Although, at present, neither Italian legislation nor technical protocols require that personal dosimetry is performed to
assess Hp(d), the ENEA Individual Monitoring Service (IMS) is able to supply thermoluminescence (TL) whole-body and extremity dosemeters
for photon and beta fields, based on LiF(Mg,Cu,P) detectors and these have been fully developed at the ENEA Institute for
Radiation Protection (IRP). All irradiation tests have been performed with ISO phantoms and ISO recommended reference radiations
at the ENEA-IRP Secondary Standard Dosimetry Laboratory. The whole-body dosemeter contains two LiF(Mg,Cu,P) (GR200) detectors
that are filtered differently. One is filtered on both sides by 290 mg.cm-2 mass per area (270 mg.cm-2 Al + 20 mg.cm-2 plastic protective layer); the other is filtered on both sides by a plastic layer of 20 mg.cm-2 mass per area. In photon radiation fields, the maximum uncertainty due to the energy dependence of the response, is ±4% for
Hp(0.07) in the energy range 13 keV to 202 keV, and ±15% for Hp(10) in the range 13 keV to 1.25 MeV. The dosemeter response in terms of Hp(d,a) in beta fields has been investigated recently. The results of a EURADOS trial performance test for photon and beta fields
are reported and discussed in this paper. The extremity dosemeter currently used at ENEA IMS consists of a GR200 detector
glued on a kapton strip identified by a bar code. Its response in terms of Hp(0.07,a) has been measured recently and the results are given. Moreover, different dosemeter assemblies have been tested to
compare the performances in photon and beta fields. Therefore, the following three constructions have been prepared: (1) an
MCP-Ns (8.5 mg.cm-2 mass per area) detector with a Mylar filter of 0.5 mg.cm-2 mass per area; (2) a polyethylene filter of 12 mg.cm-2 mass per area; and (3) a GR200 (210 mg.cm-2 mass per area) detector with a Mylar filter of 0.5 mg.cm-2. Finally, a brief discussion on international and Italian requirements for personal monitoring is given.
The pulse height weighting integration method for measuring H'(0.07) using a plastic scintillator has been developed through Monte Carlo simulations and laboratory tests. In order to design an optimal detector configuration, the parameters of input-window thickness and plastic scintillator thickness were investigated by EGS4 for their influences on the energy response for electrons. Based on the calculated deposited spectra in the plastic scintillator for electrons with various incident energies, a weighting function W(E), folded with the deposited energy spectra, was introduced in order to obtain a flat energy response in the low energy range. A constant response within +/-3% was confirmed, by calculation, for electrons with energy >0.15 MeV. In addition, a preliminary experiment was performed using three beta sources (90Sr-90Y, 147Pm, 204Tl) and the energy response within +/-6% for beta rays with the maximum energy >0.22 MeV was obtained. Some factors causing uncertainties in the measurements are also discussed in this article.
In recent years, several papers dealing with eye lens dosimetry have been published as epidemiological studies are implying
that the induction of cataracts occurs even at eye lens doses of less than 500 mGy. For that reason, the necessity to monitor
the eye lens may become more important than it was before. However, only few dosemeters for the appropriate quantity Hp(3) are available. Partial-body dosemeters are usually designed to measure the quantity Hp(0.07) calibrated on a rod phantom representing a finger while a slab phantom much better represents the head. Therefore,
in this work it was investigated whether dosemeters designed for the quantity Hp(0.07) calibrated on a rod phantom can also be worn on the head (close to the eyes) and still deliver correct results (Hp(0.07) on a head). For that purpose, different types of partial-body dosemeters from routine use were irradiated at different
photon energies on both a rod and a slab phantom. It turned out that their response values are within ±5% independent of the
phantom if the quantity value for the respective phantom is used. Thus, partial-body dosemeters designed for the quantity
Hp(0.07) calibrated on a rod phantom may be worn on the head and used to monitor the eye lens dose due to photon radiation via
the measurement of Hp(0.07) on the head.
X and gamma rays continue to remain the main contributors to the dose to humans. As these photons of varying energies are encountered in various applications, the study of photon energy response of a dosemeter is an important aspect to ensure the accuracy in dose measurement. Responses of dosemeters have to be experimentally established because for luminescence dosemeters, they depend not only on the effective atomic number (ratio of mass energy absorption coefficients of dosemeter and tissue) of the detector, but also considerably on the luminescence efficiency and the material surrounding the dosemeters. Metal filters are generally used for the compensation of energy dependence below 200 keV and/or to provide photon energy discrimination. It is noted that the contribution to Hp(0.07) could be measured more accurately than Hp(10). For the dosemeters exhibiting high photon energy-dependent response, estimation of the beta component of Hp(0.07) becomes very difficult in the mixed field of beta radiation and photons of energy less than 100 keV. Recent studies have shown that the thickness and the atomic number of metal filters not only affect the response below 200 keV but also cause a significant over-response for high energy (>6 MeV) photons often encountered in the environments of pressurised heavy water reactors and accelerators.
In August 2009, almost 1000 passive extremity dosemeters were irradiated at the Dosimetry Laboratory Seibersdorf as part of
the EURADOS intercomparison IC2009. Forty-four European individual monitoring services participated, with a total of 59 dosimetry
systems (46 finger ring, 4 finger tip and 9 wrist/ankle dosemeter systems). Additionally, finger-ring dosemeters from the
Dosimetry Service Seibersdorf were irradiated in a non-competitive manner. Dosemeter irradiations on rod and pillar phantoms
in four photon-radiation fields complying with the ISO standard 4037 were performed with personal dose equivalent values (Hp(0.07)) ranging from 4 to 480 mSv. Traceability was established by using an air-kerma-calibrated monitor ionisation chamber
together with the X-ray facility as well as a calibrated 137Cs gamma radiation field with a collimated beam geometry. The ISO-tabulated conversion coefficients from air kerma free-in-air
to Hp(0.07) were applied, resulting in the main contribution to the expanded measurement uncertainties.
Information about the size distribution of radioactive aerosols in nanometre range is essential for the purposes of air contamination
monitoring, dose assessment to respiratory tract and planning of protective measures. The diffusion battery, which allows
capturing particles in the size range of 0.1–10 nm, has developed. Interpreting data obtained from diffusion battery is very
complex. The method of expectation maximisation by Maher and Laird was chosen for indirect inversion data. The experiments
were performed in the box with equivalent equilibrium concentration of radon in the range of 7000–10 000 Bq m−3. The three modes of size distribution of radon decay products aerosols were obtained: activity median thermodynamic diameter
(AMTD) 0.3, 1.5 and 5 nm. These modes can be identified as: AMTD 0.3 nm—atoms of radon progeny (218Po in general); AMTD 1.5 nm—clusters of radon progeny atoms and non-radioactive atoms in the atmosphere; AMTD 5 nm—particles
formed by coagulation of previous mode clusters with existing aerosol particles or nucleation of condensation nuclei containing
atoms of radon progeny.
The FLUKA Monte Carlo simulations are carried out to estimate the 41Ar concentration inside accelerator vaults of various sizes when proton beams of energy 0.1–1.0 GeV are incident on thick
copper and lead targets. Generally 41Ar concentration is estimated using an empirical formula suggested in the NCRP 144, which assumes the activation is caused
only by thermal neutrons alone. It is found that while the analytical and Monte Carlo techniques give similar results for
the thermal neutron fluence inside the vault, the 41Ar concentration is under-predicted by the empirical formula. It is also found that the thermal neutrons contribute ∼41 %
to the total 41Ar production while 56 % production is caused by neutrons between 0.025 and 1 eV. A modified factor is suggested for the use
in the empirical expression to estimate the 41Ar activity 0.1–1.0-GeV proton accelerator enclosures.
Energy distributions of secondary charged particles were calculated in tissue substitutes irradiated by neutrons from 0.14
to 65 MeV, using the Particle and Heavy Ion Transport code System. The calculations were compared with experimental data measured
by tissue equivalent proportional counters (TEPC). It is found that the calculated distributions of the lineal energy, y, generally agree well with the measured ones for neutrons from several 100 keV to 15 MeV. In the case of 40 and 65 MeV neutron
irradiations, wall effects of TEPC should be considered and the fluence of alphas is underestimated by the calculations. Integrated
dose contributions of the secondary charged particles are generally in good agreement with those of the measured ones.
In conformal moving beam therapy with fast neutrons, the contributions to dose from the direct beam, scattered radiation and the gamma component vary with the position in the phantom. To determine this variation in radiation quality, microdosimetric measurements of energy deposition spectra were performed at different position in a therapy phantom. Fixed beam irradiation at different incidence angles showed strong changes in the lineal energy spectrum. An increase of slow protons (20 < y < 110 keV.micron-1) and a decrease of fast protons (2 < y < 20 keV.micron-1) was seen for irradiation outside the direct beam. During moving beam irradiation, different positions on the same isodose curves (55% or 35%) showed differences in YD of up to 5%. Variations in the quality parameter, R, determined by applying an empirical biological weighting function, were of similar magnitude. Thus, spatial variations in radiation quality should be taken into account in biological dose planning for moving beam neutron therapy.
The neoplastic transformation of human hybrid CGL1 cells is affected by perturbations from external influences such as serum
batch and concentration, the number of medium changes during the 21-day expression period and cell seeding density. Nevertheless,
for doses up to 1.5 Gy, published transformation frequencies for low linear energy transfer (LET) radiations (γ-rays, MeV
electrons or photons) are in good agreement, whereas for higher doses larger variations are reported. The 60Co γ-ray data here for doses up to 1.5 Gy, using a low-yield serum batch and only one medium change, are in agreement with
published frequencies of neoplastic transformation of human hybrid cells. For 3.4 MeV α-particles (LET = 124 keV/μm) and 0.565
MeV monoenergetic neutrons relative to low doses of 60Co γ-rays, a maximum relative biological effectiveness (RBEM) of 2.8 ± 0.2 and 1.5 ± 0.2, respectively, was calculated. Surprisingly, at higher doses of 60Co γ-rays lower frequencies of neoplastic transformation were observed. This non-monotonic dose relationship for neoplastic
transformation by 60Co γ-rays is likely due to the lack of a G2/M arrest observed at low doses resulting in higher transformation frequencies
per dose, whereas the lower frequencies per dose observed for higher doses are likely related to the induction of a G2/M arrest.
This whole body donation case (USTUR Registrant) involved two suspected PuO2 inhalation intakes, each indicated by a measurable Pu alpha activity in a single urine sample, followed about 1(1/2) y later by a puncture wound to the thumb while working in a Pu glovebox. The study is concerned with modelling simultaneously the biokinetics of deposition and retention in the respiratory tract and at the wound site; and the biokinetics of Pu subsequently transferred to other body organs, until the donor's death. Urine samples taken after the wound incident had readily measurable Pu alpha activity over the next 14 y, before dropping below the minimum detectable excretion rate (<0.4 mBq d(-1)). The Registrant died about 33 y after the wound intake, at the age of 71, from hepatocellular carcinoma with extensive metastases. At autopsy, all major soft tissue organs were harvested for analysis of their 238Pu, 239+240Pu and 241Am content. The amount of 239+240Pu retained at the wound site was 68 +/- 7 Bq (1 SD), measured by low-energy planar Ge spectrometry. A further 56.0 +/- 1.2 Bq was retained in an associated axillary lymph node, measured by radiochemistry. Simultaneous mathematical analysis (modelling) of all in vivo urinary excretion data, together with the measured lung, thoracic lymph node, wound, axillary lymph node and systemic tissue contents at death, yielded estimated intake amounts of 757 and 1504 Bq, respectively, for the first and second inhalation incidents, and 204 Bq for the total wound intake. The inhaled Pu material was highly insoluble, with an estimated long-term absorption rate from the lungs of 2 x 10(-5) d(-1). The Pu material deposited at the wound site was mixed: approximately 14% was rapidly absorbed, approximately 49% was absorbed at the rate of about 6 x 10(-5) d(-1), and the remainder ( approximately 37%) was absorbed extremely slowly (at the rate of about 5 x 10(-6) d(-1)). Thus, it was estimated that only approximately 40% of the Pu initially deposited in the wound had been absorbed systemically over the 33-y period until the donor's death. The biokinetic modelling also indicated that, in this individual case, some of the parameter values (rate constants) incorporated in the ICRP Publication 67 Pu model were up to a factor of 2 different from ICRP's recommended values (for reference man).
This whole body donation case (USTUR Registrant) involved a single acute inhalation of an acidic Pu(NO3)4 solution in the form of an aerosol 'mist'. Chelation treatment with intravenously (i.v.) Ca-EDTA was initiated on the day of the intake, and continued intermittently over 6 months. After 2.5 y with no further treatment, a course of i.v. Ca-DTPA was administered. A total of 400 measurements of 239+240Pu excreted in urine were recorded; starting on the first day (both before and during the initial Ca-EDTA chelation) and continuing for 37 y. This sampling included all intervals of chelation. In addition, 91 measurements of 239+240Pu-in-feces were recorded over this whole period. The Registrant died about 38 y after the intake, at age 79 y, with extensive carcinomatosis secondary to adenocarcinoma of the prostate gland. At autopsy, all major soft tissue organs were harvested for radiochemical analyses of their 238Pu, 239+240Pu and 241Am content. Also, all types of bone (comprising about half the skeleton) were harvested for radiochemical analyses, as well as samples of skin, subcutaneous fat and muscle. This comprehensive data set has been applied to derive 'chelation-enhanced' transfer rates in the ICRP Publication 67 plutonium biokinetic model, representing the behaviour of blood-borne and tissue-incorporated plutonium during intervals of therapy. The resulting model of the separate effects of i.v. Ca-EDTA and Ca-DTPA chelation shows that the therapy administered in this case succeeded in reducing substantially the long-term burden of plutonium in all body organs, except for the lungs. The calculated reductions in organ content at the time of death are approximately 40% for the liver, 60% for other soft tissues (muscle, skin, glands, etc.), 50% for the kidneys and 50% for the skeleton. Essentially, all of the substantial reduction in skeletal burden occurred in trabecular bone. This modelling exercise demonstrated that 3-y-delayed Ca-DTPA therapy was as effective as promptly administered Ca-EDTA.
The Soviet-produced KDT-02M system, which is still widely used for dosimetric monitoring in countries of the former Soviet Union, was compared with the Harshaw 8800 system. The comparison consisted of two stages. In the first stage workplace radiation fields were simulated in the framework of the IAEA intercomparison. In the second stage the two systems were compared when used in parallel by the personnel of Chernobyl Object 'Shelter'. Although in the first stage the Harshaw 8800 demonstrated better performance for various irradiation conditions, an obsolete KDT-02M also proved compliance with the basic requirements to the accuracy of individual dosimetric monitoring. In the second stage, more than 1200 paired measurements were performed, revealing good (r = 0.95) correlation between readouts of both systems. Deviation of the slope of the regression line may be adjusted by proper calibration. Although the KDT-02M system demonstrated adequate results for measurement of deep dose equivalent, its inability to determine shallow dose equivalent calls for its replacement with modem thermoluminesence dosemeters possessing this feature.
The thresholds of (n,xn) reactions in various activation detectors are commonly used to unfold the neutron spectra covering a broad energy span, i.e. from thermal to several hundreds of MeV. The saturation activities of the daughter nuclides (i.e. reaction products) serve as the input data of specific spectra unfolding codes, such as SAND-II and LOUHI-83. However, most spectra unfolding codes, including the above, require an a priori (guess) spectrum to starting up the unfolding procedure of an unknown spectrum. The accuracy and exactness of the resulting spectrum primarily depends on the subjectively chosen guess spectrum. On the other hand, the Genetic Algorithm (GA)-based spectra unfolding technique ANDI-03 (Activation-detector Neutron DIfferentiation) presented in this report does not require a specific starting parameter. The GA is a robust problem-solving tool, which emulates the Darwinian Theory of Evolution prevailing in the realm of biological world and is ideally suited to optimise complex objective functions globally in a large multidimensional solution space. The activation data of the 27Al(n,alpha)24Na, 116In(n,gamma)116mIn, 12C(n,2n)11C and 209Bi(n,xn)(210-x)Bi reactions recorded at the high-energy neutron field of the ISIS Spallation source (Rutherford Appleton Laboratory, UK) was obtained from literature and by applying the ANDI-03 GA tool, these data were used to unfold the neutron spectra. The total neutron fluence derived from the neutron spectrum unfolded using GA technique (ANDI-03) agreed within +/-6.9% (at shield top level) and +/-27.2% (behind a 60 cm thick concrete shield) with the same unfolded with the SAND-II code.
Anthropomorphic computational models coupled with radiation transport codes are valuable tools in radiation protection dosimetry. In particular, they are very reliable for the estimate of the energy absorbed by different organs due to an incorporated radionuclide. MIRD-based stylised analytical models are widely accepted as standards but the recent generation of voxel phantoms, developed on real anatomical data derived from tomographic images, can represent a valid alternative for radiation protection and dosimetry purposes. Specific absorbed fraction evaluation and patient-specific dose estimate in nuclear medicine and radiotherapy could be considered as the optimal area for their implementation and use. On the other hand, the accuracy of organ and body structure representation guarantees an improved dose evaluation system also for radiation protection purposes in the workplace in case of accidental internal contamination. In the present work the voxel model NORMAN-05, a modified version of NORMAN (HPA, UK) model, has been employed with the Monte Carlo code MCNPX. Some preliminary investigations were carried out to evaluate the absorbed fractions for a series of source-target organ couples in case of gamma emitters and the organ absorbed doses in case of 90Sr incorporation. The paper summarises the main preliminary outcomes of such studies.
Photons conversion coefficients from 20 keV to 100 MeV have been calculated with the voxel model NORMAN-05 using the MCNPX
code. Both kerma approximation and electronic transport were employed and the results compared with published data. In the
near future, ICRP group DOCAL will issue a new computational model, based on the GSF—GOLEM, which is intended to be the reference
adult male voxel model for ICRP. NORMAN-05 well approximates the western-caucasian standard man characteristics, in terms
of body height (176 cm) and mass (73 kg) and masses of the organs. It is not intended to substitute, or to be an alternative,
to the future official ICRP voxel model, but thanks to its accuracy and its ‘standard man structure’ can be useful to evaluate
the intrinsic uncertainties associated with the dose quantities evaluated adopting different voxel models. For such reason
data obtained with NORMAN-05 could be easily compared with those that will be derived from the future ICRP model.
Occupational exposure to radiation in medical practice in Ghana has been analysed for a 10-y period between 2000 and 2009. Monitored dose data in the medical institution in Ghana from the Radiation Protection Institute's database were extracted and analysed in terms of three categories: diagnostic radiology, radiotherapy and nuclear medicine. One hundred and eighty medical facilities were monitored for the 10-y period, out of which ~98% were diagnostic radiology facilities. Only one nuclear medicine and two radiotherapy facilities have been operational in the country since 2000. During the 10-y study period, monitored medical facilities increased by 18.8%, while the exposed workers decreased by 23.0%. Average exposed worker per entire medical institution for the 10-y study period was 4.3. Annual collective dose received by all the exposed workers reduced by a factor of 4 between 2000 and 2009. This is seen as reduction in annual collective doses in diagnostic radiology, radiotherapy and nuclear medicine facilities by ~76, ~72 and ~55%, respectively, for the 10-y period. Highest annual collective dose of 601.2 man mSv was recorded in 2002 and the least of 142.6 man mSv was recorded in 2009. Annual average values for dose per institution and dose per exposed worker decreased by 79 and 67.6%, respectively between 2000 and 2009. Average dose per exposed worker for the 10-y period was least in radiotherapy and highest in diagnostic radiology with values 0.14 and 1.05 mSv, respectively. Nuclear medicine however recorded average dose per worker of 0.72 mSv. Correspondingly, range of average effective doses within the diagnostic radiology, radiotherapy and nuclear medicine facilities were 0.328-2.614, 0.383-0.728 and 0.448-0.695 mSv, respectively. Throughout the study period, an average dose per medical institution of 3 mSv and an average dose per exposed worker of 0.69 mSv were realised. Exposed workers in diagnostic radiology primarily received most of the individual annual doses >1 mSv. The entire study period had 705 instances in which exposed workers received individual annual doses >1 mSv. On thermoluminescent dosemeter (TLD) return rates, facilities in Volta and Eastern Regions recorded highest return rates of 94.3% each. Ashanti Region recorded the least TLD return rate with 76.7%.
Institutions in the education, research and industrial sectors in Ghana are quite few in comparison to the medical sector. Occupational exposure to radiation in the education, research and industrial sectors in Ghana have been analysed for a 10 y period between 2000 and 2009, by extracting dose data from the database of the Radiation Protection Institute, Ghana Atomic Energy Commission. Thirty-four institutions belonging to the three sectors were monitored out of which ∼65% were in the industrial sector. During the 10 y study period, monitored institutions ranged from 18 to 23 while the exposed workers ranged from 246 to 156 between 2000 and 2009. Annual collective doses received by all the exposed workers reduced by a factor of 2 between 2000 and 2009. This is seen as a reduction in annual collective doses in education/research and industrial sectors by ∼39 and ∼62%, respectively, for the 10 y period. Highest and least annual collective doses of 182.0 man mSv and 68.5 man mSv were all recorded in the industrial sector in 2000 and 2009, respectively. Annual average values for dose per institution and dose per exposed worker decreased by 49 and 42.9%, respectively, between 2000 and 2009. Average dose per exposed worker for the 10 y period was least in the industrial sector and highest in the education/research sector with values 0.6 and 3.7 mSv, respectively. The mean of the ratio of annual occupationally exposed worker (OEW) doses for the industrial sector to the annual OEW doses for the education/research sector was 0.67, a suggestion that radiation protection practices are better in the industrial sector than they are in the education/research sector. Range of institutional average effective doses within the education/research and industrial sectors were 0.059-6.029, and 0.110-2.945 mSv, respectively. An average dose per all three sectors of 11.87 mSv and an average dose per exposed worker of 1.12 mSv were realised for the entire study period. The entire study period had 187 instances in which exposed workers received individual annual doses >1 mSv, with exposed workers in the education/research sector primarily receiving most of this individual dose.
This study aimed to assess the efficacy of 3,4,3-LI(1,2-HOPO) for reducing uranium, plutonium and americium in rats after
intramuscular injection of (U-Pu)O2 particles (MOX). Sixteen rats were contaminated by intramuscular injection of a 1 mg MOX suspension and then treated daily
for 7 d with LIHOPO (30 or 200 µmol kg-1) or DTPA (30 µmol kg-1). LIHOPO was inefficient for removing Pu, Am and U from the wound site. However, it reduced Pu retention in carcass and liver
by factors of 2 and 6 respectively, and Am retention in carcass and liver by factors of 10 and 30. In contrast, the effect
of LIHOPO on U was to decrease the retention in kidneys by a factor of 75. These results confirm that LIHOPO is a good candidate
for use after contamination with MOX, in combination with localised wound lavage or surgical treatment aimed at removing most
of the contaminant at the wound site.
The energy responses for the KLT-300(LiF:Mg,Cu,Na,Si, Korea), GR-200(LiF:Mg,Cu,P, China) and MCP-N(LiF:Mg,Cu,P, Poland) thermoluminescence(TL) pellets were studied for a photon radiation with energies from 1.25 MeV(60Co) to 21 MV (Microtron) to verify the usefulness of the calibration for the radiotherapy beams. The International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) have performed thermoluminescence dosimetry (TLD) audits to verify the calibration of the beams by TL powder, but TL pellets were used in this study because the element correction factor (ECF), defined as the factor to correct the variations that all TL dosemeters cannot be manufactured to have exactly the same TL efficiency, for each TL pellet could be accurately derived and be handled conveniently when compared with the powder. Also several works for the energy response of the TLDs were done for the low-energy photon beams up to 60Co, but they will be extended in this experiment to the high photon energies (up to 20 MV), which are widely used in the therapy level of a radiation. The PTW 30006 ionisation chamber was calibrated by the Korea primary standards to establish the air-kerma rates and the TL pellets were irradiated in a specially designed waterproof pellet holder in a water phantom (30 x 30 x 30 cm3) just like the IAEA postal audits programme. This result was compared with that of another type of phantom [10 (W) x 10 (L) x 10 (H) cm3 PMMA Perspex phantom for the 60Co and 6 MV photon, and 10 x 10 x 20 (H) cm3 for the 10 and 21 MV photon] for its convenient use and easy handling and installation in a hospital. The results show that the differences of the responses for the water phantom and PMMA Perspex phantom were negligible, which is contrary to the general conception that a big difference would be expected. For an application of these results to verify the therapy beams, an appropriate energy correction factor should be applied to the energies and phantom types in use.
The authors examine the radiation field produced in the vicinity of the main beam dump of the FERMI free-electron laser under
the impact of a 1.4-GeV electron beam. Electromagnetic and neutron dose rates are calculated with the Fluka Monte Carlo code
and compared with ionisation chamber and superheated drop detector measurements in various positions around the dump. Experimental
data and simulation results are in good agreement with a maximum deviation of 25 % in a single location.
Alanine response to low-energy protons was studied with alanine dosemeters of 2 mm thickness, irradiated with proton beams of energy in the 1.6-6.1 MeV range. The detector's range-averaged relative effectiveness to 60Co radiation ranged from 0.61 to 0.65. For fluence values up to 5 x 10(10) protons x cm(-2), the alanine response was linear.
Designing Air Cleaning Units (ACU) of an Engineered Safety Feature and normal atmosphere clean-up system at the renovated
APR-1000 and APR-1400 NPP, and fuel cycle facilities in Korea, is required to meet the standards of ASME AG-1 (1997), ASME
N509/N510 (1989) and KEPIC-MH (2001) to enhance the removal efficiency of aerosols and particulates from the effluents. The
revised ACU testing criteria are allowed to use alternative challenge agents of the dioctyl phthalate and Refrigerant-11 for
in situ testing of high efficiency particulate air filters and adsorption banks. The operability testing time of engineered safety
feature (ESF) trains was changed from 10 h to 15 min. The activated carbon in adsorption banks should undergo laboratory tests
at a temperature of 30°C and relative humidity 95 %. The removal criteria of methyl iodide should be over 99.5 % for ESF and
99 % for normal systems. This paper provides the background of the changed criteria for designing and testing of the ACU system
in nuclear facilities.
The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.
The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals
and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated
neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter
stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface
and 1/4 thickness but a good fitting for deeper points in vessel.
It is well known that spurious signals can occur in thermoluminescence dosimetry (TLD) whenever contaminants (i.e. dirt, oil, dust) are present on the surface of the TLD card or crystal during the read-out process. For TLD cards, the Teflon material can also contribute to the background noise and this contribution has been found to depend on the material's light absorption. These non-radiation-induced signals contribute to the total light output during TLD read-out and can lead to incorrect dosimetry especially for low-dose measurements such as personal dosimetry. However, these spurious signals are generally in the low-temperature channels and are mostly accompanied by abnormal glow curves. Most of the published reports dealing with this type of spurious TL signal are on the LiF:Mg,Ti (TLD-100) material. The relatively new TLD material, LiF:Mg,Cu,P, is more sensitive and has higher signal-to-noise ratio than the traditional LiF:Mg,Ti. In this study, the effects of disturbing signals to the LiF:Mg,Cu,P (TLD-100H) cards used in personal dosimetry are investigated and compared with those of LiF:Mg,Ti (TLD-100).
This paper describes further investigations and results in the reduction of residual signal in Harshaw TLD-100H, 600H and
700H (LiF:Mg,Cu,P). TLD-100H is an advanced, relatively new dosimetric material with near tissue-equivalence, flat energy
response, and the ability to measure beta, photon and, more importantly, neutrons all from the same base material. The simple
glow curve structure provides insignificant fade over extended dosimetric periods of up to 1 y. A criticism of the material
has been the residual as compared with TLD-100 (LiF:Mg,Ti). We will show how high-temperature peaks cause the residual signal.
We will also show how the various parameters and conditions of the residual measurement technique as well as the configuration
of the sample affect the residual measurements. A brief description of the experimental paths taken during our investigation
will be presented. We will show how we have reduced the high-temperature peaks of LiF:Mg,Cu,P in our manufacturing process
while not affecting other dosimetric properties of this material. The improvements and material properties that need to be
shown have been incorporated into our production processes. LiF:Mg,Cu,P stands as a premier choice for personal dosimetry
and has been integrated into personal, environmental and extremity configurations of the Harshaw TLD family-line of products.
One of the advantages of LiF based thermoluminescent (TL) materials is its tissue-equivalent property. The Harshaw TLD-100H
(LiF:Mg,Cu,P) material has demonstrated that it has a near-flat photon energy response and high sensitivity. With the optimized
dosemeter filters built into the holder, the Harshaw TLD-100H two-element dosemeter can be used as a whole body personnel
dosemeter for gamma, X ray and beta monitoring without the use of an algorithm or correction factor. This paper presents the
dose performance of the Harshaw TLD-100H two-element dosemeter against the ANSI N13.11-2001 standard and the results of tests
that are required in IEC 1066 International Standard.
The impact of the introduction of ICRP Publication 103 on neutron dosimetry was analysed by calculating effective doses in
various operational neutron fields, using dose conversion coefficients derived from the recommendations given in ICRP 103
and ICRP 60. It was found from the analysis that effective doses based on ICRP 103 are generally smaller than those based
on ICRP 60, mainly owing to the revision of wR assigned to neutrons. The results also indicate that H*(10) can provide a conservative estimate for ICRP 103-based effective doses in most neutron fields. These tendencies suggest
that the radiological protection system currently adopted in accelerators and nuclear facilities can be maintained after the
introduction of ICRP 103, with respect to neutron dosimetry.
Both 125I and 103Pd sources have been widely used in the permanent prostate implant. An important consideration for the choice of brachytherapy
sources is the relative biological effectiveness (RBE) for the source/seed used in the implantation. As RBE is closely related
to the microdosimetric parameter, it is desirable to calculate the dose mean lineal energies for both 125I and 103Pd at various radial distances to the seed surface. Monte Carlo simulation was performed for photons emitted from 125I and 103Pd. Energy depositions from photons and all their secondary electrons were tracked. Dose distributions of lineal energy, d(y), were calculated for spheres of 1 µm in diameter and at various radial distances to the seed surface. From the dose distribution
of lineal energy, the dose mean lineal energy, yD, was derived. The results showed that the radiation qualities are constant in the distance range from 0.5 to 5 cm. In this
distance range, the quality factor, relative to gamma rays from 60Co, is 2.2 for 125I and 2.5 for 103Pd.
In diagnostic nuclear medicine, mean absorbed doses to patients' organs and effective doses are published for standard stylised anatomic models. To provide more realistic and detailed geometries of the human morphology, the International Commission on Radiological Protection (ICRP) has recently adopted male and female voxel phantoms to represent the reference adult. This work investigates the impact of the use of these new computational phantoms. The absorbed doses were calculated for 11 different radiopharmaceuticals currently used in diagnostic nuclear medicine. They were calculated for the ICRP 110 reference computational phantoms using the OEDIPE software and the MCNP extended Monte Carlo code. The biokinetic models were issued from ICRP Publications 53, 80 and 106. The results were then compared with published values given in these ICRP Publications. To discriminate the effect of anatomical differences on organ doses from the effect of the calculation method, the Monte Carlo calculations were repeated for the reference adult stylised phantom. The voxel effect, the influence of the use of different densities and nuclear decay data were also investigated. Effective doses were determined for the ICRP 110 adult reference computational phantom with the tissue weighting factor of ICRP Publication 60 and the tissue weighting factors of ICRP Publication 103. The calculation method and, in particular, the simulation of the electron transport have a significant influence on the calculated doses, especially, for small and walled organs. Overestimates of >200 % were observed for the urinary bladder wall of the stylised phantom compared with the computational phantoms. The unrealistic organ topology of the stylised phantom leads to important dose differences, sometimes by an order of magnitude. The effective doses calculated using the new computational phantoms and the new tissue weighting factors are globally lower than the published ones, except for some radiopharmaceuticals, where the differences can reach 60 % higher than the published values. This study analyses the first set of absorbed and effective doses with the new ICRP male and female reference computational phantoms for different radiopharmaceuticals. It highlights the importance of taking into account the electron transport and the realism of the shape and inter-organ distances of the anthropomorphic model used.
The objective of this study was to perform comparative dosimetric studies of both 106Ru/106Rh plaque brachytherapy and external beam proton therapy proposed for ocular treatments at the University of Texas M. D. Anderson
Cancer Center, Houston, TX, USA. These modalities were also compared with traditional 125I plaque brachytherapy. Using a standardised eye model with a representative ocular melanoma tumour, the relative dose distributions
within the tumour and surrounding tissue were calculated using the Monte Carlo code MCNPX. Published absorbed dose distributions
benchmarked the Monte Carlo models. Results indicate that the proton beam provided superior dose uniformity within the tumour
volume, whereas the dose distribution from 106Ru/106Rh was more heterogeneous. Relative to 125I COMS plaque, both 106Ru/106Rh and protons have shown more confined dose distributions to the tumour volume in this situation, thus sparing other critical
ocular structures. For protons, it has been shown that only doses lower than the maximum dose are delivered outside the tumour
volume. Depending on the clinical situation, this may aid in the sparing of critical structures located in the sclera and
optic disc boundary. The Monte Carlo model's statistical uncertainties of the mean dose estimates for the 106Ru/106Rh plaque and proton beam were 3 and 2.5%, respectively.
This study is a part of a programme of research to provide validated dose measurement and calculation techniques for beta
emitting hot particles by the construction of well-defined model hot particle sources. This enables parallel measurements
and calculations to be critically compared. This particular study concentrates on the high-energy beta emitter, 106Ru/106Rh (Emax = 3.54 MeV). This source is a common constituent of failed nuclear fuel, particularly in accident situations. The depth dose
distributions were measured using radiochromic dye film (RDF); an imaging photon detector coupled to an LiF thermoluminescent
dosemeter (LiF-IPD) and an extrapolation ionisation chamber (ECH). Dose calculations were performed using the Monte Carlo
radiation transport code MCNP4C. Doses were measured and calculated as average values over various areas and depths. Of particular
interest are the doses at depths of 7 and 30–50 mg cm−2, and averaged over an area of 1 cm2, as recommended by the International Commission on Radiological Protection for use in routine and accidental over-exposures
of the skin. In this case, the average ratios (MCNP/measurement) for RDF, ECH and LiF-IPD were 1.07 ± 0.02, 1.02 ± 0.01 and
0.83 ± 0.16, respectively. There are significantly greater discrepancies between the ECH and LiF-IPD measurement techniques
and calculations—particularly for shallow depths and small averaging areas.
The aim of this paper is to describe the dosimetric evaluation of a point contamination that occurred in a laboratory during
the examination of an irradiated sample. The incident led to point contamination of the operator's finger due to the presence
of mainly 106Ru, with its progeny 106Rh. The paper reports on the activity and dose assessment, performed using several methods. The measured activity was obtained
using a conventional device based on a germanium detector and confirmed using software developed at IRSN, based on reconstruction
of voxel phantom associated with the Monte Carlo N-Particle code (MCNP) for in vivo measurement. Two dose assessment calculations were performed using both analytical and Monte Carlo methods, applying the
same approach as for activity assessment based on the personal computational phantom of the finger. The results are compared,
followed by a discussion on the suitability of the tools described in this study.
The impact a revision of nuclear decay data had on dose coefficients was studied using data newly published in ICRP Publication
107 (ICRP 107) and existing data from ICRP Publication 38 (ICRP 38). Committed effective dose coefficients for occupational
inhalation of radionuclides were calculated using two sets of decay data with the dose and risk calculation software DCAL
for 90 elements, 774 nuclides and 1572 cases. The dose coefficients based on ICRP 107 increased by over 10 % compared with
those based on ICRP 38 in 98 cases, and decreased by over 10 % in 54 cases. It was found that the differences in dose coefficients
mainly originated from changes in the radiation energy emitted per nuclear transformation. In addition, revisions of the half-lives,
radiation types and decay modes also resulted in changes in the dose coefficients.
The assessment of occupationally exposed medical radiation workers at the Institute of Nuclear Medicine and Oncology (INMOL),
Pakistan has been perfomed. The whole-body radiation exposure doses of 120 workers in nuclear medicine (NM), radiotherapy
(RT) and diagnostic radiology (DR) were measured by using the film badge dosimetry technique for the time interval (2007–11)
and their results presented. The annual average effective doses in NM, RT and DR were found to be well below the permissible
annual limit of 20 mSv (averaged over a period of 5 consecutive y). The declining trend observed in the annual average dose
values during the time interval (2007–11) is an indication of ameliorated radiation protection practices at INMOL, Pakistan.
Positron emission tomography (PET) is increasingly used for delineation of tumour tissue in, for example, radiotherapy treatment
planning. The most common method used is to outline volumes with a certain per cent uptake over background in a static image.
However, PET data can also be collected dynamically and analysed by kinetic models, which potentially represent the underlying
biology better. In the present study, a three-parameter kinetic model was used for voxel-wise evaluation of 11C-acetate data of head/neck tumours. These parameters which represent the tumour blood volume, the uptake rate and the clearance
rate of the tissue were derived for each voxel using a linear regression method and used for segmentation of active tumour
tissue. This feasibility study shows that it is possible to segment images based on derived model parameters. There is, however,
room for improvements concerning the PET data acquisition, noise reduction and the kinetic modelling. In conclusion, this
early study indicates a strong potential of the method even though no ‘true’ tumour volume was available for validation.
Eleven underground miners studies evaluated the risk of lung cancer from exposure in underground mines. Nearly 68 000 miners
were included in the joint study, contributing to nearly 2700 lung cancers. The resulting model of the Biological Effects
of Ionizing Radiation (BEIR) VI Committee considered linear exposure response relationship, which was modified by time since
exposure (TE), attained age and exposure rate. The effect of age at exposure (AE) was not explicitly evaluated. The presentation
aims to show that the modifying effect of AE is substantial if time-since-exposure modification is simultaneously used in
the model. When the excess relative risk per unit exposure (ERR/WLM) is adjusted for TE, the ERR/WLM corresponding to AE <15
is 0.013 and in subsequent categories decreased gradually up to the AE of 40 and more years, which was only 0.004. In comparison
with the BEIR VI model, the present model predicts higher risks at younger ages and the risk decreases more rapidly.
This study computationally investigates in situ electric field due to low-frequency contact current and specific absorption rate (SAR) due to high-frequency contact currents
in a realistic child model and compared with those in the adult model. The in situ electric fields and SAR in the child model are found to exceed the corresponding values in the adult. At the finger tip,
the electric field and SAR due to contact currents, both at the ICNIRP reference levels and IEEE Maximum Permissible Exposures,
are well beyond the corresponding basic restrictions. In the remaining part, the largest difference was observed in spinal
tissue, and the smallest effect was in the heart. With respect to brain and skin conductivities, one needs to strongly consider
which values of tissue properties are used to interpret one's results. The in situ electric fields resulting from contact with the metal plane are similar to those for contact with the wire.
The aim of this study was to carry out theoretical investigations of power frequency magnetic fields (MFs), produced inside
and outside the domain of urban 110-kV power substations and to establish a correspondence between the levels of the fields
and the specified population limits as defined by Ukrainian regulations. The fields produced by high-voltage substations were
studied based on the application of the numerical finite element methodology. The investigations have shown that magnetic
flux density values calculated inside and outside the considered 110-kV power substations do not reach the exposure limits
specified by the Ukrainian regulations (1750 μT) and by international guidelines (ICNIRP 2010). Inside the domain of the substation,
the maximum value of MFs was found under the 10-kV busbars and it equalled 420 μT.
In this study, the effective absorbed dose to human organs was estimated, following intra vascular administration of 111In-DTPA-Buserelin using biodistribution data from rats. Rats were sacrificed at exact time intervals of 0.25, 0.5, 1, 2, 4
and 24 h post injections. The Medical Internal Radiation Dose formulation was applied to extrapolate from rats to humans and
to project the absorbed radiation dose for various human organs. From rat data, it was estimated that a 185-MBq injection
of 111In-DTPA-Buserelin into the human might result in an estimated absorbed dose of 24.27 mGy to the total body and the highest
effective absorbed dose was in kidneys, 28.39 mSv. The promising results of this study emphasises the importance of absorbed
doses in humans estimated from data on rats.
According to the Euratom Directives (96/29, 97/43), the doses received by the workers as well as the family of patients and third persons during medical exposures, should conform to the dose constraint levels (DCLs), established by the authorities for each group in the context of optimisation. This study deals with the implementation of a radiation protection protocol, concerning the aforementioned group members for patients undergoing treatment with 111In-DTPA-D-Phe1-Octreotide, after intra-arterial infusion. It is shown that by applying this protocol the annual doses to the medical and technical staff are considerably reduced and remain below the established DCLs. Following the post-release behaviour instructions given to the patient, doses to the family and third persons may be kept lower than the corresponding DCLs provided by the National Regulations.
111In (Eγ=171–245 keV, t1/2=2.83 d) is used for targeted therapies of endocrine tumours. An average activity of 6.3 GBq is injected into the liver
by catheterisation of the hepatic artery. This procedure is time-consuming (4–5 min) and as a result, both the physicians
and the technical staff involved are subjected to radiation exposure. In this research, the efficiency of the use of lead
apron has been studied as far as the radiation protection of the working staff is concerned. A solution of 111In in a cylindrical scattering phantom was used as a source. Close to the scattering phantom, an anthropomorphic male Alderson
RANDO phantom was positioned. Thermoluminescent dosemeters were located in triplets on the front surface, in the exit and
in various depths in the 26th slice of the RANDO phantom. The experiment was repeated by covering the RANDO phantom by a lead
apron 0.25 mm Pb equivalent. The unshielded dose rates and the shielded photon dose rates were measured. Calculations of dose
rates by Monte Carlo N-particle transport code were compared with this study's measurements. A significant reduction of 65
% on surface dose was observed when using lead apron. A decrease of 30 % in the mean absorbed dose among the different depths
of the 26th slice of the RANDO phantom has also been noticed. An accurate correlation of the experimental results with Monte
Carlo simulation has been achieved.
The most important characteristic of the hazard due to cosmic radiation is the spectrum of linear energy transfer (LET), which enables one to estimate the dose equivalent. This has prompted us to study LET spectra of cosmic radiation aboard Cosmos-1129 using nuclear emulsions as a threshold detector.
Preliminary studies have been conducted into the occupational radiation exposure to NORMS from surface and underground mining operations in a gold mine in the Ashanti Region of Ghana. A brief description of the methods and instrumentation is presented. The annual effective dose has been estimated to be 0.26 +/- 0.11 mSv for surface mining and 1.83 +/- 0.56 mSv for the underground mines using the ICRP dose calculation method. The results obtained are found to be within the allowable limit of 20 mSv per annum for occupational exposure control recommended by the ICRP.