WO1996035131A1 - Dispositif de mesure destine a la detection non-invasive des radionuclides - Google Patents

Dispositif de mesure destine a la detection non-invasive des radionuclides Download PDF

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Publication number
WO1996035131A1
WO1996035131A1 PCT/US1996/006322 US9606322W WO9635131A1 WO 1996035131 A1 WO1996035131 A1 WO 1996035131A1 US 9606322 W US9606322 W US 9606322W WO 9635131 A1 WO9635131 A1 WO 9635131A1
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WO
WIPO (PCT)
Prior art keywords
detector
sample
output
photon
electrical signal
Prior art date
Application number
PCT/US1996/006322
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English (en)
Inventor
Arata Suzuki
Marcia Suzuki
Original Assignee
Capintec, Inc.
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Capintec, Inc. filed Critical Capintec, Inc.
Publication of WO1996035131A1 publication Critical patent/WO1996035131A1/fr

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    • G—PHYSICS
    • G01—MEASURING; TESTING
    • G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
    • G01T7/00—Details of radiation-measuring instruments
    • G01T7/02—Collecting means for receiving or storing samples to be investigated and possibly directly transporting the samples to the measuring arrangement; particularly for investigating radioactive fluids
    • G—PHYSICS
    • G01—MEASURING; TESTING
    • G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
    • G01T7/00—Details of radiation-measuring instruments

Definitions

  • This invention relates to a non-invasive activity meter for measuring pure beta emitting samples of known radio nuclides.
  • Sr-89 Phosphorous-32 (P-32)
  • Y-90 Yttrium-90
  • Sr-89 chloride injection has been developed and approved for treatment of patients who are experiencing extremely severe pain from cancer which is metastasized to the bones.
  • the Sr-89 chloride injection is used to palliate such pain.
  • Radioactive pharmaceuticals must be non-invasively, i.e., non-distractively determined accurately, and also immediately prior to administration of the pharmaceuticals to patients.
  • an ionization chamber type dose calibrator is used to determine the dose of gamma ray emitting radioactive pharmaceuticals.
  • the ionization chamber type calibrator is not suitable to accurately determine the therapy dose of pure beta emitting radio-pharmaceutical immediately prior to administration to the patient.
  • Ionization chamber type calibrators do not have the capability of discriminating photon energy other then to filter very low energy photons from entering the sensitive volume of the ionization chamber.
  • Such a low energy photon filter will produce a very low output signal from the ionization chamber for the beta ray emitting sample to be tested but also increases the relative sensitivity for impurities in the sample to unacceptable levels. For example, if a sample of Sr-89, which contains .1% of an impurity such as Sr-85, is measured by an ionization chamber type calibrator with a low energy photon filter, approximately 10% of the output current would be attributable to the Sr-85 impurity. Therefore, if the contamination level is changed by .1% it would result in a change in the activity determination of Sr-89 of approximately 10%.
  • Radio nuclei samples such as Sr-89 which are commercially supplied may contain up to .4% of Sr-85 and up to .4% of other high energy gamma ray emitting impurities which would distort the measurements. Since the level of contamination may change substantially, precise measurement of the Sr-89 could not be obtained by the ionization chamber type calibrator.
  • the substantial portion of the output signals from an ionization chamber are thus generated by those impurities, since very low levels of low energy radiation are associated with beta particles being stopped in the sample, sample container, sample holder and the wall of the ionization chamber.
  • very low levels of low energy radiation are associated with beta particles being stopped in the sample, sample container, sample holder and the wall of the ionization chamber.
  • Yttrium 90 which is a relatively high energy beta ray emitter
  • the observed variations due to sample configurations range more than 30% of the measured activities when the samples are measured by the same calibrator which is the type most commonly used in medical institutions are known in prior art.
  • a device which measures samples of radio nuclei pharmaceuticals in a non-invasive, accurate, and simple manner in a short period of time at the place and the time of the administration of such pharmaceuticals to patients. Additionally, a device is needed which may determine the activities of low level gamma ray or x-ray emitting samples of known nuclei, non-invasively, precisely, accurately, simply and quickly.
  • the present invention includes a radio nuclide detector for measuring proper dosage of a radio nuclide sample.
  • the detector has a sample holder which has a sample container which holds the radio nuclide sample.
  • a radiation detector is located in proximity to the sample holder. The radiation detector is capable of producing an electrical signal related to the radiation energy emitted by the sample.
  • the radiation detector may be a.scintillation detector having a container holding scintillation phosphor.
  • the container has a first window proximate to the sample container and a second window.
  • the surface area of the first window is such to optimize the ratio of the volume of the detector to the area of the first window to increase accuracy of the measurement.
  • a photo detector which produces an electrical signal in response to scintillation emitted by the scintillation phosphor is located near the second window.
  • a counting circuit is coupled to the radiation detector which counts photon emissions from the sample within a specified energy range.
  • the counting circuit may be a pulse shaping amplifier coupled to the radiation detector.
  • the output from the amplifier is coupled to one input of a first comparator and is compared to a voltage reference related to a certain radiation energy level.
  • a counter is coupled to the output of the comparator and outputs a signal which is related to the count of photon emissions from the sample within the specified energy range.
  • Other comparators and counters may be employed for detecting different energy levels.
  • a divider may be coupled to the output of the pulse shaping amplifier.
  • the divider divides the output of the pulse shaping amplifier.
  • a low gain pulse height unit is coupled to the divider and. the output is run to an analog to digital converter to produce a digital signal.
  • a first in first out memory is connected to the analog to digital converter. The first in first out memory has an output which is related to the counts of photon emissions from the sample within a specified energy range.
  • FIG. 1 is an activity meter according to the present invention
  • FIG. 2A is a partial cutaway view of the detector assembly and a section of a sample holder according to the present invention holding a syringe;
  • FIG. 2B is a cutaway view of the sample holder holding a vial;
  • FIG. 2C is a cutaway view of the sample holder holding a test source
  • FIG. 2D is a cutaway view of the sample holder holding a NIST ampoule
  • FIG. 3 is a circuit diagram of a six channel pulse height analyzer according to the present invention.
  • FIG. 4 is an alternate embodiment of a multichannel pulse height analyzer according to the present invention.
  • FIG. 5 is a main measurement screen according to the present invention.
  • FIG. 6 is a counter screen according to the present invention.
  • FIG. 7 is a result screen according to the present invention.
  • FIG. 8 is a bar graph screen according to the present invention
  • FIG. 9 is a calculated external Bremsstrahlung spectrum for a typical beta ray emitter
  • FIG. 10A is an energy diagram of a Sr-89 sample taken with a common scintillation detector.
  • FIG. 10B is an energy diagram of a Sr-89 sample taken with an optimized detector according to the present invention. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT
  • FIG. 1 shows a radiation activity meter 10 according to the present invention.
  • the activity meter 10 consists of the two main units, a detector unit 12 and a control unit 14.
  • Control unit 14 is coupled to detector unit 12 via a triaxial high voltage cable..16 which supplies high voltage to the detector and transmits back the detected signal from the sample for further processing in the control unit 14.
  • the control unit 14 has a key pad 18 which allows a user to select various features.
  • a LCD screen 20 in conjunction with keypad 18 allows a user to control the measurements of a sample held in the detector unit 12.
  • the detector unit 12 has an access 22 for inserting the container sample to be measured.
  • FIG. 2A shows insertion of a syringe 24 in the access 22.
  • the syringe 24 is filled with the radioactive-nuclei sample to be measured.
  • the syringe 24 has a body 26, a plunger 28, and a sheathed needle 30.
  • the syringe 24 is guided into place by means of a syringe guide 32, which has an access 34 through which the body 26 of the syringe 24 may be snugly fit.
  • a removable vial guide 36 serves to hold the upper section of body 26 of the syringe 24 in place.
  • a bottom guide 38 has an access 40 which serves to hold a needle clamp 29 of the syringe 24.
  • a scintillation detector 42 is placed in horizontal alignment with the radio nuclei substance contained in the body of the syringe 24 or other samples so that the distance between the detector and the sample will be independent of the sample volume.
  • the detector 42 is held by a collimator 44.
  • the detector 42.. is shielded from ambient radiation by collimator 44 and shield 46 which is surrounding the sample.
  • Shield 46 is preferably .25 inch (6mm) thick and made of lead which serves to prevent ambient radiation from interfering with the measurement.
  • a cylinder 48 is capped by syringe guide 32 and bottom guide 38 is inserted within cylinder 48.
  • the difference in the distance between the sample contained in a sample container such as syringe 24 and the detector 42 will have a significant effect on the precision of the sample measurement. In the preferred embodiment this distance is approximately 100 mm. A change of even 1mm in distance will change the activity measurement by 2% since the solid angle between the detector and the sample is inversely proportional to the square of the distance between the detector and the sample.
  • the syringe guide 32, bottom guide 38 and vial guide 36 reproduce sample location to within 0.5 mm from the center of the sample container.
  • the sample holder 12 may be utilized to measure the radio nuclei substances contained in a variety of different containers such as syringe 24. FIG.
  • FIG. 2B shows a 10 milliliter (ml) vial 50 which may be held in place by vial guide 36.
  • a radio nuclide may be contained by vial 50 and placed in the sample holder 12 for testing according to the present invention.
  • FIG. 2C shows a 30 ml test source 52 which is held by the cylinder 48 and bottom guide 36 of the sample holder 12.
  • FIG. 2D shows.a NIST ampoule 54 which may be held by the vial guide 36 of the sample holder 12.
  • other configurations which may hold radio nuclide substances may be tested using the techniques of the present invention.
  • the control unit 14 may also be coupled to a printer (not shown) via a standard RS-232 cable for printing results of testing samples and information " on the display 20.
  • a personal computer (not shown) may also be coupled to control unit 14 for further analysis of data taken from the sample.
  • FIG. 3 is a first embodiment of a six channel pulse height analyzer circuit used for testing the presence of impurities within the sample according to the present invention. Of course, different numbers of channels may be utilized in order to obtain greater differentiation of counts for radio nuclide samples.
  • the circuit shown in FIG. 3 is primarily located in the control unit 14.
  • the radio nuclei sample such as Sr-89 emits gamma and beta rays which are measured by the spectrum analyzer as will be explained below.
  • the detector 42 consists of a volume 56 containing scintillation phosphor, whose size is optimized for beta ray detection.
  • the scintillation phosphor volume 56 has a reflector 57 and a protective envelope 60.
  • the scintillation phosphor in volume 56 is preferably Thallium activated sodium iodide, (Nal (Tl)) which scintillates at .4nm (blue) with a decay constant of .23 microseconds.
  • the photo sensitive detector 58 may be a photomultiplier (PMT) , micro-channel plate, photo diode or similar device.
  • the scintillation phosphor in volume 56 emits light at an intensity proportional to the energy lost by ionizing radiation in the phosphor. In this case, the emission of light reflects the energy emitted by photon production from the sample contained in syringe 24.
  • the photodetector 58 converts scintillation to electrical signals for further processing.
  • a wide range of output pulses may be obtained from photodetector 58. In the preferred embodiment, the output pulse width is set at l microsecond (.5 ⁇ s-3 ⁇ s range) for a typical Nal (Tl) detector. Of course other ranges may be set depending on the desired application.
  • the high voltage is supplied to the photodetector
  • the output of the detector 42 is coupled through a capacitor 71 to the input of a pulse shaping amplifier 68.
  • the pulse shaping amplifier 68 has a feedback loop consisting of a resistor 70 and a capacitor 72.
  • the pulse shaping amplifier 68 amplifies the signal from detector 40. This signal is proportional to the energy deposited in the scintillator volume 56 from the photon emission.
  • the output of the pulse shaping amplifier 68 is coupled to inputs of a series of comparators 74.
  • the other inputs of the comparators 74 is coupled to a resistor voltage divider network 76 which is coupled between a voltage source and ground. Resistors 78 in resistor network 76 combine with the voltage source to provide different reference voltages.
  • the six comparators 74 differentiate between six different voltage levels.
  • the outputs of the comparators 74 comprise six different channels for the output of the pulse shaping amplifier 68.
  • the threshold of channel 1 corresponds to photon energies of 30 keV
  • channel 2 corresponds to 100 keV
  • channel 3 corresponds to 200 keV
  • channel 4 to 400 keV corresponds to 100 keV
  • channel 5 to 660 keV corresponds to 800 keV.
  • different ranges may be designated.
  • Each of the outputs of comparators 74 are connected to a counter unit 80 which counts the number of pulses on each channel, representing the number of photons whose energies are higher than the threshold energy.
  • the output of counter units 80 are coupled into a bus 82.
  • the bus 82 is connected to a CPU 84, a programmable read only memory (PROM) 86, a random access memory (RAM) 88, and an electrically erasable programmable read only memory (EEPROM) 90.
  • the bus 82 allows data and instructions to be shared between CPU 84, PROM 86, RAM 88, and EEPROM 90.
  • Bus 82 may • also be connected to peripheral devices such as a storage means or printer through appropriate interfaces.
  • the CPU 84 performs the analysis and output functions to be displayed on the display 20 of the control unit 14.
  • the program for analyzing and recording data is stored in the PROM 86 and loaded in RAM 88 for use by CPU 84.
  • RAM 88 is also utilized to store the results of the counters 80 for further analysis.
  • the CPU 84 is a high performance, low power 16 bit microcontroller, for example a National Semiconductor model HPC46003V20 operated at a frequency of 11.06 MHz.
  • the PROM 70 is preferably 4 megabytes divided into one 16 kilobyte non-banked basic program block, and one 16 kilobyte and fifteen 32 kilobyte banked program blocks.
  • a diagnostic program may be run whereby the PROM 86 is tested block by block via a cyclical redundancy test after confirming the functioning of a power source.
  • the RAM 88 is 0.5 megabytes. Long term but alterable data is stored in non-volatile EEPROM 90 for use during the measurement and analysis operation which .will be described below.
  • a second embodiment of a circuit used to analyze the output from detector 42 is shown in FIG. 4.
  • the detector 42 and pulse shaping amplifier 68 are the same as the circuit shown in FIG. 3.
  • the output of the pulse shaping amplifier 68 is coupled to a divider 92 which is used to set the gain of the pulse height analyzer.
  • the output of divider 92 is coupled to a linear gate and peak holding circuit (LG/PH) 94.
  • the output of the LG/PH 94 is coupled to an analog digital converter (ADC) 96.
  • the ADC 96 converts the signal from the LG/PH 94 into a digital signal.
  • ADC 96 has a precision of 14 bits or higher.
  • the output of ADC 98 is stored in a first in first out memory (FIFO) 98.
  • FIFO first in first out memory
  • the FIFO 98 serves to store the pulse height, i.e., the energy of the detected photons by the detector 42.
  • the output of FIFO 98 is coupled to bus 82.
  • a controller 100 controls both the analog digital converter 96 as well as the FIFO 98.
  • the controller 100 may also reset the LG/PH 94 in order to reset the circuit to take counts from a new sample.
  • Both the controller 100 and FIFO 98 are connected to bus 82 which transmits data and instructions between a CPU 84, PROM 86, RAM 88, and EEPROM 90.
  • the operation of activity meter 10 is as follows.
  • the user controls and makes selections as well as views results with keypad 18 and display 20.
  • the user is presented with a main measurement screen 110 as shown in FIG. 5 is displayed to the user.
  • the user may select the nuclide type to be tested by means of the keypad 18.
  • the characteristics data of up to twenty nuclides may be stored in EEPROM 90. Of course, a larger capacity EEPROM will allow more nuclide data to be stored.
  • the user may also enter nuclide data for nuclides which are not stored in EEPROM 90.
  • the selected nuclide is displayed in field 112. Users are also prompted to enter the type of container used for the holding the sample to be tested. The type of container affects the radiation counts for samples as will be explained below.
  • the sample container is displayed in a sample container field 114.
  • the user may choose between three containers, an ampoule container, a syringe, or a vial. Finally, a user may choose a measurement interval such as 2, 6, 20, 60, 180, 600 or 1800 seconds. The measurement interval is displayed in a count field 116.
  • the sample measurement is commenced via a command from the user on the keypad 18. While the measurement is taken, a counter screen 120 as shown in FIG. 6 is displayed. The counter screen 120 is displayed on display 20 during the measurement interval. Counter screen 120 is updated every second while the sample is measured. The counter screen 120 displays the nuclide type in a nuclide field 122, as well as the container type in a container field 124. The remaining time in the measurement interval is shown in a .count remainder time field 126. The count is tallied within specific energy ranges according to the vertical axis of keV graph 128 and the current counting rate is recorded as bars on keV graph 128 and as numerical values in a count field 130. The user may select whether the counts are in Curies (Ci) or Bequerels (Bq) . The user may also choose to terminate the measurement before the measurement interval is over.
  • Curies Curies
  • Bq Bequerels
  • FIG. 7 shows a result screen 140 which displays the radioactivity of a sample in a result field 142.
  • the results may be in either Ci or Bqs.
  • the sample type is shown in a nuclide field 144 and the container type is shown in a container field 146.
  • the impurity level in the sample may also be displayed in an impurity field 148.
  • FIG. 8 shows an optional bar graph screen 150 which may be selected by the user. Bar graph screen 150 shows the nuclide type in a nuclide field 152 as well as the container type in a container field 154.
  • the counts between the channels on an energy scale 156 are used to graph the activity of the current nuclide.
  • the channels which are used to calculate this activity are shown as solid bars, while other channels indicating impurities are hatched.
  • the overall count at each level is shown in user selected units of either Cis or Bqs. Both results screens shown in FIGS. 7 and 8 may be printed out or transmitted to a more permanent storage device.
  • the present invention takes advantage of non- invasively measuring the activities of pure beta emitter samples by measuring the production of continuous low energy photon emission produced by a decelerating electron or positron.
  • This continuous low energy photon emission is termed Bremsstrahlung. Since the energy and the intensity of Bremsstrahlung is very low and the intensity is dependent upon the sample configuration and the material of the container and the surroundings of the sample, the present invention may measure and account for the emissions based on impurities in the sample.
  • the Bremsstrahlung of the sample is measured by scintillation detector 42.
  • the scintillation phosphor contained in detector 42 emits a light intensity proportional to the energy lost by ionizing radiation in the phosphor.
  • the ionizing radiation in the phosphor is triggered by Bremsstrahlung of the sample.
  • the photodetector 58 converts the light intensity of the phosphor into an electrical signal.
  • the resultant electrical signal is proportioned to the energy deposited in the phosphor scintillator from the sample.
  • the present invention is capable of discriminating between levels of photon energy.
  • the levels of photon energy from Sr-89 will be in the 30-400 keV range while impurities such as Sr-85 will be in the 400 and higher keV range.
  • FIG. 9 shows the calculated external Bremsstrahlung spectrum of a typical beta ray emitter, Phosphorous-32.
  • the spectrum is expressed as the number of photons per unit energy interval as a function of the photon energy.
  • the characteristic X-rays specific for the stopping media will also be produced.
  • the intensity of the photons * is proportional to the atomic number (Z) of the stopping media.
  • Z atomic number
  • more photons are generated by a glass vial than a plastic vial of the same geometry.
  • Sensitivity to low energy photons may be reduced by inserting a filter around detector 42.
  • the filter is preferably mu-metal having a thickness of .5 mm.
  • the mu-metal filter also suppresses the effects of external magnetic fields on the stability of photo-detector 66.
  • the energy of Bremsstrahlung is rather low.
  • the sensitivity of the scintillator to high energy photons is roughly proportional to the volume of the scintillator whereas the sensitivity of the scintillator to low entry photons is roughly proportional to the area.
  • the present invention is able to maintain optimized detector sensitivity to Bremsstrahlung while reducing the sensitivity to high entry gamma rays from impurities in the sample.
  • the Nal(TI) detector thickness is approximately .3cm to 1cm.
  • FIG. 10A and 10B provide a comparison of the relative sensitivities against photon energy of a commonly used 2"x2" Na ⁇ (Tl) detector and scintillation detector 42 according to the present invention.
  • FIG. 10A shows an 18 month old Sr-89 sample which was measured by a 2"x2" Na ⁇ (Tl) detector. As may be seen by FIG. 10A, approximately 35% of the counts below 400keV are from 1% Sr-85 contamination in the sample.
  • FIG. 10B shows an 18 month old Sr- 89 sample which was measured by the scintillation detector 42 with optimized geometry according to the present invention. As may be seen by FIG. 10B, approximately 10% of the counts below 400keV are from 1% Sr-85 contamination in the sample.
  • the ratio of the volume to the frontal surface area is reduced with optimized geometry and thus the sensitivity to the higher energy gamma rays from the Sr-85 is reduced.
  • the length of the sample and relative height of the sample in the sample container will also have an influence on the counting rate measurements due to change in the average distance between the sample and the detector and the change in the angle of the detector looking from the sample. Change in the geometry also has an effect on the relative contribution of scattering and attenuation by the source and by the surrounding media.

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  • Physics & Mathematics (AREA)
  • Health & Medical Sciences (AREA)
  • Life Sciences & Earth Sciences (AREA)
  • General Physics & Mathematics (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Molecular Biology (AREA)
  • Spectroscopy & Molecular Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Analytical Chemistry (AREA)
  • Measurement Of Radiation (AREA)

Abstract

Dispositif de mesure destiné à la détection non-invasive des radionuclides et permettant de quantifier de manière précise des substances à radionuclides. Ce dispositif comprend une unité de commande accouplée à un porte-échantillon. L'échantillon est maintenu dans un récipient blindé et un détecteur à scintillation de phosphore est utilisé pour détecter l'émission photonique de faible énergie, ou rayonnement de freinage. Les émissions photoniques de faible énergie sont mesurées au moyen de relevés de l'énergie effectués par un compteur à six canaux qui compte les émissions à différents niveaux d'énergie. La mesure des énergies de scintillation provenant du phosphore et le comptage des émissions photoniques à différents niveaux d'énergie à partir de l'échantillon à radionuclides permettent d'estimer le niveau d'impuretés et de corriger la détermination de l'activité de l'échantillon.
PCT/US1996/006322 1995-05-05 1996-05-06 Dispositif de mesure destine a la detection non-invasive des radionuclides WO1996035131A1 (fr)

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Cited By (2)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
EP1416297A1 (fr) * 2002-10-31 2004-05-06 AEA Technology QSA GmbH Standard appareil d'etallonage de doses pour emetteurs beta et methode utilisant l'appareil
US20130124103A1 (en) * 2010-04-09 2013-05-16 Medrad, Inc. Radiopharmaceutical Concentration Measurement System and Method

Citations (4)

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Publication number Priority date Publication date Assignee Title
GB1131036A (en) * 1965-03-30 1968-10-16 Commissariat Energie Atomique Device for determining liquid volumes by measurement of the dilution of a radio-active solution
US3597596A (en) * 1969-02-07 1971-08-03 Atomic Energy Commission Analysis of large quantities of materials
US4682035A (en) * 1985-04-23 1987-07-21 Bioscan, Inc. Solid state counting system for high energy beta and gamma decay isotopes
EP0409776A2 (fr) * 1989-07-21 1991-01-23 Icn Biomedicals Inc. Appareil pour manipuler des échantillons et collecter des données

Patent Citations (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB1131036A (en) * 1965-03-30 1968-10-16 Commissariat Energie Atomique Device for determining liquid volumes by measurement of the dilution of a radio-active solution
US3597596A (en) * 1969-02-07 1971-08-03 Atomic Energy Commission Analysis of large quantities of materials
US4682035A (en) * 1985-04-23 1987-07-21 Bioscan, Inc. Solid state counting system for high energy beta and gamma decay isotopes
EP0409776A2 (fr) * 1989-07-21 1991-01-23 Icn Biomedicals Inc. Appareil pour manipuler des échantillons et collecter des données

Cited By (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
EP1416297A1 (fr) * 2002-10-31 2004-05-06 AEA Technology QSA GmbH Standard appareil d'etallonage de doses pour emetteurs beta et methode utilisant l'appareil
US20130124103A1 (en) * 2010-04-09 2013-05-16 Medrad, Inc. Radiopharmaceutical Concentration Measurement System and Method
EP2563409A4 (fr) * 2010-04-09 2016-10-26 Bayer Healthcare Llc Système et procédé de mesure de concentration radiopharmaceutique
US9715020B2 (en) 2010-04-09 2017-07-25 Bayer Healthcare Llc Radiopharmaceutical concentration measurement system and method
KR101803113B1 (ko) * 2010-04-09 2017-12-28 바이엘 헬쓰케어 엘엘씨 방사성의약품 농도 측정 시스템 및 방법

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