Radioactivity measurement with a plastic scintillation vial

Liquid scintillation counting is widely used to measure radioactivity, but it generates radioactive organic liquid waste. Not to generate the liquid waste using a liquid scintillation counter, novel counting method with a plastic scintillation vial was designed. The counting efficiency for ³²P was 10–40% and that for ¹²⁵I was 4–8%. The efficiency depended on the sample volume. The color quenching effect was negligible. No radioactive liquid waste was generated by this method. In addition, you can reuse the measured sample.     

Rapid determination of131I in environmental waters using a liquid anion exchanger

A rapid and specific method for the determination of¹³¹I in environmental water samples in the presence of some of the most important fission products is described. Radioiodine is separated from acidified water using tri-n-octylamine solution in toluene with dissolved iodine by one-stage static procedure with about 90% separation efficiency and 200-fold volume concentration. After the decolorization of the organic phase with NaOH in methanol, radioactivity of¹³¹I is counted by a toluene base liquid scintillator with a counting efficiency of 70%. The method is simple and enables to determine low radioactivity¹³¹I with a detection limit less than 5 pCi/1 in about 1.5 hrs.     

Radiolabeling of EGCG with 125I and its biodistribution in mice

The aim of the present study was to label EGCG with ¹²⁵I and to determine its radiopharmaceutical potential in mice. EGCG was labeled with ¹²⁵I using the iodogen method. The labeling yield and the radiochemical purity of ¹²⁵I–EGCG were determined by radio thin-layer chromatography (RTLC). The Labeling yield was approximately 89.4 %. The radiochemical purity was approximately 96.4 %. The biodistribution studies of the labeled compound (specific activity; 0.47 TBq/μg) were performed in male Kunming mice. The uptakes of ¹²⁵I–EGCG in some organs were determined at different time after injection to the mice. The radioactivity in each organ was counted and the percentage of injected activity per gram of tissue weight (%ID/g) for each organ and blood was calculated. Incorporation of radioactivity in the various tissue/organ was confirmed by microautoradiography. ¹²⁵I–EGCG uptake in the stomach and salivary gland was higher than other organ/tissue. The black silver grains was concentrated in the nucleus, cytoplasm, intercellular substance and capillaries of that various organs, and its unevenly distributed. Thus, ¹²⁵I–EGCG may be radiopharmaceutical for the imaging of the stomach and salivary gland.     

Incineration system for the disposal of 125I-radioimmunoassay test tube wastes--confirmation of 125I decay in wastes

Before incineration of disposed RIA test tubes, residual 125I radioactivity should be measured for the purpose of radiation safety. This confirmation method has been carried out according to the followings: Residual 125I radioactivity of test tubes in the plastic bag was measured from outside with the low-energy gamma-ray survey meter (S1371, Ohyo Kohken CO., Ltd.) newly developed by one of the authors. Gaseous 125I released from the residual 125I-compounds in test tubes by radiation self decomposition was adsorbed by the activated carbon fiber filter (ACFF), and was measured. From the above 2 methods, the minimum detectable concentration of residual 125I in waste tubes would be reached under a level of about 100 mBq/g (2.8 X 10(-6) muCi/g).     

A rapid method for estimating iodine-125 in water samples using X-ray fluorescence for yield correction

Large sample volumes are needed when monitoring low levels of¹²⁵I in fresh water systems. Small anion exchange resin columns concentrate the radioactivity and also act as counting sources in thin NaI(T1) well scintillation detectors. Isotope identity is confirmed by decay counting or Ge spectroscopy. Recovery corrections are made by comparing the X-ray fluorescence of stable iodide carrier added initially with that retained on the counting column. The benefits from this simple, quick and accurate procedure could be adapted to other radiochemical estimations which rely on carrier recovery.     

Speciation and estimation of radioiodine in reactor coolant water

The species of radioiodine in the primary coolant water of the heavy water moderated, heavy water cooled 100 MW research reactor have been identified. It was observed that IO ₃ ^? was the major species in the reactor coolant during reactor operation and I^? was the major species during shutdown. Organic and elemental forms amount only to less than 2% of total radioiodine. A simple method was developed for the estimation of gross iodine activity in reactor coolant water. The method involves the separation of all inorganic forms of iodine into a photographic film consisting of a thin layer of silver halide. The iodine in the film was estimated by gross counting of the film in a Geiger-Müller counter. Gross iodine activity in the reactor coolant samples estimated by the present method were in agreement with that obtained by direct γ-spectrometry with a Ge detector. It is concluded that the method can be used for the routine estimation of radioiodine in reactor coolant water.¹³³I/¹³¹I and¹³⁵I/¹³¹I ratios in the film were estimated and found to be useful in identifying split rod conditions in the reactor.     

Thyroidal 125I monitoring system using a survey meter for 125I

This paper describes the counting efficiency and detection limit of a thyroidal 125I monitoring system. Two systems were used: (1) M1 was composed of only a survey meter for 125I (SM) having a NaI(T1) crystal of 2 in phi X 5 mmt and (2) M2 was composed of the SM having an output terminal for spectroscopy and a multichannel pulse height analyzer. The counting efficiency was determined by using an anthropomorphic neck phantom embedded four simulated thyroids of 17, 20.5, 31 and 40 ml including 125I solution. The counting efficiency between 0 degrees and 45 degrees to the direction from the center of the thyroid to the front of the neck coincided within -4%. The efficiency of M1 ranged from 7.9 to 1.8% as the distance between the probe and the neck increased from 0 to 5 cm. Similarly the efficiency of M2 ranged from 8.3 to 2.2%. The detection limit of M1 ranged from 7 to 34 Bq, and the limit of M2 ranged from 1 to 5.1 Bq. M2 system was applied to monitoring a worker performing iodination with Na125I of 74 MBq. Both monitoring systems proved to sensitively detect thyroidal 125I within the uncertainty +/- 10%.     

Preparation of tritium standard gas by the use of tritiated methane

Tritium standard gas for calibrating radioactivity detectors was prepared by the use of tritiated methane synthesized by the reaction of tritiated water with aluminum carbide at ca. 470 degrees C. A long-path proportional counter was used to standardize the radioactivity of the prepared gas. Tritiated gas of less than 10(4) Bq could be measured by this counter with the overall error of 3.5% (3 sigma). Finally, two types of standard were prepared in a metal cylinder and a glass ampoule. The latter type could be used as a handy calibration source for detectors.     

Monitoring radioactive compounds in high-performance liquid chromatographic eluates: fraction collection versus on-line detection

This study compared two methods of monitoring radioisotopes in high-performance liquid chromatographic eluates (on-line radioactivity detector versus fraction collection and counting). Testing was accomplished by pumping solutions of tritiated water in acetonitrile--water mixture through the detector or to the fraction collector. At most solvent compositions, the detector's counting efficiency and detection limits were poorer than those of the scintillation counter. However, the reproducibility of the detector data was superior at acetonitrile concentrations of less than 50%. This was attributed to the difficulty in collecting fractions of small equal volumes at the lower organic solvent concentrations in short time intervals. We conclude that on-line monitoring with homogeneous detection is the preferred method for detecting radiolabeled compounds in high-performance liquid chromatographic eluates.     

Separation of 3H, 14C, 32P, 35S, 125I, and 131I in radioisotope waste

Measurement of ³H, ¹⁴C, ³²P, ³⁵S, ¹²⁵I, and ¹³¹I in radioisotope (RI) waste materials such as the vials, pipette tips, tubes, syringes, and paper generated from the industrial, medical, educational, and research organizations were conducted by a wet oxidation method. Counts were obtained by a liquid scintillation counter for ³H, ¹⁴C, and ³²P; a gas proportional counter for ³⁵S; a low energy photon spectroscopy for ¹²⁵I; and an HPGe detector for ¹³¹I. After the treatment of approximately 20 g of the sample, the counting value was determined to obtain a minimum detectable activity (MDA) of approximately 1 × 10^?3 ~ 5 × 10^?2 Bq/g. The specific activities of shor-half-life RIs (³²P, ³⁵S, ¹²⁵I, and ¹³¹I) were not detectable and/or resulted in a low value (<1 Bq/g). The waste containing ³H and ¹⁴C was observed to have the specific activities in the range of 10^?2–10⁵ and 10^?2–10⁴ Bq/g, respectively.     

外周苯二氮受体示踪剂的合成和评价

合成了用作外周苯二氮受体潜在的选择性配体的N,N-二乙基-2-(4-碘苯基)-6-三氟甲基-咪唑并[1,2-a]吡啶-3-乙酰胺(ITFZOL). 其放射性标记物[125I]ITFZOL通过碘脱锡化反应制备, 放化得率75%~85%, 比活度大于76 GBq/μmol. 小鼠尾静脉注射[125I]ITFZIOL后, 放射性集中分布于肾上腺、肺、肾、心、嗅球和小脑等外周苯二氮受体高密度区域. 预先给与外周苯二氮受体选择性配体PK11195明显减少外周苯二氮受体高密度区域放射性分布, 提示[125I]ITFZOL对外周苯二氮受体具有较高的特异亲和性. 生物活性数据表明, [125I]ITFZO是一种潜在的选择性外周苯二氮受体单光子放射性配体.     

Direct counting of soil wafers: An improved total alpha/beta screening analysis

Although potentially very useful as a screening tool, currently-applied “gross” alpha/beta analyses are often considered unreliable by the environmental monitoring community. We describe here an alternative approach based on direct counting of pressed soil “wafers” to estimate total alpha/beta activities. The system was calibrated using a series of natural soils and sediment standards with total α/β activities estimated by use of a combination of available certified values, equilibrium assumptions, and our own measurements. A set of 10 such standards that span a wide range of activities were prepared by thoroughly mixing several grams of selected NIST and IAEA natural matrix standards together with reagent grade cellulose in a 4∶1 sample-to-binder ratio and pressing in a 40-mm stainles steel die. The resulting wafers, assumed to be at infinite thickness for expected radionuclides, were counted in a gas flow proportional counter set for simultaneous α/β counting. Both the alpha and beta count rates increased in a linear, systematic manner with increasing total estimated activities. This technique should prove to be an inexpensive, simple, and waste-free approach for screening total radioactivity in soil samples.     

Measurement of neutrons from an electron accelerator by rem-counter

In the measurement of neutrons from the medical electron accelerator by a rem counter, two problems disturb accurate measurements. One is the pile-up of signals produced by X-rays during each X-ray burst and the other is the increased counting loss caused by bunched nature of yielded neutrons. The time spectrum of neutrons measured by the rem counter 2202 D (manufactured by Studsvik) rises up to a maximum value by about 20 microseconds and then falls down exponentially with a time constant of about 90 microseconds. On the other hand, that of X-rays is roughly rectangular with several microseconds width. A time discriminating system was prepared to be combined with the rem counter, which was triggered by leading edge of electron beam pulses, rejected pile-up signals due to X-ray bursts, and counted pulses of neutrons in a specified time window. The system discriminated the pile-up enough to measure neutrons at a X-ray dose rate of at least 30 mGy/h. Nonparalyzable counting loss correction was practicable upto about 10 mSv/h for the beam pulse rate of 85 Hz, in which the dead time of the rem counter was estimated as 4 microseconds.     

A comparison of classical 90Sr separation methods with selective separation using molecular recognition technology products AnaLig? SR-01 gel, 3M Empore? Strontium Rad Disk and extraction chromatography Sr?Resin

We studied the use of an extraction chromatography for determination of ⁹⁰Sr in contaminated water samples. The aim of our work was to compare selected products from the point of view of the strontium chemical yields and analysis time. Three commercial products, 3M Empore? Strontium Rad Disk, AnaLig^® Sr-01 gel, Sr^®Resin, and two classical methods, liquid?Cliquid extraction with tributhylphosphate and carbonate co-precipitation, were tested for the separation of ⁹⁰Sr. The water sample from nuclear power plant A1 Jaslovske Bohunice was used for radiochemical analysis of ⁹⁰Sr volume activity. Samples were traced with ⁸⁵Sr to monitor strontium chemical recovery and counted either by Cerenkov counting on TRI CARB 2900 TR liquid scintillation counter or low level alpha?Cbeta proportional counter.     

Mikrobestimmung von Kohlenstoff und Wasserstoff,sowie von 14C,Tritium und Deuterium im Stickstoffstrom

A microanalytical method for the determination of carbon and hydrogen is described. The procedure involves the combustion of the organic substance in a stream of nitrogen and the oxidation of the decomposition products with oxygen. Halogens and sulphur are absorbed by the combustion catalyst (according to Körbl), whereas nitrogen oxides are reduced on a copper surface. The combustion products, CO₂ and H₂O are determined gravimetrically; in the case of radioactive samples they are appropriately absorbed and counted in a liquid scintillation counter. Strongly quenching singly and doubly labelled samples, which cannot be accurately counted even by modification of the counting device, give equally high radioactive yields (75% for ¹⁴C, 30% for ³H), since counting is not influenced by other gaseous impurities. Deuterium is quantitatively determined by measuring the intensity of the OD-signal at 2500 cm^?1 with an infrared spectrometer.     

Rapid determination of 90Sr/90Y in water samples by liquid scintillation and Cherenkov counting

Strontium-90 (⁹⁰Sr) is a ubiquitous contaminant at nuclear facilities, found at high concentrations in spent nuclear fuel and radioactive waste. Due to its long half-life and ability to be transported in groundwater, an accurate method for measuring ⁹⁰Sr in water samples is critical to the monitoring program of any nuclear facility. To address this need, a rapid procedure for sequential separation of Sr/Y was developed and tested in groundwater samples collected from an area of riverbed affected by a ⁹⁰Sr groundwater plume. Sixteen samples, plus spike and water blanks, were analyzed. Five different measurements were performed to determine the ⁹⁰Sr and yttrium-90 (⁹⁰Y) activities in the samples: direct triple-to-double-coincidence ratio (TDCR) Cherenkov counting of ⁹⁰Y, liquid scintillation (LS) counting for ⁹⁰Sr following radiochemical separation, LS counting for ⁹⁰Y following radiochemical separation, Cherenkov counting for ⁹⁰Y following radiochemical separation and LS counting of the Sr samples for ⁹⁰Y in-growth. The counting was done using a low-level Hidex 300SL TDCR counter. Each measurement method was compared for accuracy, sensitivity and efficiency. The results following Cherenkov counting and radiochemical separation were in very good agreement with one another.     

Hydrodynamic optical alignment for microflow cytometry

A microfabricated flow cytometer has been developed that is capable of detecting nearly all of the microparticles in an aqueous suspension. Current design allows for integrated coupling between an optical fiber-based detection system and the particle stream via hydrodynamic focusing. By adjusting the relative flow-rates at the auxiliary inputs of the focusing manifold, the particle stream can be steered out-of-plane relative to the illuminating laser, and similarly the particle stream can be squeezed or expanded. The microfabricated device was constructed in polydimethylsiloxane with cross-sectional microfluidic dimensions of 125 μm×125 μm. Using the present device and method, fluorescent microparticles in aqueous solution were counted at an absolute counting efficiency of 91±4%. The coefficient of variation of the fluorescence pulse-heights for far-red fluorescent microparticles was 15%. The device exhibited a linear response to fluorescence intensity calibration microparticles as shown by comparison with a commercial cytometer instrument.     

A new time interval analysis system for the on-line monitoring of artificial radionuclides in airborne dust using a phosfitch type alpha/beta detector

A portable system for high-speed time interval analysis (TIA) method was constructed. The system installed with an original time counting board could record pulse incident time of 1 MHz precisely. During the measurement, calculation of time intervals for all the pulses and on-line display of the results were performed simultaneously. Using this system, the correlated decay events of ²¹⁴Bi to ²¹⁰Po in the airborne dust collected on filter paper were counted, together with a dual-phosphor type alpha/beta detector. A rapidly detective technique of artificial radioactivity under the variations of naturally occurring radioactivities is discussed.     

Further study on NAA in characterizations of reference materials

A program was initiated at Chalk River Laboratories (CRL) to determine the physical, chemical and radiological properties of wastes intended for disposal in IRUS (Intrusion Resistant Underground Structure), a below ground vault to be constructed at CRL. One of the most restrictive radionuclides for IRUS is¹²⁹I, which has been assigned a maximum activity concentration in waste of 10⁶ Bq/m³. The limit of detection for radionuclides in waste has been set at 1% of the approximate maximum activity concentration, or 10⁴ Bq/m³ for¹²⁹I. A radiochemical instrumental neutron activation analysis method has been developed to determine¹²⁹I in two waste streams, incinerator ash and liquid feed to a bituminizer. Solid samples are spiked with¹²⁵I tracer, fused at 960°C with Li₂B₄O₇ in a platinum boat in a flowing oxygen stream inside a three zone tube furnace, and the volatilized I₂ is trapped on in-line charcoal filters. The charcoal filters are irradiated together with a filter containing a spiked¹²⁵I/¹²⁹I standard, in the NRU reactor, and then subjected to post-irradiation chemistry to remove⁸²Br interference. The¹²⁹I concentration in the sample is determined by comparing the activity of the activated¹³⁰I in the sample with that of the standard, and the chemical recovery for¹²⁹I is determined from the activity of¹²⁵I tracer. Limits of detection for¹²⁹I in solids are typically 0.005 Bq/g, based on a 4 hour counting period on a 10% efficient HPGe gamma-spectrometer at a source to detector distance of approximately 12 cm. This paper presents a summary of the method and the results from analysis of two waste streams.     

Implementation of the coincidence method for determining125I activity with an estimate of error

The established coincidence method of determining¹²⁵I activities has been adapted for use with a multisample gamma counter. The method has been extended to include a formula which has been derived to indicate the precision of the measured activities. The validity of this formula has been tested and found to be satisfactory. The techniques used to select appropriate counting conditions are described.