Radiation exposure from patients treated with 165Dy-ferric hydroxide

In radiation synovectomy about 10 GBq 165Dy-ferric hydroxide is injected into major joints. Measurements of the dose rates were performed at distances of 5 cm, 0.5 m, 1 m and 2 m from the surface of the treated joints (knees) until 200 min after the application in 16 patients in order to estimate the radiation exposure of persons in the neighbourhood of the patients. The highest doses were estimated for the fingers of the technologist (320 microSv) and for the physician (700 microSv). Special shields for the syringes were constructed for dose reduction. The whole-body doses were 103 microSv for the technologist and 40 microSv for the physician. After the discharge of the patient to a ward or home, other persons at 1 m distance from the patient might receive 88 microSv, which is less than 9% of the annual permissible dose. Our results clearly demonstrate that the calculated radiation exposure to personnel and family members is well below the maximum annual dose limit for non-professionally exposed persons.     

Occupational radiation dose associated with Rb-82 myocardial perfusion positron emission tomography imaging

BACKGROUND: We determined staff radiation dose during rest and stress rubidium 82 myocardial perfusion positron emission tomography (PET) imaging. METHODS AND RESULTS: Patients received 1,587 +/- 163 MBq (42.9 +/- 4.4 mCi) Rb-82 during rest or pharmacologic stress. A pressurized ion chamber was used to monitor radiation exposure in 50 examinations. For comparison, staff exposure during pharmacologic stress in 20 other patients receiving 1,204 +/- 55.5 MBq (32.54 +/- 1.5 mCi) technetium 99m 2-methoxy isobutyl isonitrile (MIBI) was measured. For Rb-82 infusion and PET acquisition, the mean dose was 0.45 +/- 0.25 microSv (0.045 +/- 0.025 mrem). Exposure for routine stress testing at variable distances from the patient was equivalent to background. Similar exposure for pharmacologic stress testing through 7 minutes after injection of Tc-99m MIBI at variable distances was 1.075 +/- 0.32 microSv (0.108 +/- 0.03 mrem). However, exposure for stress tests starting 7 minutes after Rb-82 infusion at 0.5 m was estimated at 0.4 microSv (0.04 mrem). To determine the potential radiation dose for those responding to a medical emergency or otherwise in close proximity to a patient, we measured the mean cumulative dose at 0.5 m from 0 to 7 minutes of Rb-82 infusion, which resulted in 19.1 +/- 5.8 microSv (1.9 +/- 0.58 mrem). CONCLUSIONS: Radiation doses for all tasks during routine Rb-82 stress-rest PET are lower than measured Tc-99m MIBI values. However, the radiation dose in close proximity to the patient during or immediately after Rb-82 infusion can be considerably higher, underscoring the need for strict attention to source distance and contact times.     

Radiation dose rates from paediatric patients undergoing 99Tcm investigations.

Infants or children undergoing nuclear medicine investigations may subsequently come into close contact with nurses or parents responsible for their care. In order to estimate the radiation dose to these individuals, and to formulate appropriate recommendations, dose rates were measured at distances of 0.1, 0.5 and 1.0 m from 148 paediatric patients who had undergone one of 12 99Tcm studies. The maximum dose rates of 70, 14 and 5 microSv h-1 at these distances were not greater than the corresponding maximum values found in an earlier study of adult patients. However, the maximum dose rates per unit activity of 0.5, 0.2 and 0.1 microSv h-1 MBq-1 were greater than the corresponding maximum 99Tcm adult values, consistent with a general increase of dose rate per unit activity with decrease of body weight observed in the paediatric measurements. A parent caring for and feeding a young infant is most unlikely to receive a dose equivalent of 1 mSv, and a nurse attending to one young radioactive patient is most unlikely to receive a dose equivalent in a working day of 60 microSv. The data obtained should allow radiation doses to be estimated and appropriate recommendations to be formulated for other circumstances, including any future legislative changes in dose limits or derived levels.     

Measurement of radiation exposure to a PET institution driver from patients injected with FDG

With the recent increase in FDG-PET examinations, concern has mounted regarding radiation exposure to hospital staff and the general public from patients injected with FDG. Because our PET institution is located 15 km from the hospital that provides these examinations, a driver has been designated to transport patients injected with FDG. This study was designed to measure the radiation dose to the driver from these patients (n=28) and to compare it with the estimated dose. A pocket dosimeter was used to measure radiation exposure to the driver. When the distances between the driver and patient were 1.1 m and 1.9 m, mean measured doses were 7.31 microSv and 2.26 microSv, respectively, while mean estimated doses were 8.61 microSv and 2.82 microSv, respectively, per trip. It was presumed that maximum radiation exposure per year was between 3.02 mSv (1.1 m) and 0.92 mSv (1.9 m). According to our data, the measured dose was 20% lower than the estimated dose. This discrepancy may be due to the difference between the volume source (measured dose) and point source (estimated dose).     

Isolation period prediction in patients with differentiated thyroid carcinoma treated after thyroidectomy by radioiodine-131

In patients with differentiated thyroid carcinoma (DTC) who after thyroidectomy underwent radioiodine-131 ((131)I) treatment for the ablation of the thyroid remnant, isolation period is considered as the period of time needed for patients' radiation dose rates to be reduced below specific adopted dose rate release limits. The aim of our study was to determine mathematical equations for the prediction of the isolation period in these patients and of their discharge from special isolation rooms in nuclear medicine departments. In order, to predict the duration of the isolation period, we studied twenty-eight patients with DTC and total thyroidectomy having no metastases, who underwent (131)I ablation treatment for a minimal residual thyroid remnant. The administered (131)I activity was 5.22 +/- 0.68 GBq, ranging from 3.66 to 6.21 GBq. Dose rates, as mean +/- SD, at a distance of 1 m from the patients were 277 +/- 44 microSv/h, immediately after (131)I administration, 72 +/- 18 microSv/h at 24 h and 23 +/- 9 microSv/h at 48 h. Whole body (131)I retention was 0.261 +/- 0.05 (range 0.168-0.385) at 24 h and 0.082 +/- 0.03 (range 0.041-0.149) at 48 h, calculated as the ratio of dose rate at 24 h and 48 h versus the initial dose rate after (131)I administration. Isolation period was calculated by a mono-exponential fitting in dose rate decay data according specific dose rate release limits. For a dose rate release limit of 30 microSv/h at 1 m, isolation period was 42.7+/-7.2 h (range 31.2-56.6 h). Seventy-five percent of the patients satisfied this limit within 48 h and 25% between 48 h and 72 h. This isolation period was positively correlated with the whole body (131)I retention and the dose rates at 24 h and 48 h, but not always with the administered activity or the initial dose rate, measured immediately after (131)I administration. On the contrary, a strong negative correlation was found between patients' isolation period and dose rate release limits between 3 and 60 microSv/h at 1 m. This study indicates that isolation period is variable but can be predicted by multiple formulas, since it depends strongly on the adopted dose rate release limits, (131)I dose rates and whole body retention at 24 h and 48 h after (131)I treatment. For a dose rate release limit of 30 microSv/h at 1 m, isolation period is sufficient for 72 h for all of our patients, while 75% of them had dose rates below that limit within 48 h.     

Radiation dose rates from patients undergoing PET: implications for technologists and waiting areas

Increasingly hospitals are showing an interest in developing their imaging services to include positron emission tomography (PET). There is therefore a need to be aware of the radiation doses to critical groups. To assess the effective whole-body dose received by technologists within our dedicated PET centre, each staff member was issued with a dose rate meter, and was instructed to record the time spent in contact with any radioactive source, the dose received per working day and the daily injected activity. On average each technologist administered 831 MBq per day. The mean whole-body dose per MBq injected was 0.02 microSv/MBq(-1). The average time of close contact (<2.0 m) with a radioactive source per day was 32 min. The average effective dose per minute close contact was 0.5 microSv/min(-1), which resulted in a mean daily effective dose of 14.4 microSv. No technologist received greater than 60 microSv (the current UK limit for non-classified workers) in any one day, and in general doses received were less than 24 microSv, the daily dose corresponding to the proposed new annual limit for non-classified workers of 6.0 mSv per annum. However, we recognise that the layout of nuclear medicine departments will not mirror our own. We therefore measured the instantaneous dose rates at 0.1, 0.5, 1.0 and 2.0 m from the mid-thorax on 115 patients immediately after injection, to provide estimates of the likely effective doses that might be received by technologists operating dual-headed coincidence detection systems, and others coming into contact in the waiting room with patients who have been injected with fluorine-18 fluorodeoxyglucose. The mean (95th percentile) dose rates measured at the four aforementioned distances were 391.7 (549.5), 127.0 (199.8), 45.3 (70.0) and 17.1 (30.0) microSv/h(-1), respectively. A number of situations have been modelled showing that, with correct planning, FDG studies should not significantly increase the effective doses to technologists. However, one possible area of concern is that, depending on the number of patients in a waiting area at any one time, accompanying persons may approach the limits set by the new UK IRR 1999 regulations for members of the public.     

Proposals for the revision of radiation protection measures for doses up to 222 MBq iodine-131 for whole body scintiscan for the detection of metastatic lesions

The goal of this study was to estimate the necessary period of time, required for radiation protection instructions to be followed by patients with differentiated thyroid carcinoma (DTC) after total thyroidectomy who are given iodine-131 ((131)I) for a whole body scintiscan (WBS) in relation to the instructions of the European Commission and the ICRP. In order to estimate and evaluate the dose received by the family members and the general public, we have studied 30 patients and were given a dose of 92-222 MBq of (131)I for a diagnostic WBS. The patients studied were four men with mean age+/-standard deviation (M+/-SD)=55+/-6 y and 26 women with: M+/-SD=47+/-14 y. Dose rate measurements were carried out at the Nuclear Medicine Department of the AHEPA University Hospital; 1 h after the patients had received the (131)I dose and 48 h later when they returned to the hospital for the WBS. The calculated doses received by the in-living relatives of the patients and by the general public, assuming that radiation protection measures were applied for 2d, ranged between 76-640 microSv and 22-171 microSv respectively. In conclusion, the results of this study, compared to the dose constraints suggested by the European Commission, indicate that the duration of radiation protection guidelines for patients receiving (131)I for diagnostic purposes could be reduced to only two days without any potential risk to family members or to members of the public. The case of children of the immediate family environment, aged less than 3 y, was not investigated in this study.     

Radiation dose to PET technologists and strategies to lower occupational exposure

OBJECTIVE: The use of PET in Australia has grown rapidly. We conducted a prospective study of the radiation exposure of technologists working in PET and evaluated the occupational radiation dose after implementation of strategies to lower exposure. METHODS: Radiation doses measured by thermoluminescent dosimeters over a 2-y period were reviewed both for technologists working in PET and for technologists working in general nuclear medicine in a busy academic nuclear medicine department. The separate components of the procedures for dose administration and patient monitoring were assessed to identify the areas contributing the most to the dose received. The impact on dose of implementing portable 511-keV syringe shields (primary shields) and larger trolley-mounted shields (secondary shields) was also compared with initial results using no shield. RESULTS: We found that the radiation exposure of PET technologists was higher than that of technologists performing general nuclear medicine studies, with doses averaging 771 +/- 147 and 524 +/- 123 microSv per quarter, respectively (P = 0.01). The estimated dose per PET procedure was 4.1 microSv (11 nSv/MBq). Injection of 18F-FDG contributed the most to radiation exposure. The 511-keV syringe shield reduced the average dose per injection from 2.5 to 1.4 microSv (P < 0.001). For the longer period of dose transportation and injection, the additional use of the secondary shield resulted in a significantly lower dose of radiation than did use of the primary shield alone or no shield (1.9 vs. 3.6 microSv [P = 0.01] and 3.4 microSv [P = 0.03], respectively). CONCLUSION: The radiation doses currently received by technologists working in PET are within accepted occupational health guidelines, but improved shielding can further reduce the dose.     

Technologist radiation exposure in routine clinical practice with 18F-FDG PET

OBJECTIVE: The use of 18F-FDG for clinical PET studies increases technologist radiation dose exposure because of the higher gamma-radiation energy of this isotope than of other conventional medical gamma-radiation-emitting isotopes. Therefore, 18F-FDG imaging necessitates stronger radiation protection requirements. The aims of this study were to assess technologist whole-body and extremity exposure in our PET department and to evaluate the efficiency of our radiation protection devices (homemade syringe drawing device, semiautomated injector, and video tracking of patients). METHODS: Radiation dose assessment was performed for monodose as well as for multidose 18F-FDG packaging with both LiF thermoluminescence dosimeters (TLD) and electronic personal dosimeters (ED) during 5 successive 18F-FDG PET steps (from syringe filling to patient departure). RESULTS: The mean +/- SD total effective doses received by technologists (n = 50) during all of the working steps were 3.24 +/- 2.1 and 3.01 +/- 1.4 microSv, respectively, as measured with ED and TLD (345 +/- 84 MBq injected). These values were confirmed by daily TLD technologist whole-body dose measurements (2.98 +/- 1.8 microSv; 294 +/- 78 MBq injected; n = 48). Finger irradiation doses during preparation of single 18F-FDG syringes were 204.9 +/- 24 and 198.4 +/- 23 microSv with multidose vials (345 +/- 93 MBq injected) and 127.3 +/- 76 and 55.9 +/- 47 microSv with monodose vials (302 +/- 43 MBq injected) for the right hand and the left hand, respectively. The protection afforded by the semiautomated injector, estimated as the ratio of the doses received by TLD placed on the syringe shield and on the external face of the injector, was near 2,000. CONCLUSION: These results showed that technologist radiation doses in our PET department were lower than those reported in the literature. This finding may be explained by the use of a homemade syringe drawing device, a semiautomated injector, and patient video tracking, allowing a shorter duration of contact between the technologist and the patient. Extrapolation of these results to an annual dose (4 patients per day per technologist) revealed that the annual extrapolated exposure values remained under the authorized limits for workers classified to work in a radioactivity-controlled area.     

Assessment of radiation safety instructions to patients based on measured dose rates following prostate brachytherapy

PURPOSE: To validate radiation safety instructions to patients and to evaluate the potential radiation doses to members of the public after (125)I or (103)Pd prostate implantation. METHODS AND MATERIALS: Radiation dose rate measurements were made in the immediate postoperative period on 636 consecutive patients with stage T1-T2 prostate cancer who underwent transperineal (125)I or (103)Pd implantation at Memorial Sloan-Kettering Cancer Center during the period from August 1995 through January 2003. RESULTS: The mean radiation dose rate at the anterior skin surface following a prostate implant was 37 microSv/hr for (125)I and 8 microSv/hr for (103)Pd. At 30 cm from the anterior skin surface, these dose rates were reduced to 6 microSv/hr for (125)I and 3 microSv/hr for (103)Pd. At 1 m from the anterior skin surface the dose rates from both types of implants were reduced to less than 1 microSv/hr. The effect of body weight on dose rates from (125)I sources was examined for a select sub-group of patients and the measured dose rate was found to decrease with increasing body weight. In another group of patients, dose rate measurements were made on both lateral skin surfaces and were less than 16.8 microSv/hr in all cases. Assuming a 33% occupancy factor and utilizing the mean measured dose rate for (125)I, the time required to reach an effective dose equivalent limit of 5 mSv for caregivers was estimated to be 19 days on contact with the skin surface. Using a similar calculation, the lifetime doses for (125)I at a distance of 30 cm from the anterior skin surface, as well as the lifetime doses for (103)Pd on contact with the skin surface and at 30 cm from the anterior skin surface can be shown to be less than 5 mSv. CONCLUSIONS: The large number of cases available for this study permits a validation of radiation safety recommendations and provides concrete information from which the permitted exposure times following implantation can be estimated. The data support the conclusion that patients treated with these implants do not represent a radiation risk to members of the public.     

Exposure of medical personnel to radiation during radionuclide therapy practices

OBJECTIVES: Radioisotopes that emit beta radiation are used for the treatment of hepatocellular carcinoma, of arthritic patients (radiosynovectomy) and treatment of bone metastases with, respectively, I-labelled lipiodol, colloidal citrate of Y or and Sm-labelled EDTMP. Radiation energy of these radioisotopes that emit beta or beta and gamma radiation (from 300 to 2000 keV) leads to an increase in radiation dose received by nuclear medicine staff. In this paper we focused on clinical and laboratory staff exposure during these types of metabolic radiation therapies. METHODS: Cylindrical LiF thermoluminescence dosimeters were used to measure radiation-related whole-body doses (WBDs) and finger doses of the clinical staff. RESULTS: Exposure of the two radiopharmacists and three nurses taking part in I-labelled lipiodol, Y-colloid and Sm-EDTMP therapies, for 12 months in succession, were 146 microSv and 750 microSv, respectively, considering WBD, and 14.6 mSv and 6.5 mSv, respectively, considering finger doses. Extrapolated annual exposures (six radiosynovectomies per year) for the rheumatologists were estimated to be 21 microSv (WBD) and 13.2 mSv (finger dose). Extrapolated annual WBDs and finger doses (25 I-labelled lipiodol treatments per year) for radiologists were estimated to 165 microSv and 3.8 microSv, respectively. CONCLUSION: Fortunately, these doses were always lower than the limits reported in the European Directive EURATOM 96/29 05/13/1996 (WBD <20 mSv.year; finger dose: 500 mSv.year) but have to be added to those relative to other metabolic radiotherapies such as radioiodine treatments and new metabolic radiotherapies (Y-conjugated peptides or antibodies). Nevertheless, the global exposure of medical staff involved in all these clinical practices justifies dosimetry studies to validate protocols and radiation protection devices for each institution.     

Finger doses for staff handling radiopharmaceuticals in nuclear medicine

Radiation doses to the fingers of occupational workers handling 99mTc-labeled compounds and 131I for diagnostic and therapeutic procedures in nuclear medicine were measured by thermoluminescence dosimetry. METHODS: The doses were measured at the base of the ring finger and the index finger of both hands in 2 groups of workers. Group 1 (7 workers) handled 99mTc-labeled radiopharmaceuticals, and group 2 (6 workers) handled 131I for diagnosis and therapy. Radiation doses to the fingertips of 3 workers also were measured. Two were from group 1, and 1 was from group 2. RESULTS: The doses to the base of the fingers for the radiopharmacy staff and physicians from group 1 were observed to be 17+/-7.5 (mean+/-SD) and 13.4+/-6.5 microSv/GBq, respectively. Similarly, the dose to the base of the fingers for the 3 physicians in group 2 was estimated to be 82.0+/-13.8 microSv/GBq. Finger doses for the technologists in both groups could not be calculated per unit of activity because they did not handle the radiopharmaceuticals directly. Their doses were reported in millisieverts that accumulated in 1 wk. The doses to the fingertips of the radiopharmacy worker and the physician in group 1 were 74.3+/-19.8 and 53.5+/-21.9 microSv/GBq, respectively. The dose to the fingertips of the physician in group 2 was 469.9+/-267 microSv/GBq. CONCLUSION: The radiation doses to the fingers of nuclear medicine staff at our center were measured. The maximum expected annual dose to the extremities appeared to be less than the annual limit (500 mSv/y), except for a physician who handled large quantities of 131I for treatment. Because all of these workers are on rotation and do not constantly handle radioactivity throughout the year, the doses to the base of the fingers or the fingertips should not exceed the prescribed annual limit of 500 mSv.     

Dose rates from patients having nuclear medicine investigations

Dose rates have been measured at 0.1 m, 0.5 m and 1.0 m from patients in a Nuclear Medicine Department. Data are presented for a variety of Nuclear Medicine procedures using doses of radiopharmaceuticals within the recommendations of the administration of Radioactive Substances Advisory Committee (ARSAC). The corresponding figures when the patient left the department, and the time average dose rate over the next 8 h were calculated. At 1.0 m the dose rates do not exceed 7.5 microSv h-1, and at 0.5 m the time average dose rate does not exceed 60 microSv h-1. Assuming that the nurse is as close to the patient as 0.1 m for 20 min in a working day, the accumulated dose over a working day would not exceed 60 microSv.     

Can hand radiation absorbed dose from radiosynomicronvectomy be high?

Preparing and injecting radiopharmaceuticals containing beta emitting radionuclides, for radiosynovectomy (RS), implies the risk of exceeding the upper limit of skin and hand radiation absorbed dose, of 500 mSv/year to both technologists, who prepare and to doctors, who inject these radiopharmacuticals. A high number of RS treatments per day lack of effective radiation protection devices and skin contamination, increase the skin radiation absorbed dose. Pronounced dosimetric and radiation protection data for radionuclides used for RS, like yttrium-90, erbium-169, rhenium-186, dysprosium-165 and holmium-166, indicate the risk and the rationale for minimizing skin radiation doses to the hands of technologists and to doctors. Hands and skin radiation exposure is mainly due to direct beta radiation from yttrium-90 containing syringes. However skin contamination, may increase this dose independently of the radionuclide used for RS. Using a syringe shield with 5 mm perspex and holding the syringe by forceps, especially for the fixation of the needle to the syringe, beta radiation exposure to the finger tips may be reduced effectively. The use of radiation-resistant gloves reduces beta radiation dose to the skin only slightly, but offers a much better protection than Latex gloves for radioactive contamination. In this article we report measurements performed by us, underlining aspects of the most effective syringe shielding applied for RS. For reducing hands beta radiation exposure during RS the following are proposed: a) To use radiation protection devices, like manipulators and perspex syringe shields and b) Special training of the personnel for the proper handling of doses and for the removal of possible contamination from beta-emitting radionuclides and c) To use beta radiation personal ring dosimeters.     

Radiation dose quantities and risk in neonates in a special care baby unit

Radiographs are taken in the neonatal period most commonly to assist in the diagnosis and management of respiratory difficulties. Frequent accurate radiographic assessment is required and a knowledge of the radiation dose is necessary to justify such exposures. A survey of radiation doses to neonates from diagnostic radiography (chest and abdomen) has been carried out in the special care baby unit of the Royal Free Hospital. Entrance surface dose (ESD) was calculated from quality control measurements on the X-ray unit itself. Direct measurement of radiation doses was also performed using highly sensitive thermoluminescent dosemeters (TLDs) (LiF:Mg,Cu,P), calibrated and tested for consistency in sensitivity. ESD, as calculated from exposure parameters, was found to range from 28 microGy to 58 microGy, with a mean ESD per radiograph of 36+/-6 microGy averaged over 95 examinations. ESDs as derived from TLD crystals ranged from 18 microGy to 58 microGy for 30 radiographic examinations. The mean energy imparted, the mean whole body dose per radiograph and the mean effective dose were estimated to be 14+/-8 microJ, 10+/-4 microGy and 8+/-2 microSv, respectively. Assuming that neonates and fetuses are equally susceptible to carcinogenic effects of radiation, which involve an overestimation of risk, the radiation risk of childhood cancer from a single radiograph was estimated to be of the order (0.3-1.3) x 10(-6). Radiation doses compared favourably with the reference values of 80 microGy ESD published by the Commission of the European Communities in 1996, and 50 microGy published by the National Radiological Protection Board in 2000.     

Dosimetry of digital panoramic imaging. Part II: Occupational exposure

OBJECTIVES: To measure occupational radiation dose during panoramic exposure from five digital panoramic X-ray units. METHODS: Exposures were made with five different digital panoramic units, of which four were equipped with a direct digital CCD (charge coupled device, "direct digital" technique), and one used storage phosphor plates ("indirect digital" technique). An anthropomorphic phantom served as the patient. An ionization chamber recorded the scattered radiation at 1 m from the phantom at five different locations around the panoramic units, both at the level of the thyroid gland and the level of the gonads, and effective organ doses were calculated. Exposure parameters were set as recommended by the manufacturers for the particular image and patient size: tube potential settings ranged between 64 kV and 74 kV, exposure times between 8.2 s and 19.0 s, tube current values between 4 mA and 7 mA. RESULTS: The maximum organ equivalent dose at 1 m from the panoramic unit was 0.60 microGy, the maximum organ effective dose was 0.10 microSv. Organ equivalent doses varied between 0.18 microGy and 0.30 microGy and organ effective doses between 0.01 microSv and 0.05 microSv for the different positions around the units (average for the different panoramic units). The variations in organ doses for the various machines were 0.04-0.53 microGy organ equivalent dose and 0.01-0.08 microSv organ effective dose. CONCLUSIONS: Assuming that 500 panoramic radiographs per year are taken by a dental practitioner at 1 m distance from the panoramic unit, he or she will receive an annual additional organ effective dose between 5 microSv and 15 microSv for the thyroid gland and between 5 microSv and 40 microSv for the gonads, depending on the type of digital panoramic unit.     

Radioimmunotherapy of non-Hodgkin's lymphoma with 90Y-DOTA humanized anti-CD22 IgG (90Y-Epratuzumab): do tumor targeting and dosimetry predict therapeutic response?

A DOTA (1,4,7,10-tetraazacyclododecane-N,N',N",N"'-tetraacetic acid)-conjugated, (111)In- and (90)Y-labeled humanized antibody to CD22, epratuzumab, was studied in patients with non-Hodgkin's lymphoma (NHL) to assess biodistribution and tumor targeting, pharmacokinetics, dosimetry, and anti-antibody response. Of particular interest was to evaluate whether pretherapy targeting and tumor dosimetry could predict therapeutic responses. METHODS: Patients received a pretherapy imaging study with (111)In-DOTA-epratuzumab IgG (0.75 mg/kg), followed about 1 wk later with (90)Y-DOTA-epratuzumab starting at a dose level of 0.185 GBq/m(2) (5 mCi/m(2)) in patients who had prior high-dose chemotherapy (group 2), and at 0.370 GBq/m(2) in patients who did not have a prior transplant (group 1), with escalation in 0.185-GBq/m(2) increments. RESULTS: The effective blood half-life for (111)In-DOTA epratuzumab was 36.1 +/- 7.9 h (n = 25) compared with 35.2 +/- 7.0 h for (90)Y-DOTA-epratuzumab (n = 22). The whole-body half-life for (90)Y-DOTA-epratuzumab estimated from (111)In-DOTA-epratuzumab scintigraphy was 58.3 +/- 4.7 h (n = 20), with urine collection confirming the loss of between 2.2% and 15.9% of the injected activity over 3 d (n = 3). One-hundred sixteen of 165 CT-confirmed lesions were visualized with (111)In-DOTA-epratuzumab. Radiation-absorbed doses to liver, lungs, and kidneys averaged 0.55 +/- 0.13 (n = 17), 0.28 +/- 0.06 (n = 17), and 0.38 +/- 0.07 mGy/MBq (n = 10), respectively, with 0.14 +/- 0.02 and 0.23 +/- 0.04 mGy/MBq delivered to the whole-body and red marrow, respectively. Tumor doses (n = 14 lesions in 10 patients) ranged from 1.0 to as much as 83 mGy/MBq for a 0.5-g lesion (median, 7.15 mGy/MBq). Group 2 patients were more likely to experience significant hematologic toxicities, but doses of up to 0.370 GBq/m(2) of (90)Y-DOTA-epratuzumab were tolerated with standard support measures, whereas patients in group 1 tolerated doses of up to 0.740 GBq/m(2) with the potential for further escalation. Anti-tumor effects were seen in both indolent and aggressive NHL. The data also suggest that anti-tumor responses of potentially equal magnitude can occur irrespective of tumor targeting and tumor size. Hence, tumor response did not correlate with the radiation dose delivered or with the tumor being visualized by external imaging. An anti-antibody response to epratuzumab was detected by an enzyme-linked immunosorbent assay in only 2 of 16 patients. CONCLUSION: These results suggest that (90)Y-DOTA-epratuzmab is a promising agent for the treatment of NHL and warrants further study. There was evidence suggesting that in this system, factors other than tumor radiation dose and targeting may be involved in the success of radioimmunotherapy.     

Trial manufacture of a plunger shield for a disposable plastic syringe

A syringe-type radiopharmaceutical being supplied by a manufacturer has a syringe shield and a plunger shield, whereas an in-hospital labeling radiopharmaceutical is administered by a disposable plastic syringe without the plunger shield. In cooperation with Nihon Medi-Physics Co. Ltd., we have produced a new experimental plunger shield for the disposable plastic syringe. In order to evaluate this shielding effect, we compared the leaked radiation doses of our plunger shield with those of the syringe-type radiopharmaceutical (Medi shield type). Our plunger shield has a lead plate of 21 mm in diameter and 3 mm thick. This shield is equipped with the plunger-end of a disposal plastic syringe. We sealed 99mTc solution into a plastic syringe (Terumo Co.) of 5 ml with our plunger shield and Medi shield type of 2 ml. We measured leaked radiation doses around syringes using fluorescent glass dosimeters (Dose Ace). The number of measure points was 18. The measured doses were converted to 70 microm dose equivalent at 740 MBq of radioactivity. The results of our plunger shield and the Medi shield type were as follows: 4-13 microSv/h and 3-14 microSv/h at shielding areas, 3-545 microSv/h and 6-97 microSv/h at non-shielding areas, 42-116 microSv/h and 88-165 microSv/h in the vicinity of the syringe shield, and 1071 microSv/h and 1243 microSv/h at the front of the needle. For dose rates of shielding areas around the syringe, the shielding effects were approximately the same as those of the Medi shield type. In conclusion, our plunger shield may be useful for reducing finger exposure during the injection of an in-hospital labeled radiopharmaceutical.     

Dose rate measurements from radiopharmaceuticals: implications for nuclear medicine staff and for children with radioactive parents

Following the introduction of a number of radiopharmaceuticals, we assessed the dose received by staff working in the nuclear medicine department and also by children who may be in close contact with a radioactive parent. We measured departure dose rates (microSv.h-1) at distances of 0.1, 0.5 and 1.0 m from the skin surface at the level of the thyroid, chest and bladder of patients undergoing the following nuclear medicine procedures: MUGA scans using 99Tcm-labelled red blood cells, myocardial perfusion scans using 99Tcm-labelled radiopharmaceuticals, lymphoscintigraphy using colloidal 99Tcm (Re) sulphide, bone scans using 99Tcm-labelled oxidronate, 111In-octreotide scans, 111In-labelled leukocyte studies and cardiac reinjection studies using 201Tl. The maximum dose rates at 0.1 m were those from MUGA studies (167.3 microSv.h-1) and myocardial perfusion studies (one-day protocol = 391.7 microSv.h-1, two-day protocol = 121.8 microSv.h-1). The implications of these dose rates on both technical and nursing staff are assessed. Also, the dose received by an infant in close contact with a parent following a nuclear medicine investigation was estimated.     

A phase I/IIa study on intra-articular injection of holmium-166-chitosan complex for the treatment of knee synovitis of rheumatoid arthritis

Previous animal studies have established that the intra-articular injection of holmium-166-chitosan complex (DW-166HC) causes effective necrosis of the inflamed synovium with litle leakage of radioactivity from the injected joint. Based on these findings, we conducted a phase I/IIa study to examine the biodistribution of DW-166HC and to assess the safety of DW-166HC for the treatment of knee synovitis in patients with rheumatoid arthritis (RA). A total of 16 patients [1 man, 15 women; median age 49 (range 36-65) years] who had RA knee synovitis refractory to disease-modifying anti-rheumatic drug treatments of > 3 months' duration were randomly assigned to three treatment groups with different radiation doses of DW-166HC: 370 MBq (n = 6), 555 MBq (n = 5) and 740 MBq (n = 5). In each treatment group, blood and urine radioactivity were analysed by beta counter and biodistribution of the injected DW-166HC was evaluated using a gamma scan camera. Clinical assessment was done according to three variables (evaluation method): knee joint pain (visual analogue scale), range of motion (goniometry) and joint swelling (circumference of knee joint). The duration of follow-up observation was 3 months. Following the intra-articular injection of DW-166HC, the blood radioactivity was little changed from the baseline measurement and the accumulated radioactivity excreted in urine was minimal. Gamma scan study indicated that most of the injected radiochemical was localized within the injected joint cavity, and the extra-articular leakage was negligible at 24 h after the injection: brain, 0.3%; lung, 0.6%; abdomen, 0.7%; and pelvis, 0.8%. Major adverse events were transient post-injection knee joint pain and swelling. These results suggest that DW-166HC might be a safe agent for radiation synovectomy, particularly for the treatment of knee synovitis of RA, and further trials in a larger patient population are warranted to evaluate the therapeutic efficacy of DW-166HC.