Physica Medica: European Journal of Medical Physics - Volume 32 - Supplement 1 - February 2016

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Introduction: The latest Recommendations of the International Commis-
sion on Radiological Protection (ICRP) underlined the fact that age
distributions for workers and general population, for which effective dose
(E) is derived, can be quite different from the overall age distribution for
the patients undergoing medical procedures using ionising radiations. The
purpose of this work was to introduce a methodology for the risk assess-
ment of radiological examinations, using appropriate risk values for the age
and sex distributions of the individuals undergoing medical procedures.
Materials and Methods: Data about frequency and typical doses, relative
to the year 2012, of the most important radiodiagnostic examinations of
Projection Radiography, Computed Tomography and Nuclear Medicine were
collected directly by the 17 public Trusts of the Emilia-Romagna (ER) Re-
gional Health Service. The total number of ER region patients was split into
age and sex distributions, for all the medical procedures taken into account
in the survey.
Results: For each of the 17 public Trusts of ER and for the whole Region
26 cohorts (i.e., 13 age-specific classes for male and female) of patients were
obtained. Doses for each examination were calculated for each cohort. As
an example, during the year 2012 in ER 574 newborn patients (153 males
and 421 females) underwent pelvis and hip radiography, contributing with
3.79 and 5.87 man mSv to the annual collective effective dose, respectively.
The next step will be to associate to those doses the relative risk coeffi-
cients, age- and sex-specific, for evaluating the overall and the analytical
population risk due to medical exposures.
Conclusions: The actual age, sex and dose distributions of patients is a man-
datory step to assess appropriate risk values for medical exposures of the
population, in accordance to the latest ICRP Recommendations. Furthermore,
the large amount of collected data could be useful for retrospective studies.
D.400
DECOMMISSIONING PROCEDURES FOR A 17 MEV MEDICAL CYCLOTRON
L. D’Ambrosio S. Lastoria S. Rossan G. Calicchi C. Frattolill
R. Currarone V. Cerciello P. Luchini R. Calandrino a Istituto Nazionale
Tumori IRCCS Fondazione Pascale, Napoli, Italy; b FIS.ECO srl, Napoli, Italy;
c Progetti Plant, Altavilla Vicentina, Italy; d IRCCS San Raffaele, Milano, Italy
Introduction: The activity of the cyclotron Scanditronix MC17 was stopped
after more than a decade of production of F18. The evaluation of the ac-
tivation levels of the concrete and the different parts of the cyclotron is
mandatory for waste classification and subsequent disposal. This work de-
scribes the procedures followed for the decommissioning of the cyclotron.
Materials and Methods: The first step of decommissioning plan was the
dismantling of the easiest and hottest components from the cyclotron and
the sampling and measurements of the concrete wall to estimate the risk
level for staff involved in subsequent operations. A partial demolition of
the bunker to let exit the cyclotron was carried out. Concretes and cyclo-
tron were stored in an authorized site located in the hospital area. A new
cyclotron was installed in the same bunker.
The sampling of wall was done using concretes coming from the first 25 cm
and from different depth of the wall.
Doses to workers involved in the removal operations were monitored by
individual TL dosimeters. Urine, spit and nasal mucus samples were mea-
sured to estimate internal contamination.
The residual activities of all samples were measured with HpGe detector.
At the moment the sampling of the activated parts of the cyclotron are still
in progress.
Results: Gamma spectrometry of the concrete samples identified 4 radio-
active nuclides due to the activation processes: Eu152, Eu154, Co60, Cs134
with total specific activity of 89 Bq/kg and 9.4 Bq/kg in the superficial and
deep layers. The measured integral dose from external exposure of the
workers was <20 microSv. The effective dose from intake was negligible.
Conclusion: Cyclotron dismantling is a safe procedure. The clearance of
the cyclotron and of the building rubble has to be demonstrated by means
of accurate sampling and measurements of the induced activation. An
average waiting time between 10 and 15 years is expected to achieve the
clearance level of the main parts of the cyclotron.
D.401
DETERMINATION OF ATTENUATION PROPERTIES OF MATERIALS USED
IN PROTECTIVE DEVICES AGAINST DIAGNOSTIC MEDICAL X-RADIATION
FOLLOWING THE NEW CEI EN 61331-1
S. Donatiello *, E. Genovese, C. Orlandi, V. Cannata’. Enterprise Risk
Management/Medical Physics, Bambino Gesù Children’s Hospital, IRCCS, Rome,
Italy
Aim: The use of personal protective equipment (PPE) for the reduction of
the dose absorbed during normal clinical activity is mandatory for all health
care workers who are exposed to ionizing radiation. In particular, in
interventional radiology the operators wear PPE for several minutes. No
longer than few years ago the only material used in PPE was lead (Pb). Now-
adays there are composite materials that have led to the same protection
level with the further advantage of lightening PPE.
In this work, following the IEC 61331-1 protocol, it was investigated whether
the reported Pb equivalent (Pb-eq) material, in the different PPE brands as
used in our hospital IRCCS Pediatric hospital “Bambino Gesù” (OPBG) cor-
responds to the experimental value.
Materials and Methods: The experimental set-up made use of an X-ray
tube of the OPBG images department whose spectrumwas previously char-
acterized. For measurements of kerma rate in air RadCal AGDM Accu-
Gold Digilizer Module with 10 × 6–60/60E Ion Chamber (IC) was used by
employing the method of Inverse Broad Beam Condition as described in
IEC 61331-1 2015-07. Several models and three different thicknesses (0.25,
0.35, 0.5 Pb-eq) were tested to determine the attenuation ratio.
Results: The experimental data showed that the Pb-eq values declared by
the manufacturers comply with the attenuation values of the same thick-
ness of Pb, with small differences between brands.
Conclusions: In this preliminary study, the possibility that a service of
medical physics could perform the appropriate quality assurance on PPE
following the new regulations was investigated. In fact the CEI EN 61331-1
introduces new and more severe procedures for the determination of lead
equivalent inserts by using calculation models different from the CEI EN
61331-1 2003-09. Future evaluations will be carried out with the aid of a
spherical IC for the determination of build-up factors.
D.402
VALIDATION OF A PORTAL MONITOR FOR RADIOACTIVE HOSPITAL
WASTES
O. Ferrando *, R. Bampi, D. Zefiro, M. Baracchini, E. Nuti, A. Chimenz,
F. Foppiano. Medical Physics Department, St. Andrea Hospital, Asl5 Spezzino,
La Spezia, Italy
Aims: The aim of this work concerns the validation of a portal monitor for
detection of radioactivity in trash leaving the hospital facility. As required
by the Italian law hospital radioactive wastes can be delivered for dispos-
al if the activity fall down the limit of 1 Bq/g (radioisotopes with T1/
2< 75 days). Installation and validation of a system to detect trash containing
activities upper this limit is essential to guarantee radioprotection safety.
Materials and Methods: The monitor for detection of radioactivity in hos-
pital trash is composed by two 2 meter-scintillators placed at the exit of
the bunker containing the trash plastic boxes. The detectors are equipped
with a module for visualization of the source counts (cps) and acoustic alarm
signal. To validate the system the following measurement were per-
formed: natural background measurement, MDA (minimum detectable
activity) for all radioisotopes used in clinical practice, system response versus
radioisotope activities, dependence of the system by the source geometry.
Results: Radioactive background is defined in the range of (3.4 ± 0.5) Kcps.
Sensitivity detector depends on radioisotope and MDA for 131I, 18F, 99mTc
are respectively 81 kBq 400 kBq and 65 kBq. System linearity was con-
firmed for all radioisotopes. The alarm threshold is set to 4.5 Kcps
corresponding to 131I MDA. Counting detection depends on the source po-
sition respect to the scintillators but the system detection capability remains
unchanged.
Conclusions: From the installation in May 2015 up to now the portal
monitor checked about 22,400 plastic boxes and 20 of these were found
to contain radioactive materials over the alarm threshold. No failures oc-
curred in this period. However since radioisotope activity lower than MDA
e118 Abstracts/Physica Medica 32 (2016) e116–e123