for calculating the
activity. By solving equation (17) for the efficiency is given, equation (18)
=
Where A is the activity, N is the number of recorded events,
(18)
is the efficiency and
is the probability
for decay for each emission.
3.4 Regions of measurement and Energy response
Ion chamber detectors can be used to measure gamma and x-ray radiation down
to a few tenths of a mSv/h [8]. At lower levels the chamber dimensions needs to be
increased, for incased sensitivity. Although the increase would make the chamber
15
too large for portable use, there is however other designs of ion chambers, like the
pressurized ion chamber and liquid ion chamber that can be used at lower radiation
levels where the normal ion chamber will fail to give a reading. To measure the
radiation in the lower regions a GM-tube or scintillation detector can also be used.
A typical photon energy response function is shown in figures (5), which show that
the ion chamber gives a flat response when measuring photon energies between 0.3
to 10 MeV. At low photon energies the response decreases rapidly. GM-tubes and
scintillation detectors have a significant peak at lower photon energies, but the
response is relatively uniform at higher energies. A uniform response is achieved by
built in energy compensation, e.g. added window material. [8]
Figure (5). The relative response for different detectors plotted versus the photon energy. [8]
4. Calibration of radiation monitoring instruments
Radiation monitors are usually calibrated by using one well known standard source
with a known energy. If the monitor is used to measure at a different energy, it can
give a significant over- or under estimation of the measured radiation. This makes
checking the monitor before a measurement important, to make sure that the
correct monitor is used.
Small quantities of radioactivity that pose an insignificant external radiation hazard
can cause a significant internal radiation hazard. The amount of a radioactive
substance, from a contaminated area, that is able to cause internal hazard is
generally lower than a level that would cause an external hazard. Monitors used to
detect contamination then needs to be more sensitive than radiation survey
monitors. The contamination monitors are usually detectors with built in amplifiers
(GM tubes, proportional counters or scintillation tubes). The activity is recorded as
counting rate, and the monitor needs to be calibrated for the contamination to be
calculated [8].
16
Radiation survey monitors are usually calibrated to measure the ambient dose
equivalent ( ). This makes it easy to check that the limit of radiation exposure is not
exceeded when the energy or direction of the radiation is unknown [5].
There are two methods commonly used when calibrating radiation monitor
instruments.
• Indirect calibration (intercalibration)
In this method the response from the radiation monitoring instrument under
calibration is compared to the response of a reference instrument, see figure (6).
The reference instrument used has to be calibrated against a higher quality
reference instrument [9]. Care should be taken to minimize the scattered
radiation, it can cause problems when detectors have different energy response
[8].
Figure (6). Intercalibration [9]
• Direct calibration
The radiation monitor that is to be calibrated is placed in a known radiation field
from known standard sources. A schematic setup is presented in figure (7). This is
the method proposed in this thesis.
Figure (7). Direct calibration [9]
There are several important parameters intrinsic to the radiation monitor that has to
be known. The most important that should be thoroughly examined are, sensitivity to
radiation, energy response, rate response and temperature response. This is normally
done by the manufacturer before the monitor is released to the costumer [8]. A
calibration of radiation monitors in the true meaning of the word is beyond this thesis,
but the calibration can be investigated, see section 2.
17
The radiation monitors sensitivity is the parameter that most likely will change over
time [8]. This makes it the most important parameter to check when performing a
constancy control. This can be done by measuring the same source in the same way
and recording the results. If the measured value deviates significantly from the
expected decrease due to natural decay further investigations could be made.
The reasons for a changed response are various, here are a few examples.
-
-
Power problems.
Battery problems or damaged wires
Electric field problems
The collecting electrodes could have been bent or damaged. This would
alter the electric field.
Contamination problems
The detector gas could have been contaminated with air from a small
leak. Air (oxygen) is an electronegative gas, and could cause problems for
many detectors if the concentration is high enough. [6]
When the detector has been repaired it should be recalibrated before being put to
use.
5. Material and methods
5.1 Dose rate estimations from point sources
5.1.1 Estimation of beta point dose rates for calibration and constancy control
sources
Estimation of the beta point dose rate in air is done to get the conversion factor from
cps to dose rate in air (section 5.4). The beta particles strong dependence of air
attenuation on energy makes it hard to find a simple expression for the dose rate. No
programs like Microshield for beta point source were available. The calculations will
only give an approximate value, for methods with higher accuracy see the
discussion section for references.
•
Estimation by using energy fluence with beta attenuation coefficient
The energy fluence at a distance d from a beta point source can be used to
calculate the dose rate at a point in air, and the energy fluence is given by equation
(19)
(19)[3][10]
Where r is the distance from the source to the measurement point,
per decay,
is the beta energy attenuation coefficient and
is the mean beta energy (in MeV)
is the areal density
The areal density is given by equation (20)
18
(20)[3]
Where d is the distance from the point source to the dose point and
is the density for NTP air
The beta attenuation coefficient, in air, is given by equation (21)
(21)[3]
Where
is the maximum beta energy (MeV)
The beta attenuation coefficient, in tissue, is given by equation (22)
(22)[3]
Where
is the maximum beta energy (MeV)
To calculate the dose rate at the measuring point the energy fluence must be
converted into absorbed dose rate in the medium. This is done by multiplying
equation (19) with the beta attenuation coefficient of interest. The final expression for
the dose rate in air at distance d is;
(23)[10]
Where
is the energy fluence a distance d from the source and
is the beta attenuation
coefficient in air
By using the beta attenuation coefficient for tissue, the absorbed dose to tissue is
given. An estimation of the dose to the skin or organs can then be done by using
equation (24)
(24)[3]
Where
is the beta attenuation coefficient, t is is the distance traveled by the radiation in tissue, D is
the dose of interest and
•
is given by equation (23).
Estimation by using the beta rate constant
An estimation of the dose rate from a beta point source can be given by equation
(25)
(25)[2]
Where d is the distance in meters from the source, A is the source activity(Bq),
constant and
is the beta rate
is a function that compensates for the energy loss suffered by the beta particles
traveling the distance d.
is used as a correction function to compensate for the energy loss of the
beta particles due to interactions along the path, d.
is the specific beta rate
19
constant, which depends on the mean beta energy. See [2] for values of the
expressions of the nuclide of interest.
5.1.2 Estimation of gamma point dose rates for calibration and constancy control
sources
•
Microshield
Microshield is used to calculate the effective dose (E ) (see section 2) for an isotropic
geometry at the measurement point for the constancy control and calibration
sources at the distances of interest.
•
Approximate value by gamma rate constant
The dose rate from a gamma emitting nuclide can be estimated by equation (26)
(26)[11]
Where d is the distance from the source in meters, A is the source activity,
s the gamma rate
constant
5.2 Estimation of surface dose rate from waste containers
When measuring waste containers an estimation of the surface dose rate can be
useful to know what dose level to expect. The estimation of the dose levels were
, of each nuclide
made with the maximum allowed amount of activity, 1
investigated.
Two different waste containers are used for disposal of radioactive nuclides at the
University of Gothenburg, table (1) list some general properties for the containers.
These properties are used for the estimations.
Table (1). Properties for the two types of waste container used at University of Gothenburg.
waste
container
length
(cm)
width
(cm)
height
(cm)
volume
(
)
”large”
”small”
35
25
27
21
42
45
39690/38
23625/22
mean
weight
(kg)*
4
2
density
(g/
)
0,1
0,085
*the mean weight of a sealed waste container filled with radioactive waste.
5.2.1 Estimation of surface dose rate from waste containers filled with gamma
emitting nuclides
20
The surface dose rates, in air, from a gamma emitting volume source were
calculated with Microshield. Three different geometries were studied. For waste
containers with homogeneously distributed activity, the geometry shown in figure (8)
is used. The measurement point is placed 1 cm above the center point of the top
surface. This is done to better simulate a real measurement situation. For a complete
description on how the source is defined in Microshield, see the appendix.
Figure(8). Geometry 1, homogeneously distributed activity. The measurement point is placed 1 cm
above the surface.
For waste containers where the activity has been concentrated to the bottom half of
the container and the measurement is made at the top, the geometry shown in
figure (9) is used. The absorbing material is set to have the same density as the
volume source, but contains no activity. The thickness of the absorber was set to 20
cm for both waste containers. The measurement point is placed 1 cm above the
center point of the top surface, to better simulate a real measurement situation. For
a complete description on how the source is defined, see the appendix.
21
Figure (9). Geometry 2, inhomogeneous distribution of activity. The measurement point is placed 1 cm
above the surface.
For waste containers where most activity has been concentrated to one end, the
geometry shown in figure (10) is used. The thickness of the source is set to 10 cm and
the thickness of the absorber is set to 1 cm for both waste containers. This geometry
should estimate the higher end of the spectrum of possible dose rates. The
measurement point is placed 1 cm above the center point of the top surface. For a
complete description on how the source is defined, see the appendix.
Figure (10). Geometry 3, concentrated distribution of activity. The measurement point is placed 1 cm
above the surface.
5.2.2 Estimation of surface dose rate from waste containers filled with beta emitting
nuclides
The estimation of the surface dose rate is done by calculating the dose rate by
hand, no programs like Microshield for beta volume emitters were available.
For an infinitely thick volume source (source thickness
beta particle range), the rate
of energy absorption is equal to the energy emission for a point in the volume. This
constitutes an equilibrium called ESE (energy spatial equilibrium). The dose rate inside
the volume, figure (11), when conditions for ESE apply is given by equation (27).
(27)[3]
Where
is the concentration of the beta emitting nuclide, tps is transformations per second, and
is
the mean energy (MeV) per beta particle
22
Figure(11). The dose rate for the measurement point inside the volume under ESE conditions.
Which can be reduced to equation (28)
(28)
Where
is the concentration of the beta emitting nuclide and
is the mean energy per beta particle
At the surface of such a volume source the energy absorption rate will be
approximately half of what it is at a point within the volume, since source material will
be present only on one side of the dose point, figure (12). Using equation (28), the
surface dose rate can be written as;
=
Where
is the concentration of the beta emitting nuclide and
(29)[3]
is the mean energy per beta particle
Figure (12). The surface dose rate at the measurement point from a beta emitting volume source
If the volume source emits beta radiation with different energies or if the source is
made up of multiple nuclides, equation (29) can be rewritten as;
(30)[3]
Where
is the number of beta particles per decay with
Equation (29) or (30) will estimate the beta dose rate at the surface of the volume
source. The surface dose rates for different nuclides were calculated for the large
waste container, since the larger waste container better fulfills the conditions for ESE.
23
The bremsstrahlung produced when the beta particles interact in the volume source
could be included. The fraction of beta energy converted into bremsstrahlung is
given by equation (31)
(31)[3]
Where
is the effective atomic number of the source and
is the maximum energy in MeV of the
beta radiation
Bremsstrahlung is mostly of interest for the very low energy emitting nuclides, H-3 C14, where the beta dose rate will be hard to measure due to the short range of the
particles. The dose rate measured is primarily from bremsstrahlung contribution.
Estimations of bresmsstrahlung
-
Approximate expression
The bremsstrahlung is set to be emitted from a virtual point in the middle of the
source at a distance r form the dose point. The dose rate in air, given the
conditions above, is given by equation (32)
(32)[3]
Where
radiation,
is the effective atomic number of the source,
is the maximum energy of the beta
is the linear attenuation coefficient and r is the distance from the virtual emission point
to the dose point.
-
Microshield
A gamma emitting source, defined according to figure (8) and equation (31),
with the photon energy as the maximum energy of the beta particles could give
an upper estimation of the gamma dose rate.
The result from equations (32) or Microshield can be combined with the beta dose
rate to give the total surface dose.
5.3 General measurement steps used to measure constancy control and calibration
sources
1. Investigate the detector for any sign of damage, to wires or to the detector
itself.
Check the status of the battery or power supply. If the battery level is low most
monitors will not work, or if measurements are made it will give an incorrect
reading. Most instruments have a built in warning system to avoid this. Refer to
the instruments manual.
24
Check the monitor for a reference measurement marking, this marking should
be placed facing the source at every calibration or constancy control
measurement.
Clear the immediate surrounding area of unnecessary objects, to reduce
contribution from scattered radiation.
2
Place the detector in a jig or use some other sort of fixation at the chosen
calibration distances from the source, see figure (13). This is done to keep the
distance as constant as possible, and make repeated measurements easy to
perform.
The fixation shown on the left side of figure (13) should not be placed in a way
that interferes with the radiation from the source.
The fixation can, for example, be made up of a thin plastic tube. It is important
that the same fixation tool is used, or that the distance is the same, when
performing future measurements.
Figure (13). The measuring setup during calibration and const. control.
The distances used, for the sources, are given in table (2)
3
Measure the background radiation without the source present. If the count
rate is higher than normal, it can indicate that the monitor or the surrounding
25
area is contaminated. The background radiation level should be recorded, for
future references.
4
Perform a measurement with the source properly in place. Wait at least ten
seconds before recording the value, some detectors might respond slowly to
the radiation.
The results from step 4 can be recorded in a spread sheet program (see the
appendix), that also can be used to evaluate the result and be used for future
references.
5.4. Calibration method of radiation monitor instruments measuring counting rate
As mentioned in chapter 4, the manufacturer has performed an extensive
investigation of the instrument. The instruments sensitivity, energy response, rate
response and sensitivity to temperature variations has usually been examined. The
user is usually referred to the instrument manual for any information of interest.
There are two types of calibrations that can be performed for the monitors
measuring cps.
-efficiency calibration
The efficiency calibration is done to calculate the activity of the point source that is
indicated as counting rate on the monitor. The calibration can be done by following
the general measurement steps and recording the result from step 4. Using Equation
(18), with the corrected count rate, gives the efficiency for the nuclide measured.
With known efficiency the activity of the measured radiation can be calculated with
equation (18). If multiple calibration sources are available the efficiency over a wider
energy spectrum can be investigated by repeating the steps 1 to 4 for each
calibration source.
- Estimation of the conversion from cps to dose rate in air
A calibration concerning the conversion of counting rate to dose rate in air is done
by following the same general measurement steps. The counting rate given in step 4
can be related to the calculated dose rate in air, at the measurement point for the
source. The dose rate at the calibration point is given in table (2).
5.5 Constancy control method of radiation monitor instruments measuring counting
rate
When performing the constancy controls, the measuring setup should be as close to
identical as possible. The source is measured by following the general measurement
steps.
26
If preferred, just the counting rate could be recorded. The results from the following
constancy control measurements are compared with the result from the first
measurement by compensating for the decay of the source, see the appendix. The
results can determine if the monitor has changed its response compared to the
previous constancy control measurements.
5.6 Calibration method of radiation monitoring instruments measuring dose rate
An investigation of the calibration of instruments measuring ambient dose equivalent
(H*) or personal dose equivalent ( ) can be done by following the general
measurement steps and comparing the measured value with the calculated value
for effective dose rate (E) at the measurement point, see table (2).
The indicated value should be higher than the calculated value for a correctly
calibrated monitor, see section 2.
5.7 Constancy control of radiation monitoring instruments measuring dose rate
The source is measured by following the general measurement steps. The result from
the first constancy control is recorded. For following constancy controls, the result
can be related to the value of the first constancy control by compensating for the
decay of the source. The results can determine if the monitor has changed its
response compared to the previous constancy control measurements.
5.8 Calibration and constancy control sources
IAEA recommend that both point and surface sources are available for monitor
calibrations since they make up the extremes of the measuring geometry [9], this
thesis only deals with point sources.
The sources used for calibration have to be chosen carefully. The nuclide should
decay in as few ways as possible, to make the measurement situation during the
calibration as accurate as possible. The constancy control nuclides should have a
long half-life, to make the constancy control valid over a long period of time. For
convenience a single nuclide per monitor should be used.
The calibration and constancy control sources chosen are presented in table (2). All
sources have a relatively long half life and have a well known decay, see the
appendix.
Table(2). The constancy control sources, their activity and dose rate at the measuring point.
nuclide
,
#1
nuclide
activity
176 kBq
(2008-0601)
activity
measurement
distances
5 cm & 10 cm
measurement
distance
E **
D
3 µSv/h (5 cm)
0.83 µSv/h (10 cm)
4.4 µGy/h (5 cm)
1.2 µGy/h (10
cm)
D
E **
27
,
#2
Nuclide
#3
nuclide
183 MBq
(2008-0601)
activity
211 kBq/ml
(2008-0523)
activity
37 kBq/ml
10 cm
0.87 mSv/h (10 cm)
measurement
distances
5 cm & 10 cm
measurement
distance
1 cm
1.26 mGy/h (10
cm)
D
No source
D
No source
#4
*The
source is a
source shielded for
beta contribution, see section 5. **Calculated value
using Microshield, 1 mm pmma slab is placed over the source to remove the beta particle contribution.
5.5.3 Construction of calibration and constancy control sources
A point source for
and
has to be constructed. The sources have to be
contained in a proper way to avoid unnecessary contamination of the environment
or the detector itself.
The sources should also be constructed to make sure that the dose level outside the
container does not pose an unnecessary external radiation hazard.
The
source has to be unshielded under measurement. Any seal would remove
the beta particles and avoid detection in the instrument. The base of the source
should be constructed in aluminum to avoid the possibility that built up static
electricity, generated in e.g. plastics, would be able to force the molecules to
spread out and cause contamination. Aluminum has a low atomic number; this gives
a low bremsstrahlung contribution.
The base of the constructed point sources is presented in figure (14)
Figure (14). The base of the source. x and y is the width, z is the thickness, d is the diameter and h is the
depth of the hole.
The dimensions and material for the different sources are presented in table (3)
Table (3). The material and dimensions for the bases for the point sources that was considered.
28
material nuclide
aluminium
aluminium
The nuclides of
x (cm)
5
5
and
y (cm)
5
5
z (cm)
1-2
1
d (cm)
0.4
0.4
h (cm)
0.4
0.4
containment
1 mm PMMA
N/A
are in liquid solutions. An appropriate amount of the
solutions were placed in the base of the aluminum slab and left to dry, resulting in the
final activity seen in table (2). The handling of the solutions was done under a closed
hood.
6. Results
6.1 Dose levels around waste containers
6.1.1 Surface dose rates from gamma emitting nuclides
The results from the estimation made with Microshield are presented in table (4) and
table (5) . Table (4) shows the estimations for the smaller waste container and table
(5) shows the estimations for the lager waste container.
Table(4). The surface dose rates from the three different geometries estimated by Microshield for the
small waste container.
Nuclide (
)
with buildup
with buildup
with buildup
with buildup
with buildup
with buildup
with buildup
Small waste container
Geometry 1
Geometry 2
(µGy/h)
(µGy/h)
206
21
Geometry 3
(µGy/h)
277
220
28
297
17
1.6
24
19
2.8
27
9.7
2.3
24.7
10
2.6
25.3
82
11
203
95
17
231
0.59
1.4
0.67
1.7
1.8
0.4
4.9
2
0.5
5.2
11
2.3
12
2.6
29
31
with buildup
29
Table(5).The surface dose rates from the three different geometries estimated by Microshield for the
large waste container.
Nuclide (
)
Large waste container
Geometry 1
Geometry 2
(µGy/h)
(µGy/h)
133
19.7
with buildup
with buildup
with buildup
with buildup
with buildup
with buildup
with buildup
Geometry 3
(µGy/h)
191
146
27.5
210
11
1.5
19.5
13
2.8
23
8.9
2.4
21
9.4
2.7
21.6
67
11
157
80
18
183
0.34
0.75
0.39
0.91
1.4
0.35
3.4
1.6
0.65
3.6
8.7
2.2
20
9.4
2.7
22
with buildup
The results show that the smaller container has a higher surface dose rate for all
for geometry 2 for both waste containers
nuclides, that the surface dose rate for
are well below the legal limit (5 µGy/h) and that the surface dose rate for e.g.
is
significantly above the legal limit.
Table (6) shows the results for the large waste container if the density is reduced by
half.
Table(6). The large waste container with half the density, results are shown for geometry 1.
Large waste container
Nuclide (ALImin)
Geometry 1
(µGy/h)
9.2
with buildup
9.4
1.5
with buildup
1.6
30
As seen in table (6), the dose rate for
and
is almost identical with the dose
rate in table (5).
6.1.2 Surface dose rates from beta emitting nuclides
Table (7) shows the beta surface dose rates calculated for the large waste
container, with the activity of 1 ALImin.
Table(7). The concentration, mean energy and surface dose rates for beta emitting nuclides in the large
waste container
Nuclide
***
*
1338
330
410
595
320
1.3
1090
0.18**
52
* Calculated using equation (32). **The sum of the two dominating beta energies. *** see [11]
Table (7) shows that the surface dose rate from the beta emitting nuclides is above
. The daughter nuclide of
is
the limit of 5 µGy/h for all nuclides except
however
.
If the mass of the waste container change, the change in dose rate will be linear
(equation (30)). Due to all parameter are constant, but . A reduction in mass by
50% will lead to an increase of
by a factor two.
If the mass is 2 kg for the large container, the ESE conditions are just met for the high
, see equation (3). Which gives a range of approximately 20 cm in the
energetic
container.
Equation (31) gives f as 0.0032 for C-14 using the effective atomic weight of 7
(approximately air). The contributions from bremsstralung is low.
31
6.2 Calibration and constancy control
Table (8) shows the result for the instruments measuring the ambient dose equivalent
using the general measurement steps with source #1.
Table (8).The results from the instruments measuring the ambient dose equivalent
Instrument Department
number
1
MFT**
2
MFT**
3
Dep. Of
radiation
physics
Dep. Of
radiation
physics
Dep. Of
radiation
physics
Dep. Of
radiation
physics
4
5
6
H* instruments measured with source #1.
Instrument
Background
Measured
type
[µSv/h]
H*(10) [µSv/h]
(5 cm/10 cm)
dose error
(intrinsic
error)
RNI 10/SR
Intesimeter
Smart Ion
0.10
3.25/1.25
Calculated
E [µSv/h]
(5 cm/10
cm)
3/0.83
2.5
RNI 10/SR
Intesimeter,
S/N 59855
RNI 10/SR
Intesimeter,
S/N 59857
SRV-2000
0.20
Not
considered***
3.55/1.75
Not
considered
3/0.83
Not
considered
20% [16]
0.21
3.69/1.55
3/0.83
20% [16]
0.15
3.34/1.52
3/0.83
20% [16]
Canberra
Radiagem
SAC 100
0.20
3.96/*
3/0.83
15% [16]
20% [16]
.*The lowest recommended dose rate for measurement was 3 uSv/h, the distance 10 cm was therefore
not considered. **Dep. Of Medical physics and biomedical engineering Sahlgrenska University Hospital
*** due to the high background radiation value measured. c the uncertainty in positioning is also
considered (estimated to 10% for 5 cm and 7% for 10 cm)
The results from table (8) show that all instruments, except instrument #2, measures a
dose rate higher than the calculated effective dose at the measurement point,
taking the statistical factors in mind. The instrument #2 measured a significantly high
dose rate, and was therefore not considered.
Table (9) shows the results for the instruments measuring counting rate using the
general measurement steps with source #1.
32
Table (9). The results for the instruments measuring cps.
Department
MFT**
Dep. Of
radiation
physics
Dep. Of
radiation
physics
Counting rate instruments measured with source #1.
Instrument type
Background cps
ε*
(cps)
(5cm/10cm) (5cm)
Berthold LB 1210 B
Miniseries 900, 44B
9 cps
20 cps
510/260
500/210
0.0033
0.0032
Dair/cps
(5 cm)
[µGy/h
cps]
0.0087
0.0092
Exploranium Gr100 G NaI
320
2450/1419
0.014
0.0021
*calculated with equation (19) **Dep. Of Medical physics and biomedical engineering Sahlgrenska
University Hospital
The result from table (9) show that the instruments are relatively insensitive to the
point source, the instrument Berthold LB 1210 B and Miniseries 900 with probe 15EL
should be re-measured with the beta source #3 when available.
Table (10) shows the result for the instruments measuring the ambient dose
equivalent using the general measurement steps with source #2.
Table (10).The results from the instruments measuring the ambient dose equivalent.
H* instruments measured with source #2.
Instrument Department Instrument
Background
Measured
number
type
[µSv/h]
H*(10)
[mSv/h]
(10 cm)
1
Dep. Of
RNI 10/SR
0.21
1.22
radiation
Intesimeter,
physics
S/N 59855
2
Dep. Of
RNI 10/SR
0.26
1.16
radiation
Intesimeter,
physics
S/N 59857
SRV-2000
0.17
1.34
3
Dep. Of
radiation
physics
4
Dep. Of
Canberra
0.22
1.41
radiation
Radiagem
physics
SAC 100
Calculated
dose
E [mSv/h]
error
(10 cm)
(intrinsic
error) c
0.87
20%
[16]
0.87
20%
[16]
0.87
20%
[16]
0.87
15%
[16]
The results from table (10) shows that all the instruments indicated a measured dose
rate that is higher than the calculated effective dose rate.
Table (11) shows the results for the instruments measuring cps with source #2. The
source was too strong for all instruments, and not considered for the Miniseries 900,
15EL instrument.
33
Table (11). The results for the instruments measuring cps.
Counting rate instruments measured with source #2.
Department Instrument type
Background cps
(cps)
(10cm)
Dep. Of
Miniseries 900, 44B 25 cps
overload
radiation
physics
Dep. Of
Miniseries 900,
3 cps
radiation
15EL
physics
Dep. Of
Exploranium Gr320
overload
radiation
100 G NaI
physics
Result from table (8) can be summated into figure (15) for the instruments measuring
the ambient dose equivalent with source #1 at 5 cm
Figure (15). The measured ambient dose equivalent for source #1 at 5 cm
Results from table (8) can be summarized into figure (16) for the instruments
measuring the ambient dose equivalent with source #1 at 10 cm
34
Figure (16).The measured ambient dose equivalent for source #1 at 10 cm
Result from table (10) can be summarized into figure (17) for the instruments
measuring the ambient dose equivalent with source #2 at 10 cm
Figure (17).The measured ambient dose equivalent for source #2 at 10 cm
7. Discussion
The investigated instruments that measure the ambient dose equivalent could not
be said to give a reading that is lower than the calculated effective dose, taking
35
statistical factors in mind. This is an important result as it means that they will not
underestimate the effective dose that is related to the risk of the exposure.
To better be able to evaluate the monitors measuring the ambient dose equivalent,
a calibrated monitor (with traceable calibration) for one or several energies of
interest should be used. The evaluation is then done by the method of indirect
calibration, and the calibrated monitor is used as a reference. References [9] and [8]
discussed this method, and it is the recommended method if several detectors are
supposed to be investigated. Care should be taken to reduce the scattered
radiation if the monitors have different energy response. Another way is to order
calibration nuclides that have the specified dose rate of interest given in a
calibration certificate.
The results from the measured instruments, Figure (15), with source #1 at 5 cm show
that the deviations are quite high. If the instruments are to be controlled with the
source, the distance of 10 cm is recommended, figure (16). A jig is also
recommended, for complete fixation during the measurement time.
The result from source #2, figure (17), gave a very stable read out for each of the
instruments measured. However the activity of the source is quite high and gives an
unnecessary exposure to the worker for a simple constancy control, and it needs to
be handled carefully.
Unfortunately either source #1 or #2 is an ideal constancy control source. A more
appropriate constancy and control source would be a Cs-137 source with about 10
to 20 times the activity of source #1. The effective dose from the new source could
be calculated with Microshield, and the instruments could be re-measured with the
new source by following the general measurement steps.
The instruments measuring counting rate was too insensitive to source #1 (except
Exploranium Gr-100 G NaI) and too sensitive for source #2, the ideal source could be
the one mentioned above. There is not enough data to draw any conclusions for the
conversion factors from cps to dose rate in air.
The beta source that would be used for most instruments measuring cps, and surface
contamination, is unfortunately still in the making. The response for the beta source
should be better, since it is easier for most gas based monitors to detect the
radiation. To get a calibration for Bq/cm3 the contamination measuring instruments
have to be calibrated with a volume source.
The equations used in this thesis just give an estimation of the beta dose rate in air. To
better estimate the dose rates the ICRU report 56 [15] should be consulted.
The surface dose rate for the small waste container is higher than the large
container, due to the higher concentration of nuclides per unit mass.
The waste container geometry 3 should give a good estimation of the possible high
region doses, and the geometry 2 should give a good estimation of a low region of
36
possible doses. Many calculated values are above the legal limit, it can easily be
reduced just by waiting for the nuclide to decay if it has a short half life. Another
way is to repackage the container into a larger container for added attenuation. A
split of one waste container into two separate would also reduce the dose rate by
half, given a uniform distribution.
The dose rates could also be calculated with a measurement point placed on the
side on one waste container, the top point was chosen for convenience and to
simulate three possible types of distributions.
The surface dose rates from the beta emitting nuclides is just a theoretical value, and
will differ significantly from the measured value for the most low energetic beta
radiation (especially H-3). H-3 has an extremely short range and cannot be
measured with detectors like GM-tubes or proportional counters. Even C-14 will be
hard to measure; most of the signal could be a result from the low bremsstrahlung
contribution.
37
References
[1] Joniserande strålningens växelverkan med material, Lars Hallstadius & Sven
Hertzman, radiofysiska inst., Lund, 1983 [in Swedish]
[2] The Physics of radiation protection, B Dörschel et.al, Nuclear Technology
Publishing, 1996
[3] Introduction to health physics, Herman Cember, Mc Graw-Hill, Third edition, 1996
[4] Introduction to radiological physics and radiation dosimetry, Frank Herbert Attix,
Wiley, 2004
[5] Dosbestämning I strålskyddsarbete, Lennart Lindborg, SSI, Strålskyddsnytt nr 4, 1998
[in Swedish]
[6] Radiation detection and measurement, Glenn F Knoll, Wiley, Third edition, 1999
[7] Grundläggande strålningsfysik, Mats Isaksson, Studentlitteratur, 2002 [in Swedish]
[8] An introduction to radiation protection, Alan Martin & Sam Harbison, Hoder
Arnold, Fifth edition, 2006
[9] IAEA Safety report series no 16, Calibration of radiation protection monitoring
instruments, 2000
[10] Health Physics Society, especially
http://www.hps.org/publicinformation/ate/q3896.pdf, visited in 050508
[11] Strålskydd, Curt Bergman et.al., natur och kultur, 1988 [in Swedish]
[12] SSI FS 2000:7, Statens strålskyddsinstituts föreskrifter om laboratorieverksamhet
med radioaktiva ämnen i form av öppna strålkällor, SSI, 2000 [in Swedish]
[13] SSI FS 1983:7, Statens strålskyddsinstituts föreskrifter m.m. om icke
kärnenergianknutet radioaktivt avfall, SSI, 1983 [in Swedish]
[14] ICRP 30, Limits for the Intake of Radionuclides by Workers, ICRP, 1978
[15] ICRU 56, Dosimetry of External Beta Rays for Radiation Protection, ICRU, 1997
[16] Inventory of Radiation Monitors at Göteborg University and development of
control and calibration protocol, Elen Monsen, Master Thesis, Department of
radiation physics, University of Gothenburg, 2007
[17] Radiation safety manual, office of radiation, chemical and biological safety,
Michigan state university, 1996
[18] The Lund/LBNL Nuclear Data Search,
http://nucleardata.nuclear.lu.se/nucleardata/toi/, visited in 051408
38
Nuclides used at University of Gothenburg (see [11] [17] [18])
Halflife
12.3 y
ALImin 3000 MBq
Radio toxicity D
General properties
decays to
by emission of beta particles. The beta particles have a
maximum energy of 18.6 keV and an average energy of 5.7 keV. The beta particles
from
has a short range, approximately 6 mm in air and 6 µm in tissue. The
external dose contribution from bremsstrahlung is negligible.
Detection and measurement
Portable detectors like GM-tubes or NaI-detectors will not detect
, due to the low
energy of the beta particles. The particles cannot penetrate the entrance window
of the detector. Wipes should be taken over the area of interest and measured in a
liquid scintillation detector. Measurement of airborne activity can be performed
when dealing with high activity levels. The airborne activity can be measured using
monitors where the air is filtered through water, and the activity in the water is
measured.
Radiation protection
The beta particles can not penetrate the dead layer of the skin or be measured
with TLD or film dosimeters, due to the short range. Thus a radiation monitoring
dosimeter will not give a reading.
can be hazardous if it enters the body,
causing internal contamination. Many solutions marked with
evaporates, this
causes airborne activity and a risk to the lungs when inhaling.
distributes itself
evenly upon entry in the body. The critical organs are body water or tissue.
The beta dose rate, from a 37 MBq point source at 2.5 mm, is 103 Gy/h. At 5 mm the
dose rate has reduced to 0,3 Gy/h.
39
Halflife
5730 y
ALImin 97 MBq
Radio toxicity C
General properties
decay to
by emission of beta particles. The beta particles have a maximum
energy of 156 keV and an average energy of 49 keV. The range from the beta
particles is approximately 24 cm in air and 0,3 mm in tissue. The photon contribution
from bremsstrahlung is negligible
Detection and measurement
Measurement can be done with a GM-tube fitted with a thin entrance window, the
measurement must be performed at close range, ca 1 cm. Wipes taken from the
area of interest can be measured in a liquid scintillation detector.
Radiation protection
Due to the short range of the beta particles, radiation monitoring dosimeters will not
give a reading and 1% of the beta particles can penetrate the dead layer of the
skin.
is not significantly volatile at room temperature.
is hazardous if it enters
the body causing internal contamination. Inhalation of airborne activity or
absorption through the skin is a large risk. The critical organs are fat tissue or bone. In
dealing with activity levels (above 40 MBq)
should be handled under a closed
hood. Checking for contamination is important. With its long half-life, it can cause a
waste management problem.
The beta dose rate, from a 37 MBq point source at 1 cm, is 12,4 Gy/h. At 2 cm the
dose rate has reduced to 2,5 Gy/h.
40
Halflife
14.3 d
ALImin 10 MBq
Radio toxicity C
General properties
decay to
by emission of beta particles. The beta particles have a maximum
energy of 1,709 MeV and an average energy of 0.690 MeV. The range from the
beta particles is approximately 6 m in air and 8 mm in tissue.
gives an external
exposure due to photon contribution from bremsstrahlung, with a HVL value of 2
mm in tissue.
Detection and measurement
The preferred detector is a GM-tube with a thin entrance window or pancake
probe. NaI-detectors can be used to detect bremsstrahlung photons. Wipes from
the area of interest can be measured in a liquid scintillation detector to detect
removable surface contamination.
Radiation protection
gives both an external and internal dose. Radiation monitoring dosimeters
should be worn when handling
. As well as hand dosimeters. Eyes and skin are at
risk when exposed to the external radiation. Shielding should always be used and
protective glass should be worn to prohibit exposure to the eyes. Distance tools
should be used when opening voiles, to reduce skin doses. If
enters the body,
causing internal contamination, the critical organs are bone (for soluble forms), lung
& GI tract (for insoluble form) when inhaling or ingesting the nuclide.
When dealing with large activities, above 400 MBq, a TLD dosimeter should be worn
inside the protective glove and urine samples should be collected. An intake of 8
MBq gives a dose of 50 mSv in a year.
The dose rate, from a
MBq point source at 1 cm is 3.48 Gy/h. At 15 cm the dose
rate has reduced to 0,01 Gy/h.
Halflife
87,4 d
ALImin 80 MBq
Radio toxicity C
General properties
decay by emission of beta particles. The beta particles have a maximum energy
of 167 keV and an average energy of 53 keV. The range from the beta particles is
approximately 26 cm in air and 0.3 mm in tissue.
gives a small external exposure
due to bremsstrahlung contribution.
Detection and measurement
A GM-Tube with a thin entrance window can be used, due to the low energy of the
beta particles the measurement must be performed at a distance around 1 cm. The
detection efficiency is low, typically around 4-6%. Wipes measured in a liquid
scintillation detector can be used to find removable contamination.
Radiation protection
beta energy is low, and the external exposure is not hazardous. Shielding is
optional, a 3 mm Plexiglas shield can be used. Radiation monitoring badges will not
41
give a reading, due to the low energetic beta particles. 12% of the beta radiation
enters the body, causing internal
can penetrate the dead layer of the skin. If
contamination, the critical organs are the testis.
The beta dose rate, from a
Bq point source at 1 cm, is 11.7 Gy/h. At 2.5 cm
the dose rate has reduced to 0,94 Gy/h.
Halflife
27.8 d
ALImin 700 MBq
Radio toxicity C
General properties
decay by photon and beta emisson. The dominating beta energy has a
maximum energy of 752 keV. The dominating gamma energy has a energy of 320
keV.
Detection and measurement
The preferred detector of choice is a liquid scintillation detector, for example a NaIdetector. GM-tubes or similar detectors are very inefficient at detecting
due to
the low photon energies emitted. Wipes measured in liquid scintillation detectors
can be used to detect removable surface contamination.
Radiation protection
Personal radiation monitoring dosimeters should be worn, both whole body and
extremity detector badges. Shielding should always be used. A 5 mm lead plate will
reduce the dose rate significantly. If
enters the body, causing internal
contamination, it enriches in the lower large intestine.
The dose rate, from a 37 MBq point source at 1 cm, is 1.6 Gy/h. At 10 cm the dose
rate has reduced to 0,02 Gy/h.
Halflife
64 h
ALImin 20 MBq
Radio toxicity B
General properties
decay by emission of beta particles. The beta particles have a maximum
energy of 2.27 MeV and an average energy of 0.76 MeV. The range from the beta
particles is approximately 1000 cm in air and 1 cm in tissue.
Detection and measurement
GM-tube
Radiation protection
Dose rate from a 37 MBq point source at 1 cm is 6.77 Gy/h
Halflife
13.3 h
ALImin 100 MBq
Radio toxicity C
General properties
decay by emission of positron and gamma radiation. The beta (positron) particles
42
have a maximum energy of 1.07 MeV
Detection and measurement
A GM-tube, scintillation, and liquid scintillation detector can be used.
Radiation protection
If
enters the body, causing internal contamination, it enriches in the thyroid.
Halflife
60.1 d
Radio toxicity B
General properties
decay to an excited state of
instantaneously to
ALImin 1 MBq
with a probability of 7%, that decays
by emission of a photon with the energy of 35.3 keV.
decay with a probability of 93 % to
by internal conversion and emission of x-
ray gamma photons with a mean energy of 30 kev.
Detection and measurement
Due to the low energetic photons emitted by
GM-tubes become inefficient as
contamination monitors, typically with efficiency around 0.1 %. A low energy
scintillation detector, like NaI-detectors, is the preferred detector of choice. Wipes
of the area of interest can be measured in a liquid scintillation detector.
Radiation protection
gives an external exposure due to photon decay. Shielding should always be
used when handling
as well as distance tools. When there is a risk for airborne
activity, the work should be performed in a fume cupboard.
is hazardous when
it enters the body, causing internal contamination. The critical organ is the thyroid,
where
enriches. This makes even small intakes leads to a high radiation dose to
the organ. An intake of 40 kBq gives a radiation dose to the thyroid of 25 mSv.
The dose rate, from a
Bq point source, at 1 cm is 1.56 – 2.75 Gy/h. At 10 cm
the dose rate has reduced to 0,15-0.27 Gy/h. HVL value of 0.022 mm.
Halflife
8d
ALImin 1 MBq
Radio toxicity B
General properties
decay by emission of beta particles and photons. The beta particles have a
maximum energy of 806 keV and an average energy of 269 keV. The range from
the beta particles is approximately 1.6 m in air and 20 mm in tissue. The three
dominating gamma energies are 248 kev, 364 kev and 637 kev.
Detection and measurement
A GM-tube (or similar detectors) and Liquid scintillation detectors can be used.
Radiation protection
Handling
must be done with great caution since it evaporates. If
enters the
body, causing internal contamination, it enriches in the thyroid.
43
Halflife
7.2 h
Radio toxicity B
General properties
decay by emission of alpha particle and x-ray photons. The energy for the
alpha particle is 5867 keV and the x-ray photons have a mean energy of 92 keV.
Detection and measurement
Liquid scintillation or NaI(Tl) detector can be used.
Radiation protection
does not cause an external radiation dose of radiological concern. The
danger comes from internal contamination. The alpha particles are densely ionizing
and can cause great damage if it enters the body.
Microshield
Volume source definition in Microshield
Waste container
large
small
Dimensions [cm]
42/35/27
46/25/21
(length/width/height)
Dose point [cm] (x/y/z)
43/13.5/1.75 46/10.5/12.5
Integrations (x/y/z)
50/40/40
50/40/40
44
Spread sheet results
Första kontrollen
Datum
Avdelning:
Instrument:
Dosfel: (se tabell)
2008-06-03
Radiofysik
RNI 10/SR
Intensimeter
S/N 59855
20%
Källa nr:
Mätavstånd
1
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,2
1,51
Instrumentet bör ha ett mätutslag som ligger mellan (95 % konfidens intervall)
1,10
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund)
Steg 4 (indikerad dosrat)
Ok
Ok, fixerad i hållare på 10 cm
0,2
1,75
Mätt dosrat (uSv/h)
1,55
och
1,93
Första kontrollen
Datum
Avdelning:
Instrument:
Dosfel: (se manual)
2008-06-03
Radiofysik
RNI 10/SR
Intensimeter
S/N 59857
20%
Källa nr:
Mätavstånd
1
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,21
1,31
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,90
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund)
Steg 4 (indikerad dosrat)
Ok
Ok, fixerad i hållare på 10 cm
0,21
1,55
Mätt dosrat (uSv/h)
1,34
och
1,72
Första kontrollen
Datum
Avdelning:
Instrument:
2008-06-03
MFT
RNI 10/SR
Intensimeter
Dosfel: (se manual)
20%
Källa nr:
Mätavstånd
1
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,10
1,12
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,71
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund)
Steg 4 (indikerad dosrat)
Ok
Ok, fixerad i hållare på 10 cm
0,1
1,25
Mätt dosrat (uSv/h)
1,15
och
1,53
Första kontrollen
Datum
Avdelning:
Instrument:
2008-06-03
Radiofysik
SRV 2000
Dosfel: (se manual)
20%
Källa nr:
Mätavstånd
1
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,21
1,34
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,93
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund) (uSv/h)
Steg 4 (indikerad dosrat)
Ok
Ok, fixerad i hållare på 10 cm
0,15
1,52
Mätt dosrat (uSv/h)
1,37
och
1,75
Första kontrollen
Datum
Avdelning:
Instrument:
Dosfel: (se manual)
2008-06-03
Radiofysik
RNI 10/SR
Intensimeter
S/N 59855
20%
Källa nr:
Mätavstånd
2
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,2
1,19
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,78
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund) (uSv/h)
Steg 4 (indikerad dosrat) (mSv/h)
Ok
Ok, fixerad i hållare på 10 cm
0,21
1,22
Mätt dosrat (mSv/h)
1,22
och
1,60
Första kontrollen
Datum
Avdelning:
Instrument:
Dosfel: (se manual)
2008-06-03
Radiofysik
RNI 10/SR
Intensimeter
S/N 59857
20%
Källa nr:
Mätavstånd
2
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,2
1,13
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,72
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund) (uSv/h)
Steg 4 (indikerad dosrat) (mSv/h)
Ok
Ok, fixerad i hållare på 10 cm
0,26
1,16
Mätt dosrat (mSv/h)
1,16
och
1,54
Första kontrollen
Datum
Avdelning:
Instrument:
2008-06-03
Radiofysik
SRV-2000
Dosfel: (se manual)
20%
Källa nr:
Mätavstånd
2
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,17
1,31
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
0,90
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund) (uSv/h)
Steg 4 (indikerad dosrat) (mSv/h)
Ok
Ok, fixerad i hållare på 10 cm
0,17
1,34
Mätt dosrat (mSv/h)
1,34
och
1,72
Första kontrollen
Datum
Avdelning:
Instrument:
Dosfel: (se manual)
2008-06-03
Radiofysik
Canberra
Radigem
S/N 59855
15%
Källa nr:
Mätavstånd
2
10 cm
Efterföljande kontroll
Datum
2009-06-03
Dagar sedan
första kontrollen:
365
Mätt bakgrund
Sönderfallskorrigerat mätutslag
0,22
1,38
Instrumentet bör ha ett mätutslag som ligger mellan (95% koinf. intervall)
1,05
Steg 1 (OK/EJ OK)
Steg 2 (OK/beskriv fixeringen)
Steg 3 (Indikerad bakgrund) (uSv/h)
Steg 4 (indikerad dosrat) (mSv/h)
Ok
Ok, fixerad i hållare på 10 cm
0,22
1,41
Mätt dosrat (mSv/h)
1,41
och
1,79