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The HURRA filter: an easy method to eliminate collimator artifacts in high-energy gamma camera images / Filtro HURRA: una manera sencilla de eliminar los artefactos del colimador en la imagen con gammacámara de alta energía

Purpose. The correct determination and delineation of tumor/organ size is crucial in 2-D imaging in 131I therapy. These images are usually obtained using a system composed of a Gamma camera and high-energy collimator, although the system can produce artifacts in the image. This article analyses these artifacts and describes a correction filter that can eliminate those collimator artifacts. Methods. Using free software, ImageJ, a central profile in the image is obtained and analyzed. Two components can be seen in the fluctuation of the profile: one associated with the stochastic nature of the radiation, plus electronic noise and the other periodically across the position in space due to the collimator. These frequencies are analytically obtained and compared with the frequencies in the Fourier transform of the profile. A specially developed filter removes the artifacts in the 2D Fourier transform of the DICOM image. This filter is tested using a 15-cm-diameter Petri dish with 131I radioactive water (big object size) image, a 131I clinical pill (small object size) image, and an image of the remainder of the lesion of two patients treated with 3.7GBq (100mCi), and 4.44GBq (120mCi) of 131I, respectively, after thyroidectomy. Results. The artifact is due to the hexagonal periodic structure of the collimator. The use of the filter on large-sized images reduces the fluctuation by 5.8-3.5%. In small-sized images, the FWHM can be determined in the filtered image, while this is impossible in the unfiltered image. The definition of tumor boundary and the visualization of the activity distribution inside patient lesions improve drastically when the filter is applied to the corresponding images obtained with HE gamma camera. Conclusion. The HURRA filter removes the artifact of high-energy collimator artifacts in planar images obtained with a Gamma camera without reducing the image resolution. It can be applied in any study of patient quantification because the number of counts remains invariant. The filter makes possible the definition and delimitation of small uptakes, such as those presented in treatments with 131I (AU)

Objetivo. En terapia con 131I es crucial la correcta delimitación y determinación del tamaño del tumor/órgano en la imagen obtenida con un equipo de gammacámara y colimador de alta energía. Sin embargo, en estas imágenes aparecen artefactos debidos al colimador que dificultan dicha tarea. En este trabajo se analizan dichos artefactos y se describe un filtro corrector que puede eliminar los artefactos producidos por el colimador. Métodos. Utilizando software de descarga gratuita, Image J, se obtiene un perfil central en la imagen que se analiza. Pueden distinguirse 2 componentes en la fluctuación del perfil: uno asociado a la naturaleza estocástica de la desintegración radiactiva más el ruido electrónico y otro con periodicidad espacial en la posición debida al colimador. Las frecuencias correspondientes a esta periodicidad se obtienen analíticamente y se comparan con las frecuencias conseguidas en la transformada de Fourier del perfil. Se desarrolla un filtro capaz de suprimir estos artefactos en la transformada de Fourier 2D de la imagen DICOM. Este filtro se verifica con la imagen de una placa Petri de 15cm de diámetro rellena de agua radiactiva con 131I (objeto de tamaño grande), con la imagen de una pastilla clínica de 131I (objeto de pequeño tamaño) y con la imagen obtenida con gammacámara y colimador de alta energía del resto tiroideo de 2 pacientes tratados con 3,7GBq y 4.44GBq de 131I tras sufrir tiroidectomía. Resultados. Los artefactos se deben a la estructura periódica hexagonal del colimador. La aplicación del filtro en imágenes de objetos de gran tamaño reduce la fluctuación del 5,8% al 3,5%. En imágenes de pequeño tamaño la aplicación del filtro permite obtener el FWHM (ancho a mitad de altura), lo que resulta imposible en la imagen sin filtrar. La definición del contorno y la distribución de actividad en el interior de lesiones tiroideas mejoran substancialmente cuando se aplica el filtro a las imágenes obtenidas con gammacámara y colimador de alta energía. Conclusión. El filtro denominado HURRA suprime los artefactos en la imagen planar obtenida con colimador de alta energía sin reducir la resolución de la imagen. Puede ser utilizado en cualquier estudio de cuantificación en imagen de pacientes ya que el número de cuentas permanece invariante. El filtro permite la definición y delimitación de pequeñas captaciones como las que se producen en tratamientos con 131I (AU)

The HURRA filter: An easy method to eliminate collimator artifacts in high-energy gamma camera images.

PURPOSE: The correct determination and delineation of tumor/organ size is crucial in 2-D imaging in 131I therapy. These images are usually obtained using a system composed of a Gamma camera and high-energy collimator, although the system can produce artifacts in the image. This article analyses these artifacts and describes a correction filter that can eliminate those collimator artifacts. METHODS: Using free software, ImageJ, a central profile in the image is obtained and analyzed. Two components can be seen in the fluctuation of the profile: one associated with the stochastic nature of the radiation, plus electronic noise and the other periodically across the position in space due to the collimator. These frequencies are analytically obtained and compared with the frequencies in the Fourier transform of the profile. A specially developed filter removes the artifacts in the 2D Fourier transform of the DICOM image. This filter is tested using a 15-cm-diameter Petri dish with 131I radioactive water (big object size) image, a 131I clinical pill (small object size) image, and an image of the remainder of the lesion of two patients treated with 3.7GBq (100mCi), and 4.44GBq (120mCi) of 131I, respectively, after thyroidectomy. RESULTS: The artifact is due to the hexagonal periodic structure of the collimator. The use of the filter on large-sized images reduces the fluctuation by 5.8-3.5%. In small-sized images, the FWHM can be determined in the filtered image, while this is impossible in the unfiltered image. The definition of tumor boundary and the visualization of the activity distribution inside patient lesions improve drastically when the filter is applied to the corresponding images obtained with HE gamma camera. CONCLUSION: The HURRA filter removes the artifact of high-energy collimator artifacts in planar images obtained with a Gamma camera without reducing the image resolution. It can be applied in any study of patient quantification because the number of counts remains invariant. The filter makes possible the definition and delimitation of small uptakes, such as those presented in treatments with 131I.

Estimation of neutron-equivalent dose in organs of patients undergoing radiotherapy by the use of a novel online digital detector.

Neutron peripheral contamination in patients undergoing high-energy photon radiotherapy is considered as a risk factor for secondary cancer induction. Organ-specific neutron-equivalent dose estimation is therefore essential for a reasonable assessment of these associated risks. This work aimed to develop a method to estimate neutron-equivalent doses in multiple organs of radiotherapy patients. The method involved the convolution, at 16 reference points in an anthropomorphic phantom, of the normalized Monte Carlo neutron fluence energy spectra with the kerma and energy-dependent radiation weighting factor. This was then scaled with the total neutron fluence measured with passive detectors, at the same reference points, in order to obtain the equivalent doses in organs. The latter were correlated with the readings of a neutron digital detector located inside the treatment room during phantom irradiation. This digital detector, designed and developed by our group, integrates the thermal neutron fluence. The correlation model, applied to the digital detector readings during patient irradiation, enables the online estimation of neutron-equivalent doses in organs. The model takes into account the specific irradiation site, the field parameters (energy, field size, angle incidence, etc) and the installation (linac and bunker geometry). This method, which is suitable for routine clinical use, will help to systematically generate the dosimetric data essential for the improvement of current risk-estimation models.

Neutron dosimetry in organs of an adult human phantom using linacs with multileaf collimator in radiotherapy treatments.

PURPOSE: To calculate absorbed doses due to neutrons in 87 organs/tissues for anthropomorphic phantoms, irradiated in position supine (head first into the gantry) with orientations anteroposterior (AP) and right-left (RLAT) with a 18 MV accelerator. Conversion factors from monitor units to µGy per neutron in organs, equivalent doses in organs/tissues, and effective doses, which permit to quantify stochastic risks, are estimated. METHODS: MAX06 and FAX06 phantoms were modeled with MCNPX and irradiated with a 18 MV Varian Clinac 2100C/D accelerator whose geometry included a multileaf collimator. Two actual fields of a pelvic treatment were simulated using electron-photon-neutron coupled transport. Absorbed doses due to neutrons were estimated from kerma. Equivalent doses were estimated using the radiation weighting factor corresponding to an average incident neutron energy 0.47 MeV. Statistical uncertainties associated to absorbed doses, as calculated by MCNPX, were also obtained. RESULTS: Largest doses were absorbed in shallowest (with respect to the neutron pathway) organs. In µGyMU(-1), values of 2.66 (for penis) and 2.33 (for testes) were found in MAX06, and 1.68 (for breasts), 1.05 (for lenses of eyes), and 0.94 (for sublingual salivary glands) in FAX06, in AP orientation. In RLAT, the largest doses were found for bone tissues (leg) just at the entrance of the beam in the body (right side in our case). Values, in µGyMU(-1), of 1.09 in upper leg bone right spongiosa, for MAX06, and 0.63 in mandible spongiosa, for FAX06, were found. Except for gonads, liver, and stomach wall, equivalent doses found for FAX06 were, in both orientations, higher than for MAX06. Equivalent doses in AP are higher than in RLAT for all organs/tissues other than brain and liver. Effective doses of 12.6 and 4.1 µSvMU(-1) were found for AP and RLAT, respectively. The organs/tissues with larger relative contributions to the effective dose were testes and breasts, in AP, and breasts and red marrow, in RLAT. Equivalent and effective doses obtained for MAX06/FAX06 were smaller (between 2 and 20 times) than those quoted for the mathematical phantoms ADAM/EVA in ICRP-74. CONCLUSIONS: The new calculations of conversion coefficients for neutron irradiation in AP and RLAT irradiation geometries show a reduction in the values of effective dose by factors 7 (AP) and 6 (RLAT) with respect to the old data obtained with mathematical phantoms. The existence of tissues or anatomical regions with maximum absorbed doses, such as penis, lens of eyes, fascia (part of connective tissue), etc., organs/tissues that classic mathematical phantoms did not include because they were not considered for the study of stochastic effects, has been revealed. Absorbed doses due to photons, obtained following the same simulation methodology, are larger than those due to neutrons, reaching values 100 times larger as the primary beam is approached. However, for organs far from the treated volume, absorbed photon doses can be up to three times smaller than neutron ones. Calculations using voxel phantoms permitted to know the organ dose conversion coefficients per MU due to secondary neutrons in the complete anatomy of a patient.

Ambient neutron dose equivalent outside concrete vault rooms for 15 and 18 MV radiotherapy accelerators.

In this work, the ambient dose equivalent, H*(10), due to neutrons outside three bunkers that house a 15- and a 18-MV Varian Clinac 2100C/D and a 15-MV Elekta Inor clinical linacs, has been calculated. The Monte Carlo code MCNPX (v. 2.5) has been used to simulate the neutron production and transport. The complete geometries including linacs and full installations have been built up according to the specifications of the manufacturers and the planes provided by the corresponding medical physical services of the hospitals where the three linacs operate. Two of these installations, those lodging the Varian linacs, have an entrance door to the bunker while the other one does not, although it has a maze with two bends. Various treatment orientations were simulated in order to establish plausible annual equivalent doses. Specifically anterior-posterior, posterior-anterior, left lateral, right lateral orientations and an additional one with the gantry rotated 30° have been studied. Significant dose rates have been found only behind the walls and the door of the bunker, near the entrance and the console, with a maximum of 12 µSv h(-1). Dose rates per year have been calculated assuming a conservative workload for the three facilities. The higher dose rates in the corresponding control areas were 799 µSv y(-1), in the case of the facility which operates the 15-MV Clinac, 159 µSv y(-1), for that with the 15-MV Elekta, and 21 µSv y(-1) for the facility housing the 18-MV Varian. A comparison with measurements performed in similar installations has been carried out and a reasonable agreement has been found. The results obtained indicate that the neutron contamination does not increase the doses above the legal limits and does not produce a significant enhancement of the dose equivalent calculated. When doses are below the detection limits provided by the measuring devices available today, MCNPX simulation provides an useful method to evaluate neutron dose equivalents based on a detailed description of linac, patient and bunker.

131I activity in urine to the sewer system due to thyroidal treatments.

In nuclear medicine, estimating the radioactivity contained in the urine of patients treated with I and discharged to the environment could prevent the exposure of a population to radioactive effluents and the pollution of the aquatic environment with ionizing radiation. This can be a regulatory requirement (as in Spain) or requested by the sewer authority. Seventy-nine differentiated thyroid cancer cases (undergone as inpatients) and 187 hyperthyroidism cases (undergone as outpatients) were treated in our hospital with I throughout the year 2009. In hyperthyroidism treatments, the effective elimination constant was used to calculate the corresponding discharged activity in the urine, giving an activity level always below 0.7 GBq. In differentiated thyroid cancer treatments, patient's urine was collected in storage tanks during the hospitalization. Measurements of external exposure at 1 m made every day were used to calculate the activity contained in the urine. The tank activity was always below 15 GBq, but always higher than 2 GBq. Obtained results show that effective doses to sewage workers, received from liquid discharges, can only be reduced to less than 10 µSv if storage tanks are installed. Without tanks, 157 µSv can be reached, above the constrain dose used in nuclear installations (100 µSv). Our calculations may be helpful to the regulatory authority to review the clinical radiation waste normative, especially in countries where the discharges are released directly into public sewage plants.

Neutron dose equivalent and neutron spectra in tissue for clinical linacs operating at 15, 18 and 20 MV.

In this work, the dose equivalent due to photoneutrons and the neutron spectra in tissue was calculated for various linacs (Varian Clinac 2100C, Elekta Inor, Elekta SL25 and Siemens Mevatron KDS) operating at energies between 15 and 20 MV, using the Monte Carlo code MCNPX (v. 2.5). The dose equivalent in an ICRU tissue phantom has been calculated for anteroposterior treatments with a detailed simulation of the geometry of the linac head and the coupled electron-photon-neutron transport. Neutron spectra at the phantom entrance and at 1-cm depth in the phantom, depth distribution of the neutron fluence in the beam axis and dose distributions outside the beam axis at various depths have also been calculated and compared with previously published results. The differences between the neutron production of the various linacs considered has been analysed. Varian linacs show a larger neutron production than the Elekta and Siemens linacs at the same operating energy. The dose equivalent due to neutrons produced by medical linacs operating at energies >15 MeV is relevant and should not be neglected because of the additional doses that patients can receive.

A comprehensive study of the Gafchromic EBT2 radiochromic film. A comparison with EBT.

PURPOSE: Radiochromic film has become an important tool to assess complex dose distributions. In particular, EBT was accepted by the scientific community as a reference two-dimensional detector. Recently, Gafchromic EBT2 has replaced old film, providing new improvements in both accuracy and handling. METHODS: This work presents a dosimetric study of the new Gafchromic EBT2 using an Epson 10000XL flatbed scanner, also comparing the results with EBT film as reference when necessary. The most important film characteristics have been studied, such as ambient light sensitivity, different possibilities of the three RGB color channels, postirradiation development, high dose behavior, exposition at temperatures similar to the human body, and dependence on orientation during the scanning process. RESULTS: The results obtained confirm a considerably lower sensitivity to ambient light of EBT2, as well as a fast stabilization of the film within 2 h. It has also been found that the green channel has a better behavior at high dose levels up to 35 Gy, in addition to good behavior of the red channel at doses below 10 Gy. Other features, such as temperature independence and scanning orientation dependence, have also been shown. CONCLUSIONS: Gafchromic EBT2 can be used for clinical practice in the same way as the old EBT film. However, a much easier handling as the result of all new enhancements improves film behavior, expanding in this way the potential applications of radiochromic film dosimetry.

Evaluation of CdZnTe as neutron detector around medical accelerators.

The operation of electron linear accelerators (LINACs) and cyclotrons can produce a mixed gamma-neutron field composed of energetic neutrons coming directly from the source and scattered lower energy neutrons. The thermal neutron detection properties of a non-moderated coplanar-grid CdZnTe (CZT) gamma-ray detector close to an 18 MV electron LINAC and an 18 MeV proton cyclotron producing the radioisotope (18)F for positron emission tomography are investigated. The two accelerators are operated at conditions producing similar thermal neutron fluence rates of the order of 10(4) cm(-2) s(-1) at the measurement locations. The counting efficiency of the CZT detector using the prompt 558 keV photopeak following (113)Cd thermal neutron capture is evaluated and a good neutron detection performance is found at the two installations.

Correlation between external exposure and activity in patients undergoing 131I thyroid cancer therapy.

The uptake and clearance of 131I activity for inpatients undergoing cancer therapy were determined from routine external dose survey measurements. A bi-exponential behavior was found, with the two time constants representing the iodine dynamics in the thyroid on one hand and in the rest of the body on the other. The external dose at 1 m from the patient was correlated to the activity in the thyroid remnant and inside the body, the averaged value being 52.8 +/- 11.4 microSv GBq(-1) h(-1). The temporal evolution of activity in the body, the urinary system and the thyroid remnant area were determined taking into account the clearance from thyroid and whole body (effective retention constants averages 0.23 +/- 0.14 d(-1) and 1.46 +/- 0.34 d(-1)) and the uptake in thyroid (3.15 +/- 3.36%). Applications of this study in the public and environmental radiation protection areas are presented.

Liquid discharges from patients undergoing 131I treatments.

This work discusses the production and management of liquid radioactive wastes as excretas from patients undergoing therapy procedures with 131I radiopharmaceuticals in Spain. The activity in the sewage has been estimated with and without waste radioactive decay tanks. Two common therapy procedures have been considered, the thyroid cancer (4.14 GBq administered per treatment), and the hyperthyroidism (414 MBq administered per treatment). The calculations were based on measurements of external exposure around the 244 hyperthyroidism patients and 23 thyroid cancer patients. The estimated direct activity discharged to the sewage for two thyroid carcinomas and three hyperthyroidisms was 14.57 GBq and 1.27 GBq, respectively, per week; the annual doses received by the most exposed individual (sewage worker) were 164 microSv and 13 microSv, respectively. General equations to calculate the activity as a function of the number of patient treated each week were also obtained.

Liquid discharges from the use of radionuclides in medicine (diagnosis).

The production and discharge of liquid radioactive wastes as excreta from patients undergoing Nuclear Medicine Diagnostic (NMD) in a hospital were studied. Instantaneous and accumulated activity, discharged from the hospital to the sewage system, has been estimated keeping in mind radionuclide decay. This study would enable estimation of the environmental impact due to NMD procedures. Annual accumulated activities of 2.2 GBq (131I), 1.847 GBq (99mTc), 0.743 GBq (123I), 0.337 GBq (67Ga), 0.169 GBq (111In) and 0.033 GBq (201Tl) result from our model when applied to a European hospital. A comparison is made with calculations by other authors that do not consider the radionuclide decay and who overestimate by two orders of magnitude. Doses to critical people as sewage treatment workers are also significantly reduced. So, our results stress the importance of including the decay in the calculations.

Neutron KERMA factors of human tissues

A program to calculate the neutron KERMA in human tissues has been developed. The program was developed in Mathcad and contains the neutron kerma factors of those elements that are present in different human tissues. Having the elemental composition of any human tissue the neutron kerma can be easily calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic. Neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large.

Neutron spectrometry in a PET cyclotron with a Bonner sphere system.

Positron emission tomography (PET) is a non-invasive medical imaging technique normally used for diagnostic purposes to determine the location and concentration of physiologically active compounds in a human body. An unshielded cyclotron is used for PET at the Clinica Universitaria de Navarra to produce short-lived positron emitting radionuclides ((15)O, (13)N, (11)C and (18)F) by bombarding appropriate target material with proton or deuteron beams with energies up to 18 and 9 MeV, respectively. Subsequent nuclear reactions may generate undesirable neutrons that should be evaluated and controlled. In this study, the neutron measurements performed with an active and a passive Bonner sphere systems at different locations outside and inside the cyclotron vault during operation have been presented. The neutron spectrum at each location was determined with an unfolding code developed by the authors.

Monte Carlo neutron doses estimations inside a PET cyclotron vault room.

Neutron organ equivalent doses, effective doses and dose equivalents received inside a positron emission tomography vault room in a maximum credible accident have been estimated with the Monte Carlo code MCNPX. While an operator was inside the vault room of a Cyclone 18/9 IBA cyclotron, this was producing (18)F with 30 muA proton current in the target and the operator had to activate a stopped emergency device placed on the wall. MC simulation of the cyclotron vault were carried out to estimate the organ and tissue equivalent doses in a mathematical male mannequin simulating the operator facing the wall on which the emergency device is placed. Doses were calculated at two emergency devices for each one of the two targets of the cyclotron, which were able to produce (18)F. The maximum effective dose in the mannequin was 6.70 Sv/h and the maximum organ equivalent dose was 18.47 Sv/h in spleen.

Campos neutrónicos producidos por un acelerador lineal para radioterapia / Neutron fields produced by linear accelerator for radiotherapy

Mediante una serie de medidas y cálculos Monte Carlo se han determinado las características dosimétricas y los espectros de los fotoneutrones que se producen en torno a un acelerador lineal para radioterapia de 18 MV. Las medidas se realizaron con dosímetros termoluminiscentes TLD 600 y TLD 700 que se expusieron desnudos y emparedados con Cd, así como dentro de una esfera de parafina y dentro de esferas Bonner.

Measurements and Monte Carlo calculations has been utilized to determine the dosimetric features as well as the neutron spectra of photoneutrons produced around an 18 MV linear accelerator for radiotherapy. Measurementes were carried out with bare and Cd covered thermolumiscent dosimeters, TLD600 and TLD700, as well as inside a paraffine moderator. TLD pairs were also utilized as thermal neutrons inside a Bonner sphere spectrometer.

Study of the neutron field in the vicinity of an unshielded PET cyclotron.

The neutron field in the proximity of an unshielded PET cyclotron was investigated during 18F radioisotope production with an 18 MeV proton beam. Thermoluminescent detector (TLD) models TLD600 and TLD700 as well as Bonner moderating spheres were irradiated at different positions inside the vault room where the cyclotron is located to determine the thermal neutron flux, neutron spectrum and dose equivalent. Furthermore, from a combination of measurements and Monte Carlo simulations the neutron source intensity at the target was estimated. The resulting intensity is in good agreement with the IAEA recommendations. Neutron doses derived from the measured spectra were found to vary between 7 and 320 mSv per 1 microA h of proton-integrated current. Finally, gamma doses were determined from TLD700 readings and amounted to around 10% of the neutron doses.

Monte Carlo simulation estimates of neutron doses to critical organs of a patient undergoing 18 MV x-ray LINAC-based radiotherapy.

Absorbed photoneutron dose to patients undergoing 18 MV x-ray therapy was studied using Monte Carlo simulations based on the MCNPX code. Two separate transport simulations were conducted, one for the photoneutron contribution and another for neutron capture gamma rays. The phantom model used was of a female patient receiving a four-field pelvic box treatment. Photoneutron doses were determinate to be higher for organs and tissues located inside the treatment field, especially those closest to the patient's skin. The maximum organ equivalent dose per x-ray treatment dose achieved within each treatment port was 719 microSv/Gy to the rectum (180 degrees field), 190 microSv/Gy to the intestine wall (0 degrees field), 51 microSv/Gy to the colon wall (90 degrees field), and 45 microSv/Gy to the skin (270 degrees field). The maximum neutron equivalent dose per x-ray treatment dose received by organs outside the treatment field was 65 microSv/Gy to the skin in the antero-posterior field. A mean value of 5 +/- 2 microSv/Gy was obtained for organs distant from the treatment field. Distant organ neutron equivalent doses are all of the same order of magnitude and constitute a good estimate of deep organ neutron equivalent doses. Using the risk assessment method of the ICRP-60 report, the greatest likelihood of fatal secondary cancer for a 70 Gy dose is estimated to be 0.02% for the pelvic postero-anterior field, the rectum being the organ representing the maximum contribution of 0.011%.