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Internal Bremsstrahlung (IB) is a continuous electromagnetic radiation accompanying beta decay; however, this process is not considered in radiation protection studies, particularly when estimating exposure from beta-decaying radionuclides. The aims of the present work are: i) to show that neglecting the IB process in Monte Carlo (MC) simulation leads to an underestimation of the energy deposited in a ionization chamber, in the case of a high-energy pure beta emitter such as Yttrium-90 (90Y), and ii) to determine the most reliable choice of source term for 90Y IB to be used in MC simulations. For this radionuclide, commonly employed in nuclear medicine and radiochemistry applications, experimental data acquired with a well ionization chamber have been compared with Monte Carlo (MC) calculations carried out in the GAMOS framework. Simulations that do not include the effect of the IB process, are found to give results underestimating the experimental values by 12–14%. Consequently, two models for the IB energy spectra, previously described by Italiano et al. [1] , have been implemented using MC simulation and a good agreement has been achieved with one of them. We therefore conclude that inclusion of IB process in Monte Carlo simulation packages is advisable for a more accurate and complete treatment of electromagnetic interactions.

Reverse osmosis (RO) is a water purification process that uses a semi-permeable membrane to remove ions from potable water. It has high rejection throughput, low energy consumption, and negligible pollution load when compared to conventional treatment methods. Comparative percentage salt rejection (SR) efficiency for surrogates of cesium and molybdenum was performed using commercially available membranes. Polyamide, polysulfone, polyamide–polysulfone composite, and cellulose acetate were subjected to various operating conditions in a domestically developed high-pressure membrane test cell. Five different concentrations of surrogate salts ranging from 100 to 500 ppm and varied pressures of 15–17 kg/cm2 combined with varying temperatures of the feed solution from 25°C to 45°C were used in this experimental work. It was found that the %SR efficiency of these membranes increased with the increase in salt concentration and feed temperature. However, a significant decrease in SR was observed with increasing pressure. A comparative study of these commercially available RO membranes was also performed against short-lived radioisotope Technetium-99m, and was determined by activity counts of feed and filtered samples using a well counter. The results suggest that the rejection efficiency was found to be highest in the case of polyamide–polysulfone composite, followed by polysulfone, polyamide, and cellulose acetate.

This paper presents a test of a methodology to describe the shape of true and accidental coincidence gamma-ray spectra. The effect of the coincidence summing was enhanced by using a 50 kBq 22Na source in a NaI well counter. The spectra were described by convoluting Monte Carlo simulated spectra in a spreadsheet for true, as well accidental coincidence summing. In this paper we show that applying convolution techniques to the MCNP simulated probability spectra, not only allows a reconstruction of the shape of the true coincidence spectrum, but the simulations also prove to be an invaluable tool to understand the continuum in the spectra made up of double and even triple random coincidences. Each of these summing processes was used to generate so-called standard spectra that may be used in a full-spectrum analysis of the measured spectra to quantitatively derive the contributions of each of the summing processes. For the case presented it is shown that the accidental coincidences can be accounted for to a 0.1% level, a considerable improvement to the precision in previous papers.

In nuclear medicine, (68)Ge is used to generate (68)Ga for imaging by positron emission tomography (PET) and sealed sources containing (68)Ge/(68)Ga in equilibrium have been adopted as long-lived calibration surrogates for the more common PET nuclide, (18)F. We prepared several (68)Ge sources for measurement on a NaI(Tl) well counter and a pressurized ionization chamber, following their decay for 110 weeks (≈ 2.8 half-lives). We determined values for the (68)Ge half-life of T(1/2) = 271.14(15) d and T(1/2) = 271.07(12) d from the NaI(Tl) well counter and ionization chamber measurements, respectively. These are in accord with the current Decay Data Evaluation Project (DDEP) recommended value of T(1/2) = 270.95(26) d and we discuss the expected impact of our measurements on this value.

Purpose Genomic profiling of biopsied tissue is the basis for precision cancer therapy. However, biopsied materials may not contain sufficient amounts of tumor deoxyribonucleonic acid needed for the analysis. We propose a method to determine the adequacy of specimens for performing genomic profiling by quantifying their metabolic activity. Methods We estimated the average density of tumor cells in biopsy specimens needed to successfully perform genomic analysis following the Memorial Sloan Kettering Integrated Mutation Profiling of Actionable Cancer Targets (MSK-IMPACT) protocol from the minimum amount of deoxyribonucleonic acid needed and the volume of tissue typically used for analysis. The average 18 F-FDG uptake per cell was assessed by incubating HT-29 adenocarcinoma tumor cells in 18 F-FDG containing solution and then measuring their activity with a scintillation well counter. Consequently, we evaluated the response of two devices around the minimum expected activities which would indicate genomic profiling adequacy of biopsy specimens obtained under 18 F-FDG PET/CT guidance. Surrogate samples obtained using 18G core needle biopsies of gels containing either 18 F-FDG-loaded cells in the expected concentrations or the corresponding activity were measured using autoradiography and a scintillation well counter. Autoradiography was performed using a CCD-based device with real-time image display as well as with digital autoradiography imaging plates following a 30-min off-line protocol for specimen activity determination against previously established calibration. Results Cell incubation experiments and estimates obtained from quantitative autoradiography of biopsy specimens (QABS) indicate that specimens acquired under 18 F-FDG PET/CT guidance that contained the minimum amount of cells needed for genomic profiling would have an average activity concentration in the range of about 3 to about 9 kBq/mL. When exposed to specimens with similar activity concentration, both a CCD-based autoradiography device and a scintillation well counter produced signals with sufficient signal-to-background ratio for specimen genomic adequacy identification in less than 10 min, which is short enough to allow procedure guidance. Conclusion Scintillation well counter measurements and CCD-based autoradiography have adequate sensitivity to detect the tumor burden needed for genomic profiling during 18 F-FDG PET/CT-guided 18G core needle biopsies of liver adenocarcinoma metastases.

In the standard practice of neutron multiplicity counting , the first three sampled factorial moments of the event triggered neutron count distribution are used to quantify the three main neutron source terms: the spontaneous fissile material effective mass, the relative ( α , n ) production and the induced fission source responsible for multiplication. This study compares three methods to quantify the statistical uncertainty of the estimated mass: the bootstrap method, propagation of variance through moments, and statistical analysis of cycle data method. Each of the three methods was implemented on a set of four different NMC measurements, held at the JRC-laboratory in Ispra, Italy, sampling four different Pu samples in a standard Plutonium Scrap Multiplicity Counter (PSMC) well counter.

The National Institute of Standards and Technology (NIST) performed new standardization measurements for 64Cu. As part of this work the photon emission probability for the main gamma-ray line and the half-life were determined using several high-purity germanium (HPGe) detectors. Half-life determinations were also carried out with a NaI(Tl) well counter and two pressurized ionization chambers.

Introduction Quantitative small animal radionuclide imaging studies are often carried out with the intention of estimating the total radioactivity content of various tissues such as the radioactivity content of mouse xenograft tumors exposed to putative diagnostic or therapeutic agents. We show that for at least one specific application, positron projection imaging (PPI) and PET yield comparable estimates of absolute total tumor activity and that both of these estimates are highly correlated with direct well-counting of these same tumors. These findings further suggest that in this particular application, PPI is a far more efficient data acquisition and processing methodology than PET. Methods Forty-one athymic mice were implanted with PC3 human prostate cancer cells transfected with prostate-specific membrane antigen (PSMA (+)) and one additional animal (for a total of 42) with a control blank vector (PSMA (−)). All animals were injected with [ 18 F] DCFPyl, a ligand for PSMA, and imaged for total tumor radioactivity with PET and PPI. The tumors were then removed, assayed by well counting for total radioactivity and the values between these methods intercompared. Results PET, PPI and well-counter estimates of total tumor radioactivity were highly correlated (R 2 > 0.98) with regression line slopes near unity (0.95 Conclusion Total mouse xenograft tumor radioactivity can be measured with PET or PPI with an accuracy comparable to well counting if certain experimental and pharmacokinetic conditions are met. In this particular application, PPI is significantly more efficient than PET in making these measurements.

Introduction We describe a simple fixture that can be added to the imaging bed of a small-animal PET scanner that allows for automated counting of multiple organ or tissue samples from mouse-sized animals and counting of injection syringes prior to administration of the radiotracer. The combination of imaging and counting capabilities in the same machine offers advantages in certain experimental settings. Methods A polyethylene block of plastic, sculpted to mate with the animal imaging bed of a small-animal PET scanner, is machined to receive twelve 5-ml containers, each capable of holding an entire organ from a mouse-sized animal. In addition, a triangular cross-section slot is machined down the centerline of the block to secure injection syringes from 1-ml to 3-ml in size. The sample holder is scanned in PET whole-body mode to image all samples or in one bed position to image a filled injection syringe. Total radioactivity in each sample or syringe is determined from the reconstructed images of these objects using volume re-projection of the coronal images and a single region-of-interest for each. We tested the accuracy of this method by comparing PET estimates of sample and syringe activity with well counter and dose calibrator estimates of these same activities. Results PET and well counting of the same samples gave near identical results (in MBq, R 2 =0.99, slope=0.99, intercept=0.00-MBq). PET syringe and dose calibrator measurements of syringe activity in MBq were also similar ( R 2 =0.99, slope=0.99, intercept=− 0.22-MBq). Conclusion A small-animal PET scanner can be easily converted into a multi-sample and syringe counting device by the addition of a sample block constructed for that purpose. This capability, combined with live animal imaging, can improve efficiency and flexibility in certain experimental settings.

Our previous study demonstrated that a cadmium telluride (CdTe)-based single-photon emission computed tomography (SPECT) system with good energy resolution could provide dual-radioisotope SPECT images of a patient who had been administered with two kinds of radioisotope simultaneously. To obtain two quantitative physiological parametric images from these radioisotope images by using compartment model analysis, the concentrations of the dual-radioisotope in the blood sample need to be measured. In this paper, a CdTe-based well counter was developed to measure the concentrations of a dual-radioisotope of 99mTc and 123I, and its performance was evaluated. The counter consists of four detector panels. Each detector panel consists of two detector modules, and each module has $16\,\, {\times } \,\, 16$ CdTe detector pixels with a pitch of 2.5 mm; the eight detector modules give our system a field of view ( $ {x} {\times } \,\, {y} {\times } {z}$ ) of $4\,\, {\times } \,\, 4\,\, {\times } \,\, 8$ cm3. The crosstalk correction coefficient $ {{{\alpha }}}$ , which is the ratio of the count rate in the 99mTc-energy window to that in the 123I-energy window, was obtained by measuring the single-radioisotope solution of 123I. When the dual-radioisotope of 99mTc and 123I was measured, the true count rate of 99mTc was obtained by subtracting the crosstalk count rate ( $ {{{\alpha }}} \,\, {\times }$ the count rate of 123I) from the uncorrected count rate of 99mTc. To evaluate the measurement accuracy, dual- and single-radioisotope solutions were measured, and the mean percentage error of the measured count rate of dual-radioisotope solutions to that of the single-radioisotope solution was obtained. The energy resolutions full width at half maximum of the CdTe-based well counter were 6.95% and 6.93% at 140.5 keV and 159 keV, respectively, and were higher than that of a conventional NaI (Tl) scintillation detector (10% at 140.5 keV). The count loss in the CdTe detector was 1.7% at 44 kcps. For the dual-radioisotope measurement, the percentage errors (mean percentage% ± standard deviation%, ${N} \,\, {=}$ 3) of the count rate of 99mTc and 123I at a high count rate (99mTc: 44 kcps, 123I: 28 kcps) were −1.9% ± 0.0% and −3.5% ± 1.1%, respectively. On the other hand, the percentage errors of the count rate of 99mTc and 123I at a low count rate (99mTc: 3 kcps, 123I: 4 kcps) were −0.4% ± 0.3% and 0.2% ± 0.6%, respectively. The CdTe-based well counter measured each concentration of 99mTc and 123I in mixed solutions with high accuracy.

AIM: To correlate between the radionuclide in vitro plasma sampling method (using single and dual blood samples) and Gates’ GFR measurement using Tc-99m diethylene triamine penta-acetic acid (Tc-99m DTPA) renal scintigraphy (in vivo method). METHODS: This study included 40 renal donors (group 1) and 40 patients with obstructive uropathy (group 2). Group 1 included 22 males and 18 females with an age range from 22 to 65 years, while group 2 included 24 males and 16 females with age range 27 to 64 years. Both groups subjected to renal Scintigraphy after administration of 5 mCi 99m-Tc DTPA, GFR was calculated using Gates’ method (in vivo method), then plasma sampling was acquired at 60 mins and 180 mins post-injection of the tracer, samples were counted in well counter and GFR was calculated using in vitro technique either using single plasma sample (SPSM 60 mins) or dual sample (DPSM 60 & 180 min). Additionally, GFR was measured by estimated equations based on serum creatinine. RESULTS: In group 1, the mean GFR using in vivo Gates’ method was 115.7 ± 29 ml/min, while using the SPSM was 100.1 ± 16.1 ml/min, and the DPSM was 100.3 ± 20.1 ml/min. In group 2, mean GFR using in vivo method was 74.1 ± 14.5 ml/min, while using in vitro SPSM it was 77.5 ± 24.9 ml/min and DPSM was 76.8 ± 24.8 ml/min. There was no significant difference between mean GFR values using in vivo and in vitro methods (single or dual samples) in group 1 and 2 (p > 0.05). There is high significant correlation between SPSM and DPSM in groups 1 and 2 (r = 0.90, r = 0.91 respectively), moderate significant correlation was found between in vivo Gates’ method and in vitro SPSM in group 1 and 2 (r = 0.46 and 0.57 respectively) and moderate correlation was evident between in vivo and in vitro DPSM in both groups (r = 0.42 and 0.68 respectively). By using the DPSM as the reference standard significant high correlation was found with SPSM and significant-high moderate correlation with in vivo Gates’ scintigraphic method. Conclusion: In vitro plasma sampling considered as a reliable, accurate |method for GFR calculation yet it considered relatively complex, both single and dual sample in vitro techniques showed a very high correlation, and hence SPSM can replace DPSM. CONCLUSION: Renal scintigraphy and GFR estimation using Gates’ in vivo method is considered inaccurate, yet given its simplicity in performance it can still be used if corrected GFR is standardised for Egyptian population-based on studies with large numbers of patients from multiple centres.

BACKGROUND: Multi-tracer positron emission tomography (PET) imaging can be accomplished by applying multi-tracer compartment modeling. Recently, a method has been proposed in which the arterial input functions (AIFs) of the multi-tracer PET scan are explicitly derived. For that purpose, a gamma spectroscopic analysis is performed on blood samples manually withdrawn from the patient when at least one of the co-injected tracers is based on a non-pure positron emitter. Alternatively, these blood samples required for the spectroscopic analysis may be obtained and analyzed on site by an automated detection device, thus minimizing analysis time and radiation exposure of the operating personnel. In this work, a new automated blood sample detector based on silicon photomultipliers (SiPMs) for single- and multi-tracer PET imaging is presented, characterized, and tested in vitro and in vivo. RESULTS: The detector presented in this work stores and analyzes on-the-fly single and coincidence detected events. A sensitivity of 22.6 cps/(kBq/mL) and 1.7 cps/(kBq/mL) was obtained for single and coincidence events respectively. An energy resolution of 35% full-width-half-maximum (FWHM) at 511 keV and a minimum detectable activity of 0.30 ± 0.08 kBq/mL in single mode were obtained. The in vivo AIFs obtained with the detector show an excellent Pearson's correlation (r = 0.996, p < 0.0001) with the ones obtained from well counter analysis of discrete blood samples. Moreover, in vitro experiments demonstrate the capability of the detector to apply the gamma spectroscopic analysis on a mixture of 68Ga and 18F and separate the individual signal emitted from each one. CONCLUSIONS: Characterization and in vivo evaluation under realistic experimental conditions showed that the detector proposed in this work offers excellent sensibility and stability. The device also showed to successfully separate individual signals emitted from a mixture of radioisotopes. Therefore, the blood sample detector presented in this study allows fully automatic AIFs measurements during single- and multi-tracer PET studies. This work was supported by grants from the Carlos III Institute of Health of Spain and Fondo Europeo de Desarrollo Regional (FEDER, “Una manera de hacer Europa”) (FIS-FEDER PI14-01427), and from the Comunidad de Madrid (2016-T1/TIC-1099). CV holds a fellowship from the Spanish Ministry of Education (FPU014/01794). The CNIC is supported by the Instituto de Salud Carlos III (ISCIII), the Ministerio de Ciencia, Innovación y Universidades (MCNU), and the Pro CNIC Foundation, and is a Severo Ochoa Center of Excellence (SEV-2015-0505).

Software for Tomographic Image Reconstruction (STIR: http://stir.sf.net) is an open source C++ library available for reconstruction of emission tomography data. This work aims at the incorporation of the GE SIGNA PET/MR scanner in STIR and enables PET image reconstruction with data corrections. The data extracted from the scanner after an acquisition includes a list of raw data files (emission, normalisation, geometric and well counter calibration (wcc) factors), magnetic resonance attenuation correction (MRAC) images and the scanner-based reconstructions. The listmode (LM) file stores a list of ’prompt’ events and the singles per crystal per second. MRAC images from the scanner are used for attenuation correction. The modifications to STIR that allow accurate histogramming of this LM data in the same sinogram organisation as the scanner are also described. This allows reconstruction of acquisition data with all data corrections using STIR, and independent of any software supplied by the manufacturer. The implementations were validated by comparing the histogrammed data, data corrections and final reconstruction using the ordered subset expectation maximisation (OSEM) algorithm with the equivalents from the GE-toolbox, supplied by the manufacturer for the scanner. There is no difference in the histogrammed counts whereas an overall relative difference of 6.7 × 10−8% and from 0.01% to 0.86% is seen in the normalisation and randoms correction sinograms respectively. The STIR reconstructed images have similar resolution and quantification but have some residual differences due to wcc factors, decay and deadtime corrections, as well as the offset between PET and MR gantries that will be addressed in future work. This work will enable the use of all current and future STIR algorithms, including penalized image reconstruction, motion correction and direct parametric image estimation, on data from GE SIGNA PET/MR scanners.

Two legacy AmLi sources have been actively used across several projects within Pacific Northwest National Laboratory, however, applicability of source measurement data has been substantially limited due to the lack of original manufacturer documentation and cross comparison to other sources. This paper summarizes measurements conducted to better quantify source radionuclide content, neutron energy spectrum and emission rate of the AmLi sources. The measured neutron energy spectra of AmLi sources produced by Monsanto Research Corporation have not been found in open literature before and therefore the focus of the study was to obtain the high definition energy distribution of the sources. The initial gamma spectrometry measurements, aimed to estimate the radionuclide content and potential impurities, revealed a significant contamination of the source material with beryllium. Consequent neutron spectrum measurements confirmed beryllium contamination through the presence of a high energy tail contributing up to 17% to the total neutron emission rate. The obtained lithium-only portion of the measured source energy distribution was compared against other AmLi spectra discussed in the past and recent literature. Determination of the total neutron yields of the sources was made with measurements in a neutron well counter. Detection efficiency of the well counter was calculated specifically for these sources via detailed Monte Carlo simulations based on the obtained neutron energy distribution.

Aim Radionuclide imaging and therapies produce radioactive liquid waste that may lead to significant radiation exposure to the general public. The study aims to assess the radiation exposure rate to public sewerage from a modified delay tank facility. We shall also evaluate the exposure rates and overall radioactivity at several points. Materials and methods After having appropriate permission from the AERB, we measured the radiation exposure from the radionuclide therapy ward. Ward has three isolation beds and a single delay and decay tank of a capacity of 7500 liters. Effluents from the delay tank are processed at the filtration plant of the institute and subsequently released in the public sewerage. We obtained samples from several sites to determine discharged radioactivity. Results A total of 38 patients received 129.4 ± 42 mCi (Range 40- 200) radioiodine therapy during the study. Discharge of the tanks was done two times during the study. The radioactivity discharges into aeration plant were 89.2 and 71.2 mCi that correspond to 440.05 and 351 MBq/m3, respectively. This was diluted by the aeration tank (6 million liters). Finally, at the discharge time, the radioactivity in the discharge was 1.6 and 1.5 MBq/m3, respectively. The highest exposure rates were 14 μSv/h near the delay tank, which rapidly decreased on moving to the surrounding. Conclusion Our study indicates that the addition of the dilution method and close monitoring may significantly reduce the radiation exposure and overall radioactivity release from the facility. Old facilities that do not have space to add up the tank capacity may get a benefit from it. A small change in the practice, such as admitting patients alternate months or providing extra decay time for radioactive waste, may lead to a cost-effective alternative.

Ra-223 is an alpha-emitter that is being used as a bone-seeking radiotherapeutic agent. The relatively large uncertainty on its evaluated half-life (0.26%, Be et al., 2011 ) is an impediment to precision activity assays, which often involve measurements by various methods over time spans of days or weeks. We have performed two series of measurements using an ionization chamber (IC) and a NaI(Tl) well counter (γ-wc) to determine new, precise values for the 223Ra half-life. We have endeavored to realistically assess the uncertainties on the derived half-lives, looking beyond the fit uncertainties to identify uncertainty components acting on multiple timescales. We recovered respective values of 11.447(6) d and 11.445(13) d from the IC and γ-wc measurements. Our values are in accord with the evaluated value of 11.43(3) d, but with smaller combined uncertainties.

Tests for chemical stability were carried out on carrier-added (68)Ge solutions prepared and calibrated in 2007 and 2011 to evaluate the suitability of the specific composition as a potential Standard Reference Material. Massic count rates of the stored solutions were measured using a NaI(Tl) well counter before and after gravimetric transfers. The present activity concentration of the 2007 solution was also measured using live-timed anticoincidence counting (LTAC) and compared to the 2007 calibrated value. The well counter data indicated no change in massic count rate to within uncertainties for either solution. The LTAC measurements gave a difference of -0.49% in the activity concentration 2007 solution over 7 years. However, the uncertainty in the decay correction over that time, due to the uncertainty in the (68)Ge half-life, accounted for the majority (0.67% out of 0.83%) of the standard uncertainty on the activity concentration. The results indicate that these carrier-added solutions are stable with regard to potential activity losses over several half-lives of (68)Ge.

Background: In order to obtain a reliable 177Lu-DOTATATE therapy dosimetry, it is crucial to acquire accurate and precise activity measurements with the radionuclide calibrator, the SPECT/CT camera, and the NaI(Tl) well counter. The aim of this study was to determine, in a clinical context, the accuracy and the precision of their activity quantification over a range of activities and time. Ninety-three 177Lu sources from the manufacturer were measured in the radionuclide calibrator over 2.5 years to evaluate its calibration accuracy and precision compared to the manufacturer’s value. A NEMA 2012/IEC 2008 phantom was filled with a 177Lu activity concentration sphere-to-background ratio of five. It was acquired with the SPECT/CT camera to determine the reconstruction parameters offering the best compromise between partial volume effect and signal-to-noise ratio. The calibration factor was computed accordingly. The calibration quality was monitored over 2.5 years with 33 phantom acquisitions with activities ranging from 7040 to 0.6 MBq. Home-made sources were used to calibrate the well counter. Its reliability was evaluated with activities ranging from 150 to 0.2 kBq measured 34 times over 2.5 years. Results: For the radionuclide calibrator, median [interquartile range] for the error on activity measurement was −0.99 [1.31] %. The optimal SPECT reconstruction parameters were obtained with 16 iterations, 16 subsets and a 12-mm Gaussian post-filter. The calibration factor was 9.87 cps/MBq with an error of −1.05 [2.12] %. The well counter was calibrated with 31.5 cps/kBq, and the error was evaluated to −12.89 [16.55] %. Conclusions: The accuracy and the precision of activity quantification using dedicated quality control were found to be sufficient for use in dosimetry implemented in clinical routine. The proposed methodology could be implemented in other centres to obtain reproducible 177Lu-based treatment dosimetry., SCOPUS: ar.j, info:eu-repo/semantics/published

After the Fukushima nuclear disaster, social concern about radiocesium (137Cs and 134Cs) contamination in food increased. However, highly efficient instruments that can measure low level radioactivity are quite expensive and heavy. A compact, lightweight, and reliable radiation detector that can inexpensively monitor low level radiocesium is highly desired. We developed a compact and highly efficient radiocesium detector to detect ~32 keV X-rays from radiocesium instead of high energy gamma photons. A 1-mm thick GAGG scintillator was selected to effectively detect ~32 keV X-rays from 137Cs to reduce the influence of ambient radiation. Four sets of 25 mm×25 mm×1 mm GAGG plates, each of which was optically coupled to a triangular-shaped light guide, were optically coupled to a photomultiplier tube (PMT) to form a square-shaped well counter. Another GAGG plate was directly optically coupled to the PMT to form its bottom detector. The energy resolution of the GAGG well counter was 22.3% FWHM for 122 keV gamma rays and 32% FWHM for ~32 keV X-rays. The counting efficiency for the X-rays from radiocesium (mixture of 137Cs and 134Cs) was 4.5%. In measurements of the low level radiocesium mixture, a photo-peak of ~32 keV X-rays can clearly be distinguished from the background. The minimum detectable activity (MDA) was estimated to be ~100 Bq/kg for 1000 s measurement. The results show that our developed GAGG well counter is promising for the detection of radiocesium in food.

The goal of this study is to evaluate the quantitative accuracy of measuring in vivo tracer concentrations using multiplexed multi-pinhole microSPECT with CT-based attenuation correction (AC) and simple methods for scatter compensation. Phantom and in vivo rat cardiac images were acquired and reconstructed with no photon AC, with AC only, with attenuation and dual-energy window scatter correction, and using a reduced attenuation coefficient to compensate for scatter. Absolute calibration was also acquired using small sources to minimize self-attenuation and scatter. The phantom tracer concentrations measured with SPECT were compared to dose calibrator measurements. The rats were sacrificed and the cardiac activity measured in vivo was compared to well counter measurements of cardiac activity. With no correction, the tracer concentrations measured in phantoms was as much as 30% below the true value for the largest phantom (52 mm diameter). AC improved the accuracy of quantification, but overestimated activity concentrations by up to 5% for the larger phantoms. With DEW scatter correction, activity concentration measured with SPECT agrees with the dose calibrator measurements to within $ -2 \pm 2\% $ and with a reduced attenuation coefficient, the agreement was better than $ -1 \pm 1\% $ . In rats injected with $^{99{\rm m}}{\rm Tc}$ -tetrofosmin, SPECT measurements of total cardiac activity were $24\pm 2\% $ below well counter measurements with no corrections, $5\pm 3\% $ above well counter measurements with only attenuation correction, $3\pm 3\% $ below with DEW scatter correction, and $1\pm 3\% $ below with reduced attenuation correction.

The purpose of this study was to assess the accuracy and practicality of well counter– and thyroid probe–based methods, commonly available in nuclear medicine facilities, for measuring the concentration of 18F-FDG in blood samples. The degree to which the accuracy of such methods influences quantitative analysis of dynamic PET scans was also assessed. Methods: Thirty-five patients with cancer of the head and neck underwent dynamic PET imaging as part of a study intended to evaluate the utility of quantitative, image-based metrics for assessment of early treatment response. The activity in blood samples from the patients, necessary to provide an estimate of the input function for quantitative analysis, was measured both using a thyroid probe and using a well counter. Three calibration techniques were compared: single-point calibration using a standard solution for the thyroid probe (ProbePoint technique), single-point calibration using a standard solution for the well counter (WellPoint technique), and multiple-point calibration over the full range of expected blood activities for the well counter (WellCurve technique). The WellCurve method was assumed to provide the most accurate estimate of blood activity. The precision of measuring blood volume using a micropipette was also evaluated by obtaining multiple blood samples. Simplified-kinetic-analysis multiple-time-point (SKA-M) uptake rates for the primary tumor were calculated for all 35 patients using PET images and each of the 3 methods for assessing blood concentration. Results: Errors in blood activity measurements ranging from −9.5% to 7.6% were found using the ProbePoint method, whereas the error range was much less (from −1.3% to 0.9%) for the WellPoint method. The precision in blood volume measurements ranged from −6% to 12% in the 10 patients assessed. The errors in blood activity and volume measurements were reflected in the SKA-M measurements in the same range. Conclusion: The WellPoint method provides a compromise between accuracy and clinical practicality. Random errors in both blood activity and volume measurements accumulate and may compromise parameters—such as the SKA-M estimate of tumor uptake rate—that depend not only on images but also on blood concentration data.

The use of a thallium-doped sodium-iodide well-type scintillation detector for the assay of the low-activity radioisotope 40K, in open-source potassium chloride aqueous solutions, is described. The hazards, safety concerns and radiowaste generation associated with using open-source radioactive isotopes can present significant difficulties, the use of hot cells and escalated costs in radioanalytical laboratory research. A solution to this is the use of low-hazard alternatives that mimic the migration and dispersion characteristics of notable fission products (in this case 137Cs). The use of NaI(Tl) as a detection medium for naturally-abundant levels of 40K in a range of media is widespread, but the use of 40K as a radioactive tracer has not been reported. The use of such low-activity sources is often complicated by the ability to detect them efficiently. In this paper a scintillator detector designed to detect the naturally-abundant 40K present in potassium chloride in tracer applications is described. Examples of the use of potassium chloride as a tracer are given in the context of ion exchange and electrochemical migration studies, and comparisons in performance are drawn from literature with hyper pure germanium semiconductor detectors, which are more commonly utilised detectors in high-resolution counting applications.

Purpose The purpose of this study is to specify a simple procedure for a robust data conversion of radioactivity value between plastic scintillator (PL) and NaI scintillator (NaI) devices. Materials and methods The radioactivity estimate of 100 blood samples was measured by the two devices. The two radioactivities were plotted on the same graph. The least-squares method was applied to obtain the conversion function. The differences between the actual radioradioy (N) from the NaI device and the estimated radioactivity for NaI (N’) from the PL device activity (P) were statistically analyzed. Results N’ was determined from P as N’ = 4.45 P + 6.28 with high correlation (r = 0.997). The Bland-Altman analysis between N’ and N showed no fixed bias and no proportional bias. Conclusions A hundred blood samples using a fixed type of sample tubes and a fixed radionuclide may be required to set up the robust conversion function.

Background In order to obtain a reliable 177Lu-DOTATATE therapy dosimetry, it is crucial to acquire accurate and precise activity measurements with the radionuclide calibrator, the SPECT/CT camera, and the NaI(Tl) well counter. The aim of this study was to determine, in a clinical context, the accuracy and the precision of their activity quantification over a range of activities and time. Ninety-three 177Lu sources from the manufacturer were measured in the radionuclide calibrator over 2.5 years to evaluate its calibration accuracy and precision compared to the manufacturer’s value. A NEMA 2012/IEC 2008 phantom was filled with a 177Lu activity concentration sphere-to-background ratio of five. It was acquired with the SPECT/CT camera to determine the reconstruction parameters offering the best compromise between partial volume effect and signal-to-noise ratio. The calibration factor was computed accordingly. The calibration quality was monitored over 2.5 years with 33 phantom acquisitions with activities ranging from 7040 to 0.6 MBq. Home-made sources were used to calibrate the well counter. Its reliability was evaluated with activities ranging from 150 to 0.2 kBq measured 34 times over 2.5 years. Results For the radionuclide calibrator, median [interquartile range] for the error on activity measurement was −0.99 [1.31] %. The optimal SPECT reconstruction parameters were obtained with 16 iterations, 16 subsets and a 12-mm Gaussian post-filter. The calibration factor was 9.87 cps/MBq with an error of −1.05 [2.12] %. The well counter was calibrated with 31.5 cps/kBq, and the error was evaluated to −12.89 [16.55] %. Conclusions The accuracy and the precision of activity quantification using dedicated quality control were found to be sufficient for use in dosimetry implemented in clinical routine. The proposed methodology could be implemented in other centres to obtain reproducible 177Lu-based treatment dosimetry.

The aim of this study is to investigate the outpatient treatment protocol and radiation safety of a new-emerging lutetium-177 ((177)Lu) prostate specific membrane antigen (PSMA) therapy. This work analyzed the dose rate of 23 patients treated with 7400 MBq (177)Lu-PSMA at different distances (0, 0.25, 0.50, 1.0 and 2.0 m) and variable time marks (0, 1, 2, 4, 18, 24, 48 and 120 h) after the termination of infusion. Blood samples were withdrawn from 17 patients within the same group at 3, 10, 20, 40, 60 and 90 min and 2, 3, 24 h after termination of infusion. Seven different patients were asked to collect urine for 24 h and a gamma well counter was used for counting samples. Family members were invited to wear an optically stimulated luminescence dosimeter whenever they were in the proximity of the patients up to 4-5 d. The total dose of the medical team including the radiopharmacist, physicist, physician, nurse, and nuclear medicine technologist was estimated by an electronic personnel dosimeter. The finger dose was determined using a ring thermoluminescent dosimeter for the radiopharmacist and nurse. The mean dose rate at 1 m after 4 h and 6 h was 23 ± 6 μSv h(-1) and 15 ± 4 μSv h(-1) respectively. The mean total dose to 23 caregivers was 202.3 ± 42.7 μSv (range: 120-265 μSv). The radiation dose of the nurse and radiopharmacist was 6 and 4 μSv per patient, respectively, whereas the dose of the physicist and physician was 2 μSv. The effective half life of blood distribution and early elimination was 0.4 ± 0.1 h and 5 ± 1 h, respectively. Seven patients excreted a mean of 45% (range: 32%-65%) from the initial activity in 6 h. Our findings demonstrate that (177)Lu-PSMA is a safe treatment modality to be applied as an outpatient protocol, since the dose rate decreases below the determined threshold of

We have been developing a medical imaging technique using a Compton camera. This study evaluates the feasibility of clear imaging with 99mTc and 18F simultaneously, and demonstrates in vivo imaging with 99mTc and/or 18F. We used a Compton camera with silicon and cadmium telluride (Si/CdTe) semiconductors. We estimated the imaging performance of the Compton camera for 141 keV and 511 keV gamma rays from 99mTc and 22Na, respectively. Next, we simultaneously imaged 99mTc and 18F point sources to evaluate the cross-talk artifacts produced by a higher energy gamma-ray background. Then, in the in vivo experiments, three rats were injected with 99mTc-dimercaptosuccinic acid and/or 18F-fluorodeoxyglucose and imaged. The Compton images were compared with PET images. The rats were euthanized, and the activities in their organs were measured using a well counter. The energy resolution and spatial resolution were measured for the sources. No apparent cross-talk artifacts were observed in the practical-activity ratio (99mTc:18F = 1:16). We succeeded in imaging the distributions of 99mTc and 18F simultaneously, and the results were consistent with the PET images and well counter measurements. Our Si/CdTe Compton camera can thus work as a multi-tracer imager, covering various SPECT and PET probes, with less cross-talk artifacts in comparison to the conventional Anger cameras using a collimator. Our findings suggest the possibility of human trials.

A new class of dose calibrator called a micro-dose calibrator, designed to measure radioactivity rates in the range between 1 kBq up to 370 kBq, has been designed and an 8 NaI(Tl)/photo multiplier tube (PMT) module prototype built to evaluate its performance. The design of the micro-dose calibrator is based on the concept of a segmented well counter. By segmenting a single crystal well counter into multiple crystals or segments, with each segment operated and read out independently, one can extend the operational range of a standard single crystal well counter while maintaining its accuracy and spectrographic capabilities. The micro-dose calibrator is operated in N-fold coincidence mode to ensure that a single radioactive decay resulting in multiple segments detecting gamma ray events are counted as a single decay. The N fold coincidence detection is achieved using a pipeline coincidence processing algorithm implemented in software as part of a multi-threaded acquisition system. A tri-functional model pulse pileup dead time correction is applied to extend the operational range of the micro-dose calibrator. The micro-dose calibrator achieved an operation range between 1 kBq up to 3.4 MBq for (18)F with a 1% linearity precision using a 60 s integration time window. Over a 9 d period, a daily count rate measure of a (137)Cs point source had a coefficient of variation of 0.17% with only one high voltage calibration performed on day 0. The central 20 mm up to 35 mm axial length in the sample well showed less than 1% change in count rate depending on the radionuclide. Using a 0.5 ml insulin syringe, a volume of dose between 0.13 ml up to 0.50 ml showed less than 1% deviation from volumetric linearity. A 20 cc scintillation vial showed similar volumetric linearity performance for radionuclide solution volumes between 4 cc up to 20 cc.

Background Although NaI(Tl) gamma counters play an important role in many quantitative positron emission tomography (PET) protocols, their calibration for positron-emitting samples has not been standardized across imaging sites. In this study, we characterized the operational range of a gamma counter specifically for positron-emitting radionuclides, and we assessed the role of traceable 68Ge/68Ga sources for standardizing system calibration. Methods A NaI(Tl) gamma counter was characterized with respect to count rate performance, adequacy of detector shielding, system stability, and sample volume effects using positron-emitting radionuclides (409- to 613-keV energy window). System efficiency was measured using 18F and compared with corresponding data obtained using a long-lived 68Ge/68Ga source that was implicitly traceable to a national standard. Results One percent count loss was measured at 450 × 103 counts per minute. Penetration of the detector shielding by 511-keV photons gave rise to a negligible background count rate. System stability tests showed a coefficient of variation of 0.13% over 100 days. For a sample volume of 4 mL, the efficiencies relative to those at 0.1 mL were 0.96, 0.94, 0.91, 0.78, and 0.72 for 11C, 18F, 125I, 99mTc, and 51Cr, respectively. The efficiency of a traceable 68Ge/68Ga source was 30.1% ± 0.07% and was found to be in close agreement with the efficiency for 18F after consideration of the different positron fractions. Conclusions Long-lived 68Ge/68Ga reference sources, implicitly traceable to a national metrology institute, can aid standardization of gamma counter calibration for 18F. A characteristic feature of positron emitters meant that accurate calibration could be maintained over a wide range of sample volumes by using a narrow energy window centered on the 511-keV peak.

Positron emission tomography (PET) has gained widespread use in cancer diagnosis and treatment, but how many malignant cells are required for a tumour to be detected by PET?Three human cancer cell lines [glioblastoma and two subtypes of small cell lung cancer (SCLC)] in concentrations from 10(4) to 10(7) were seeded on six-well plates or plastic tubes and treated with [(18)F]fluorodeoxy-glucose (FDG) in vitro. FDG retention was measured in a PET/CT scanner and in a calibrated well counter. The clinical situation was simulated using a cylinder phantom with a background concentration of FDG.The theoretical detection limit was found to be around 10(5) malignant cells. In a cylinder phantom the detection limit was increased by a factor of 10. The FDG retention by the glioblastoma cell line was significantly higher than the activity of the SCLC cell line. FDG retention measured by PET and a gamma counter was closely correlated to the number of cells and a linear relationship was found.The detection limit of PET is in the magnitude of 10(5) to 10(6) malignant cells. The experimental set-up was robust and well suited as a platform for further investigations of factors influencing the detection limit of PET.

Coincidence counting of neutron pairs is an effective way to non-destructively determine the amount of fissile material within a sample of special nuclear material (SNM) [1]. Multiplicity counting is more versatile and precise, but also more demanding, requiring the detection of three or more neutrons per single fission event. Detecting a triple coincidence of neutrons depends on the 3rd power of the detection efficiency and so on. The detection efficiency quickly becomes the critical determining factor in evaluating the utility of a particular neutron multiplicity detection technique. The purpose of this feasibility study was to characterize the performance of an LLNLdesigned water Cherenkov based Multiplicity Well counter.

Three differing exact methods of inverting the two-dimensional (2D) exponential Radon transform were implemented and evaluated quantitatively with a phantom study. The phantom had the shape of a pie-chart divided into six cavities, each 480 ml in volume and 10 cm in height, that were symmetrically positioned in a cylinder that was 20 cm in diameter and 10 cm in height. This phantom tests for linearity between true activity concentration and measured activity concentration, and it is denoted as a linearity phantom in the present study. Each cavity contained a different concentration of a homogeneous solution of 99mTc (74, 148, 222, 296, 370 and 444 kBq ml(-1)). Data acquisition was performed with two energy windows: a 20% photopeak energy window set symmetrically over the 140 keV of 99mTc and a secondary 5% energy window set over the 122 keV peak. We optimized a triple-energy window scatter correction method for a gamma camera-collimator system to obtain accurate scatter-corrected projections. A circular ROI 3 cm in diameter was identified over each cavity region, and count density (counts per pixel) was calculated. This value was converted to activity concentration (kBq ml(-1)) using a cross-calibration coefficient between SPECT counts and the gamma well counter. The relation between true activity (x) and measured activity concentration (y) was fitted to a line using the least-squares method. Regression lines were y = 0.63 + 1.0255x (R2 = 0.9987), y = -2.62 + 1.0278x (R2 = 0.9995), and y = 0.092 + 1.0241x (R2 = 0.9989) for the Bellini, Inouye and Metz-Pan methods respectively. In another phantom study using two different types of phantoms, contrast of a cold region in the two was 96% and 101% for all three methods. Combined optimized scatter correction and analytical attenuation correction methods achieve good accuracy in quantification of activity distribution with a uniform attenuating medium.

Single-photon-emitting or positron-emitting radionuclides employed in nuclear medicine are detected by using sophisticated imaging devices, whereas simpler detection devices are used to quantify activity for the following applications: measuring doses of radiopharmaceuticals, performing radiotracer bioassays, and monitoring and controlling radiation risk in the clinical environment. Detectors are categorized in terms of function, the physical state of the transducer, or the mode of operation. The performance of a detector is described by the parameters efficiency, energy resolution and discrimination, and dead time. A detector may be used to detect single events (pulse mode) or to measure the rate of energy deposition (current mode). Some detectors are operated as simple counting systems by using a single-channel pulse height analyzer to discriminate against background or other extraneous events. Other detectors are operated as spectrometers and use a multichannel analyzer to form an energy spectrum. The types of detectors encountered in nuclear medicine are gas-filled detectors, scintillation detectors, and semiconductor detectors. The ionization detector, Geiger-Muller detector, extremity and area monitor, dose calibrator, well counter, thyroid uptake probe, Anger scintillation camera, positron emission tomographic scanner, solid-state personnel dosimeter, and intraoperative probe are examples of detectors used in clinical nuclear medicine practice.

SummaryQuantitative positron emission tomography (PET) requires accurate scanner calibration, which is commonly performed using phantoms. It is not clear to what extent this procedure ensures quantitatively correct results in vivo, since certain conditions differ between phantom and patient scans. Aim: We, therefore, have evaluated the actual quantification accuracy in vivo of PET under clinical routine conditions. Patients, methods: We determined the activity concentration in the bladder in patients undergoing routine [18F]FDG whole body investigations with three different PET scanners (Siemens ECAT EXACT HR+ PET: n = 21; Siemens Biograph 16 PET/CT: n = 16; Philips Gemini-TF PET/CT: n = 19). Urine samples were collected immediately after scan. Activity concentration in the samples was determined in well counters cross-calibrated against the respective scanner. The PET (bladder) to well counter (urine sample) activity concentration ratio was determined. Results: Activity concentration in the bladder (PET) was systematically lower than in the urine samples (well The patient-averaged PET to well counter ratios for the investigated scanners are (mean ± SEM): 0.881 ± 0.015 (ECAT HR+), 0.898 ± 0.024 (Biograph 16), 0.932 ± 0.024 (Gemini-TF). These values correspond to underestimates by PET of 11.9%, 10.2%, and 6.8%, respectively. Conclusions: The investigated PET systems consistently underestimate activity concentration in the bladder. The comparison of urine samples with PET scans of the bladder is a straightforward means for in vivo evaluation of the expectable quantification accuracy. The method might be interesting for multi-center trials, for additional quality assurance in PET and for investigation of PET/MR systems for which clear proof of sufficient quantitative accuracy in vivo is still missing.

Background We propose a new approach to facilitate sentinel node biopsy examination by multimodality imaging in which radioactive and near-infrared (NIR) fluorescent nanoparticles depict deeply situated sentinel nodes and fluorescent nodes with anatomical resolution in the surgical field. For this purpose, we developed polyamidoamine (PAMAM)-coated silica nanoparticles loaded with technetium-99m (99mTc) and indocyanine green (ICG). Methods We conducted animal studies to test the feasibility and utility of this dual-modality imaging probe. The mean diameter of the PAMAM-coated silica nanoparticles was 30 to 50 nm, as evaluated from the images of transmission electron microscopy and scanning electron microscopy. The combined labeling with 99mTc and ICG was verified by thin-layer chromatography before each experiment. A volume of 0.1 ml of the nanoparticle solution (7.4 MBq, except for one rat that was injected with 3.7 MBq, and 1 μg of an ICG derivative [ICG-sulfo-OSu]) was injected submucosally into the tongue of six male Wistar rats. Results Scintigraphic images showed increased accumulation of 99mTc in the neck of four of the six rats. Nineteen lymph nodes were identified in the dissected neck of the six rats, and a contact radiographic study showed three nodes with a marked increase in uptake and three nodes with a weak uptake. NIR fluorescence imaging provided real-time clear fluorescent images of the lymph nodes in the neck with anatomical resolution. Six lymph nodes showed weak (+) to strong (+++) fluorescence, whereas other lymph nodes showed no fluorescence. Nodes showing increased radioactivity coincided with the fluorescent nodes. The radioactivity of 15 excised lymph nodes from the four rats was assayed using a gamma well counter. Comparisons of the levels of radioactivity revealed a large difference between the high-fluorescence-intensity group (four lymph nodes; mean, 0.109% ± 0.067%) and the low- or no-fluorescence-intensity group (11 lymph nodes; mean, 0.001% ± 0.000%, p < 0.05). Transmission electron microscopy revealed that small black granules were localized to and dispersed within the cytoplasm of macrophages in the lymph nodes. Conclusion Although further studies are needed to determine the appropriate dose of the dual-imaging nanoparticle probe for effective sensitivity and safety, the results of this animal study revealed a novel method for improved node detection by a dual-modality approach for sentinel lymph node biopsy.

A modular back-propagation ANN has been implemented for the non-destructive localization of a source of Even Plutonium Isotopes (EPI) contained in sealed tanks. The ANN has been trained on data obtained from a simulation of a well counter (filtered and Fourier transformed signals of the neutron detectors surrounding the well counter) for known positions of the EPI. After training, the ANN can predict the position of EPI within sealed tanks from the corresponding detector signals. The introduction of median and majority ANNs has been found to significantly improve the accuracy of prediction. Furthermore, these ANNs perform in a satisfactory manner when noise is injected to the detector signals; prediction is corrupted in a manner which is directly related to the extent and amount of noise. The motivation for using back-propagation ANNs is twofold: on one hand (theoretical importance), they are capable of learning to approximate complex functions such as the strongly non-linear relation that exists between the neutron detector signals and the EPI position; on the other hand, they accomplish on-line localization which is of practical interest.

In the course of your day, when you interpret or count a radioimmunoassay or a sample of blood or body fluid, you will probably use a variant of the well counter invented by Hal O. Anger in the early 1950s. If you obtain or read a patient image, it probably came from some type of his gamma camera. If yoy do or see a single photon emission computed tomography study, Hal Anger showed the feasibility. Even if you are working with positrons, some of the early imaging work was done by devices Hal Anger invented. This article deals primarily with the years I spent with Hal, 1962 to 1964. I was fortunate to have the opportunity to spend this time with him when he had just completed the development of his 11-in gamma camera, the first multipurpose camera suitable for diverse clinical examinations. In retrospect, it is easy to see why Hal was honored by the Society of Nuclear Medicine as one of its pioneers and why he was the first recipient of the Ben Cassen Prize. However, it was not always clear that Hal's creative road would lead to such deserved distinction. I have tried to portray the beginnings of this journey to the best of my recollections.