RADIOPHARMACEUTICALS METHODS OF ANALYSIS. September Revised Draft for adoption

Transcription

1 September 2008 RESTRICTED RADIOPHARMACEUTICALS METHODS OF ANALYSIS September 2008 Revised Draft for adoption World Health Organization 2008 All rights reserved. This draft is intended for a restricted audience only, i.e. the individuals and organizations having received this draft. The draft may not be reviewed, abstracted, quoted, reproduced, transmitted, distributed, translated or adapted, in part or in whole, in any form or by any means outside these individuals and organizations (including the organizations concerned staff and member organizations) without the permission of WHO. The draft should not be displayed on any web site. Please send any request for permission to: Dr Sabine Kopp, Quality Assurance Programme, Medicines Quality Assurance Programme, Quality & Safety: Medicines (QSM), Department of Essential Medicines and Pharmaceutical Policies (EMP), World Health Organization, CH-1211 Geneva 27, Switzerland. Fax: (41-22) ; s: kopps@who.int. The designations employed and the presentation of the material in this draft do not imply the expression of any opinion whatsoever on the part of the World Health Organization concerning the legal status of any country, territory, city or area or of its authorities, or concerning the delimitation of its frontiers or boundaries. Dotted lines on maps represent approximate border lines for which there may not yet be full agreement. The mention of specific companies or of certain manufacturers products does not imply that they are endorsed or recommended by the World Health Organization in preference to others of a similar nature that are not mentioned. Errors and omissions excepted, the names of proprietary products are distinguished by initial capital letters. All reasonable precautions have been taken by the World Health Organization to verify the information contained in this draft. However, the printed material is being distributed without warranty of any kind, either expressed or implied. The responsibility for the interpretation and use of the material lies with the reader. In no event shall the World Health Organization be liable for damages arising from its use. This draft does not necessarily represent the decisions or the stated policy of the World Health Organization.

2 page 2 Contents R1. Physical and Physicochemical Methods R1.1 Detection and measurement of radioactivity: general introduction R1.2 Radiation spectrometry R1.2.1 Crystal scintillation spectrometry R1.2.2 Semiconductor detector spectrometry R1.2.3 Liquid scintillation counting R1.3 Determination of radionuclidic purity R1.4 Electrophoresis R2. Chemical methods R2.1 Tin analysis R2.1.1 Tin estimation by gas chromatography or high performance liquid chromatography R2.1.2 Tin estimation by polarography R2.1.3 Tin estimation by potentiometric titration with potassium iodate (for kits) R2.1.4 Tin estimation by UV absorption R3. Biological methods R3.1 Biological distribution new page Radiopharmaceuticals Methods of Analysis R1. Physical and Physicochemical Methods R1.1 Detection and measurement of radioactivity: general introduction When measuring the radioactivity of radiopharmaceutical preparations it is necessary to use standardized solutions of the appropriate radionuclide. Standardized solutions of radionuclides are available from laboratories recognized by the relevant national or regional authority (see General Notice on Reference substances). When measuring the radioactivity of radiopharmaceutical preparations containing 99m Tc, a good approximation may be obtained using an ionization chamber and employing a standardized solution of 57 Co provided that correction for the differences in the radiations emitted are made.

3 page 3 Radioactive decay may involve the emission of charged particles, the process of electron capture, or the process of isomeric transition. The charged particles emitted from the nucleus may be alpha particles (helium nuclei of mass number 4) or beta particles (electrons of negative or positive charge, beta or beta + respectively, the latter known as positrons). The emission of charged particles from the nucleus may be accompanied by gamma rays, which are of the same physical nature as X-rays. Gamma rays are also emitted in the process of isomeric transition (IT). X-rays, which may be accompanied by gamma rays, are emitted in the process of electron capture (EC). Positrons are annihilated on contact with matter. Each positron annihilated is accompanied by the emission of 2 gamma rays, at 180 degrees to one another, each with energy of MeV. The methods employed for the detection and measurement of radioactivity are dependent upon the nature and energy of the radiation emitted. Radioactivity may be detected and/or measured by a number of different instruments based upon the action of radiation in causing the ionization of gases and solids, or the scintillation in certain solids and liquids, or by the effect of radiation on a photographic emulsion. In general, a counting assembly consists of a sensing unit and an electronic scaling device. The sensing unit may be a Geiger-Müller tube, a proportional counter, a scintillation detector in which a photomultiplier tube is employed in conjunction with a scintillator, or a solid-state semi-conductor. Geiger-Müller counters and proportional counters are generally used for the measurement of the beta emitters. Scintillation counters employing liquid or solid phosphors may be used for the measurement of alpha, beta, and gamma emitters. Solid-state devices may also be used for alpha, beta, and gamma measurements. The electronic circuitry associated with a detector system usually consists of a high-voltage supply, an amplifier, a pulse-height selector, and a scaler, a rate meter, or other readout device. When the electronic scaling device or the scaler in a counting assembly is replaced by an electronic integrating device, the resultant assembly is a rate meter. Rate meters are used for the purpose of monitoring and surveying radioactivity and are somewhat less precise as measuring instruments than the counters. Ionization chambers are often used for measuring gamma-ray emitters and, similar type of thin-walled instruments for measuring X-rays. Dose calibrators are ionization chambers used for measuring the amount of radioactivity in a vial or the dose to a patient in a syringe. Radiation from a radioactive source is emitted in all directions that is, isotropically. Procedures for the standardization and measurement of such sources by means of a count of the emissions in all directions are known as 4π-counting; those based on a count of the emissions in a solid angle of 2π steradians are known as 2π-counting; and those based on a fraction of the emissions defined by the solid angle subtended from the detector to the source are known as counting in a fixed geometry. It is customary to assay the radioactivity of a preparation by comparison with a standardized preparation using identical geometry conditions. The validity of such an assay is critically dependent upon the reproducibility of the spatial relationships of the source to the detector and its surroundings and upon the accuracy of the standardized preparation. In the primary

4 page 4 standardization of radionuclides coincidence techniques are employed in preference to simple 4π-counting whenever the decay scheme of the radionuclide permits. One of the most commonly employed coincidence techniques is 4π-beta/gamma coincidence counting, which is used for nuclides in which some or all of the disintegrations are followed by prompt photon emission. An additional adjacent detector, sensitive only to photons, is used to measure the efficiency in the 4π-counter of those disintegrations with which the photons are coincident. 4π-Gamma/gamma coincidence counting techniques are often employed for the standardization of pure gamma emitters. The construction and performance of instruments and accessory apparatus could vary to a great extent. The preparation of samples must therefore, be modified to obtain satisfactory results with a particular instrument. The operator must carefully follow the manufacturer's instructions for obtaining optimum instrument performance. The results must be substantiated by careful examination of known samples. Proper instrument functioning and reliability must be monitored on a day-to-day basis through the use of secondary reference preparations. Radioactivity occurring in materials of construction, or caused by cosmic rays, and to spontaneous discharges in the atmosphere contributes to what is known as the background activity. All sample radioactivity measurements must be corrected by subtracting the respective background activity. When counting of samples at high activity levels, corrections must be made also for loss of counts due to inability of the equipment to resolve pulses arriving in close succession. Such coincidence-loss corrections must be made prior to the background correction. The corrected count rate, R, is given by the formula: Where r is the observed count rate, and τ is the resolving time. A radioactivity count is a statistical value, i.e., it is a measure of nuclear decay probabilities, and is not exactly constant over any given time interval. The magnitude of the standard deviation is approximately equal to the square root of the number of counts. In general, at least counts are necessary to obtain a standard deviation of 1 %. Absorption Ionizing radiation is absorbed in the material surrounding the source of the radiation. Such absorption occurs in air, in the sample itself (self-absorption), in sample coverings, in the window of the detection device, and in any special absorbers placed between the sample and the detector. Since alpha particles have a short range of penetration in matter, beta particles have a somewhat greater range, and gamma rays are deeply penetrating, identification of the type and energy of radiation emitted from a particular radionuclide may be determined by the use of absorbers of varying thickness. In practice, this method

5 page 5 is seldom used, and that too mainly in connexion with beta emitters. Therefore, variations in counting rate due to (small) differences in thickness and density of sample containers could give rise to major problem with beta emitters and with X-ray emitters, such as iodine-125. Plastic containers, in which variations of density and thickness are minimal, are therefore often employed in such cases. Plastic tubes with defined density and thickness are therefore employed frequently. The absorption coefficient (µ), which is the reciprocal of the thickness expressed in mg/cm 2, or the half-thickness (the thickness of absorber required to reduce the radioactivity by a factor of two), is commonly determined to characterize the beta radiation emitted by a radionuclide. This equation is valid only for monoenergetic radiation. R1.2 Radiation spectrometry R1.2.1 Crystal scintillation spectrometry When the energy of beta or gamma radiation is dissipated within some materials known as scintillators, light is produced in an amount proportional to the energy dissipated. This quantity of light may be measured by suitable means, and is proportional to the energy absorbed in the scintillator. The light emitted under the impact of a gamma photon or a beta particle is converted into an electric output pulse by a photomultiplier. Scanning of the output pulses with a suitable pulse-height analyser results in an energy spectrum of the source. The scintillators most commonly used for gamma spectrometry are single crystals of thallium-activated sodium iodide. Gamma-ray scintillation spectra show one or more sharp, characteristic photoelectric peaks, corresponding to the energies of the gamma radiation of the source. They are thus useful for identification purposes and also for the detection of gamma-emitting impurities in a preparation. These peaks are accompanied by other peaks due to secondary effects of radiation on the scintillator and its surroundings, such as backscatter, positron annihilation, coincidence summing, and fluorescent X-rays. In addition, broad bands known as the Compton continua arise from the scattering of the gamma photons in the scintillator and in surrounding materials. Calibration of the instrument is performed with the use of reference preparations of radionuclides whose energy spectra have been characterized. The shape of the spectrum produced will vary with the instrument used, owing to such factors as differences in the shape and size of the crystal, in the shielding materials used, the distance between the source and the detector, and in the types of discriminator employed in the pulse-height analysers. When using the spectrum for identification of radionuclides it is therefore necessary to compare the spectrum with that of a reference preparation of the radionuclide obtained in the same instrument under identical conditions. Certain radionuclides, for example, iodine-125, emit characteristic X-rays of well-defined energies that will produce photoelectric peaks in a suitable gamma spectrometer. Beta radiation also interacts with the scintillators, but the spectra are continuous and diffuse

6 page 6 and generally of no use for identification of the radionuclide or for the detection of betaemitting impurities in a radiopharmaceutical preparation. R1.2.2 Semiconductor detector spectrometry Gamma-ray spectra may be obtained using solid-state detectors. The peaks obtained do not suffer to the same extent the broadening shown in crystal scintillation spectrometry, and the resolution of gamma photons of similar energies is very much improved. However, the efficiencies of such detectors are much lower. The energy required to create an electron-hole pair or to promote an electron from the valence band to the conduction band in a semi-conductor is far less than the energy required to produce a photon in a scintillation crystal. In gamma-ray spectrometry a high purity germanium (HPGe) detector can provide an energy resolution of % for the 1.33 MeV photon of cobalt-60. R1.2 3 Liquid scintillation counting For beta-emitters like 35 S, 14 C and 3 H, where self-absorption of the low-energy beta particles is significant, the preferred counting method is by liquid scintillation, which can occasionally be employed also for emitters of X-rays, alpha-particles, and gamma-rays. If the sample to be counted is dissolved in, or mixed with, a solution of an appropriate scintillator material, the decay energy from the sample is converted into light photons. These are sensed by a photomultiplier, which converts them into an electric pulse, whose intensity is proportional to the energy of the initial radiation. Thus, simultaneous counting of several radionuclides differing in the energy of emitted radiation can be effected with suitable discriminators (pulse-height analysers), provided the energy separation is adequate. Detection efficiencies approaching 95 % for 14 C and 60 % for 3 H are reached because self-absorption is minimized. The scintillator (to check the chemical) solute usually consists of a polycyclic aromatic compound, such as p-terphenyl or 2,5-diphenyloxazole (primary solute), together with a secondary solute, such as 1,4-di[2-(4-methyl-5-phenyloxazole)]benzene (Dimethyl- POPOP), that shifts the wavelength of the light emitted to match the highest sensitivity of the photomultiplier tube. Water-immiscible solvents, such as toluene, or water-miscible solvents, such as dioxan, can be used. To facilitate the counting of aqueous solutions, special solvents have been developed. Alternatively, samples may be counted as suspensions in scintillator gels. As a means of attaining compatibility and miscibility with aqueous specimens to be assayed, many additives, such as surfactants and solubilizing agents, are also incorporated into the scintillator. For accurate determination of sample radioactivity, care must be taken to prepare a sample that is truly homogeneous. The presence of impurities and colour in the solution causes a decrease in the number and energy of photons reaching the photomultiplier tube; such a decrease is known as quenching. Accurate radioactivity measurement requires correcting for count-rate loss due to quenching. Solutions containing organic scintillators are prone to photo-excitation and samples may need to be prepared in subdued light and kept in darkness before and during counting process.

7 page 7 R1.3 Determination of radionuclidic purity For gamma emitters the most useful method of examination for radionuclide purity is gamma spectrometry. It does have limitations, however, because: beta-emitting impurities are, in general, not detected; When sodium iodide detectors are employed, the photoelectric peaks due to impurities may be obscured by those due to the major radionuclide, or, in other words, the degree of resolution of the instrument could be insufficient. This problem could be solved by the use of high resolution solid state semiconductor detectors, such as high purity germanium (HPGe) detector. Unless the instrument has been calibrated with a standard source of known radionuclide purity under identical conditions of geometry, it is difficult to determine whether additional peaks are due to impurities or whether they result from such secondary effects as backscatter, coincidence summation, or fluorescent X-rays. The range of gamma spectrometry may be extended in two ways first, by observing changes in the spectrum of a preparation with time (this is especially useful in detecting the presence of long-lived impurities in a preparation of a short-lived radionuclide); secondly, by the use of chemical separations, whereby the major radionuclide may be removed by chemical means and the residue examined for impurities, or whereby specific impurities may be separated chemically and then quantified. It is evident that chemical means will not separate an impurity that is isotopic with the major radionuclide. Radionuclide impurities are directly related to the production process of a radionuclide. Based on technical limitations and safety requirements limits have been set for radionuclidic impurities in radiopharmaceutical preparations, expressed as a percentage of the total radioactivity. For identification of gamma emitters the method of choice is gamma spectrometry. In order to interpret the energy spectrum of radionuclides it is necessary that the energy range be calibrated with appropriate reference preparations. Gamma spectrometry may be performed using high resolution germanium detectors. Beta emitting impurities are not detected by gamma spectrometry. Long lived impurities in a preparation of a short-lived radionuclide may be determined after the decay of the short-lived radionuclide. Chemical separation of impurities is an effective method both during the production process and as an analytical procedure. The exact measurement of trace amounts of betaand alpha-emitting radionuclides in preparations of generally applied gamma radionuclides requires special techniques. Chemical separation of the radioactive impurities is used prior to the measurement of non-penetrating radiation. R1.4 Electrophoresis Use the method as described under 1.15 Electrophoresis, but using counting devices and detectors suitable for radiopharmaceuticals. These methods are particularly suitable for

8 page 8 charged radiopharmaceuticals (anionic, e.g. technetium ( 99m Tc) mebrofenin complex, radioiodinated o-hippuric acid, technetium ( 99m Tc) mertiadale injection or cationic, e.g. technetium ( 99m Tc) sestamibi complex injection. R2. Chemical methods R2.1 Tin analysis Tin is used for many technetium based radiopharmaceuticals and since this is the main radiopharmaceutical that is most widely used clinically the assessment of tin is essential. For an optimal radiopharmaceutical formulation milligram amounts are used and for some microgram amounts are used. The actual levels can affect the final radiochemical purity and alter the pharmacokinetics of the radiopharmaceutical. Well-established methods are identified and used as the standard methods of analysis for tin estimation. Analytical methods approved by the relevant regional or national authority for application to environmental samples may be suitable. Additionally, analytical methods are included that modify previously used methods to obtain lower detection limits and/or to improve accuracy and precision. The specific requirements are included in the relevant individual monographs. R2.1.1 Tin estimation by gas chromatography or high performance liquid chromatography Tin is usually determined as the total metal, but it may also be measured as specific organo-tin compounds. Flame atomic absorption analysis is the most widely used and straightforward method for determining tin; furnace atomic absorption analysis is used for very low analyte levels and inductively coupled plasma atomic emission analysis is used for multi-analyte analyses that include tin. The preferred separation technique for organo-tin compounds is gas chromatography (GC) due to its high resolution and detector versatility. High performance liquid chromatography (HPLC) has also been used in the analysis of organo-tin compounds. The advantage of HPLC over GC is that no derivatization step is needed after extraction. For determination of tin in biological samples, the sample is digested in an oxidizing acid mixture followed by atomic spectrometric determination. Determination of organo-tin compounds in biological materials will require extraction, derivatization, separation, and detection, as described. Whole blood samples are typically analysed by spectrophotometry and photometry. R2.1.2 Tin estimation by polarography Tin can be effectively analysed by polarography, which is also called polarographic analysis, or voltammetry method of analysing solutions of reducible or oxidizable substances. Polarography technique involves electric potential (or voltage) varied in a regular manner between two sets of electrodes (indicator and reference) while the current is monitored. The shape of a polarogram depends on the method of analysis selected, the type of indicator electrode used, and the potential ramp that is applied. The method is

9 page 9 useful in detecting several substances simultaneously and is applicable to relatively small concentrations, e.g up to about 0.01 mole per litre, or approximately 1 to 1000 parts per million. R2.1.3 Tin estimation by potentiometric titration with potassium iodate (for kits) Potentiometric titration is based on the principle of measuring the change in redox potential when tin solution is titrated against potassium iodate solution. The redox potential is measured with redox-electrode couple. This method is ideal for estimating stannous (tin II) contents in radiopharmaceutical vials sealed in nitrogen or inert gases. Tin estimation by potentiometric titration is not possible in vials containing antioxidants such as ascorbic acid or gentisic acid. Since antioxidants are commonly found in radiopharmaceutical preparations this method is not suitable for such formulations. Reagents Prepare the following two reagents as described: Potassium iodate stock solution VS Potassium iodate R, dissolved in water R, purged with nitrogen R for 5 minutes before use, to contain g in 1000 ml (1.667x10-3 mol/l). Method of standardization. Ascertain the exact concentration of the solution following the method described under potassium iodate (0.05 mol/l) VS. Prepare a fresh solution every three months. Potassium iodate working solution VS Dilute 10ml of Potassium iodate stock solution VS to 50ml with water R, purged with nitrogen R for 5 minutes before use (0.334 x10-3 mol/l). Method of standardization. Ascertain the exact concentration of the solution following the method described under potassium iodate (0.05 mol/l) VS. Prepare a fresh solution each day. Titration method The apparatus consists of a suitable titration cell assembly 2 with a redox-electrode operating in milli-volt mode. Pass a gentle stream of nitrogen R i through the assembly to mix the solution and provide an inert atmosphere. Reconstitute the stannous tincontaining test preparation with 4.0ml of sodium chloride(9g/l) TS and dispense 1.0ml of the resulting solution into the titration cell. Add 2.0ml of hydrochloric acid (1mol/l) VS ) and titrate immediately with either potassium iodate stock solution VS or potassium iodate working solution VS, as appropriate using a microburette until the endpoint (a marked, persistent jump in redox-electrode potential) is achieved. Record the volume of titrant in ml. 2 Metrohm is suitable (NfS: as footnote to be deleted before publication)

10 page 10 Titrate radiopharmaceutical kits containing high stannous (tin II) content (e.g. PYP and Phytate colloid kits) with potassium iodate stock solution VS which contains 594 microgram Sn(II) per ml solution. The volume of titrant required to achieve the projected end-point is indicated in the following table. Type of kit Theoretical Sn(II)content /ml ml of titrant Pyrophosphate (PYP) PHYTATE Titrate radiopharmaceutical kits containing low stannous (tin II) content (e.g. DTPA, DISIDA kits) with potassium iodate working solution VS, which contains 119 microgram Sn(II) per ml. The volume of titrant required to achieve the projected endpoint is indicated in the following table. Type of kit Theoretical Sn(II) content /ml ml of titrant Pentetate complex (DTPA) Di-iso propyl imino deacetic acid (DISIDA) Imidodiphosphonate (IDP) Stannous fluride SnF 2

11 page 11 Unless otherwise specified, the radiopharmaceutical kit being tested contains more than 85% of the theoretical content of tin (II). R2.1.4 Tin estimation by UV absorption Prepare the test and reference solutions as described in the monograph. To 1.0 ml of each solution add 0.05 ml of thioglycollic acid R, 0.1 ml of dithiol reagent R, 0.4 ml of a 20g/l solution of sodium laurilsulfate R and 3.0 ml of hydrochloric acid (0.2mol/l) VS. Mix each of the solutions thoroughly. Measure the absorbance (1.6) of of a 1-cm layer of each solution at 540 nm, against a solvent cell containing hydrochloric acid (0.2mol/l) VS. The absorbance of the test solution is not greater than that of the reference solution. New reagents [Note from Secretariat. Details to be included.] Thioglycollic acid R Dithiol reagent R R3. Biological Methods R3.1 Biodistribution A physiological distribution test is prescribed, if necessary, for certain radiopharmaceutical preparations. Specific requirements are set out in individual monographs. The distribution pattern of radioactivity observed in specified organs, tissues or other body compartments of an appropriate animal species (usually rats or mice) can be a reliable indication of the expected distribution in humans and thus of the suitability of the intended purpose. The individual monograph prescribes the details concerning the performance of the test and the physiological distribution requirements, which must be met for the radiopharmaceutical preparation. A physiological distribution conforming to the requirements will assure appropriate distribution of the radioactive compounds to the intended biological target in humans and limits its distribution to nontarget areas. Selection of animals The animals used in this test are healthy animals, drawn from uniform stocks that have not previously been treated with any material that will interfere with the test. If relevant, the species, sex, strain and weight and/or age of the animals are specified in the monograph. Unless otherwise stated, mice weigh not less than 20g and not more than 30g; rats weigh not less than 150g and not more than 250g; and guinea pigs (especially for cardiac radiopharmaceuticals) weigh not less than 250g.

12 page 12 Method Where applicable, reconstitute the test preparation according to the manufacturer s instructions. In most cases, dilution immediately before injection may be necessary to ensure optimal radioactivity count characteristics. Unless otherwise stated, inject the specified dose (x) of the radiopharmaceutical preparation into the caudal vein of three animals previously weighed and, where necessary, warmed to room temperature under an infrared lamp. Swab the injection site with cotton wool and retain the cotton wool and the residual dose in the syringe after injecting for counting (y) and (z) respectively. Actual injected dose (a) = x-(y+z). Immediately after injection, place each animal in a separate cage that is designed to allow collection of excreta and to prevent contamination of the body surface of the animal. After the time period specified in the monograph, kill the animals. Collect a sample of blood by cardiac puncture and record the weight of the sample. Dissect out the required organs and tissues, e.g. gall bladder, liver, stomach, intestines, bones and kidneys and place in separate labelled counting tubes. Remove the tail above the injection site and place in a labelled counting tube. Prepare three dose standards (0.2ml) in counting tubes. Count remaining organs and standards in an automatic gamma-well counter or other suitable device. Determine the percentage of injected radioactivity in all organs according to the following formula: 100 x (A/a) where: A = radioactivity in organ; a = actual injected radioactivity. The percentage of radioactivity in blood is determined according to the formula: [100x(C/Ws) x 0.07 x (Wr)] / a where C = Radioactivity in specimen of blood; Ws = weight in grams of blood specimen and Wr = weight in grams of animal. (Normally blood is approx. 7% of total body weight.) Calculate the physiological distribution and express as the percentage of the injected dose/gram wet weight of tissue. Tissues are counted in optimally calibrated gamma counters. Specification The preparation meets the requirements of the test, if the distribution of radioactivity in at least two of the three animals complies with the criteria specified in the monograph. Disregard the results from any animal showing evidence of extravasation of the injection (observed at the time of injection or revealed by subsequent assay of tissue radioactivity). ***