CH661215A5 - METHOD FOR PRODUCING A LIQUID CONTAINING A RADIOISOTOP. - Google Patents

Description

The present invention relates to a method for producing a liquid for radiopharmaceutical applications, which liquid contains a radioisotope, and to an apparatus and a reservoir for carrying out the method. The method is used to produce a liquid for radiopharmaceutical applications, which liquid contains a radioisotope, by eluting a radioactive daughter isotope from a mother isotope, which is adsorbed on an adsorbent, with the aid of a physiological solution. The method is defined in claim 1. A device for its implementation is defined in claim 9, the reservoir in claims 14 and 15.

Radioisotopes with a half-life of up to a few days are used for diagnostic purposes in medicine. In order to minimize tissue damage from radiation, it is advisable to use radioisotopes that only emit gamma rays. The radioisotope "mTc is a pure gamma radiator and has a relatively short half-life. For this reason, this isotope is excellently suitable s

10th

15

20th

25th

30th

35

40

45

50

55

60

65

3rd

661 2i5

for use as a diagnostic, but it can also be used for radioactive labeling of other substances, e.g. Proteins. The 99mTc isotope is generated by radioactive decay of the mother isotope "Mo. It is known, for example from US Pat. No. 3,970,583, to adsorb the mother isotope in the form of a molybdate onto a suitable adsorbent and then elute the daughter isotope "mTc using physiological saline. A device suitable for producing an "mTc-containing liquid in this way is an isotope generator, as is also described in the above-mentioned US patent.

As a result of the rapid development of radiodiagnostics in the past 10 years, a need has arisen for a liquid for radiopharmaceutical use which contains a radioisotope and which has a higher concentration of radioactive material and a greater chemical purity than the radiodiagnostics used hitherto. The previous "technetium eluate was generated in an isotope generator from natural or enriched molybdenum, which was irradiated in a nuclear reactor. The radioactive isotope" Mo is present in this product in very small concentrations; the main part of the product consists of non-radioactive molybdenum and serves as a carrier for "Mo. The dimensions of the column which contains the adsorbent for the mother isotope are limited because a column which is too large cannot be eluted effectively. This relates in particular to withdrawing small volumes of elution from the column which may be necessary for certain purposes in which a higher concentration of isotope is required Since restrictions are imposed on the dimensions of the column and the adsorption capacity of the adsorbent, only a relatively small amount can be used on the mother isotope in the generator, so that the required high concentration of radioactivity in the eluate is not available with the previously known isotope generators.

Meanwhile radioactive isotopes e.g. radioactive molybdenum and cerium, produced in a different way, namely by a cleavage reaction. For example, "Molybdenum is produced by splitting 235U; 235U is irradiated with neutrons in a nuclear reactor, whereupon the other fission products can be removed from the" Mo by a chemical separation process. A radioisotope generated by cleavage is purified to an acceptable level of radionuclide purity, but it still contains traces of contaminants such as 115Cd, 136Cs, 140La, 156Eu, 89Sr, 90Sr, 95Zr, 140Ba and actinides. In addition to the gamma rays emitted by most of these radioisotopes, these impurities also emit corpuscular rays, namely alpha or beta rays. These alpha or beta rays are very undesirable in pharmaceutical preparations because they can seriously attack the tissues; the strontium isotopes and the actinides are considered the most toxic isotopes.

The invention has for its object to provide a method for producing a liquid which contains a radioisotope suitable for radiopharmaceutical use. This object is achieved by the method defined in claim 1, in which an eluate obtained by elution of a mother isotope produced by cleavage, which is adsorbed on a suitable adsorbent, and which contains the daughter isotope, with the aid of a cation exchange material, preferably one Cation exchange resin is cleaned. Strongly acidic cation exchange resins, which are neutralized, and a are particularly suitable for this purpose

Particle size of e.g. B. 37 to 300 microns, preferably 74 to 150 microns. An example of a resin suitable for this purpose is «Dowex» or «Bio-Rad 50W-X8». These strongly acidic resins are preferably neutralized by treatment with an alkali metal base, e.g. NaOh, KOH or with NH4OH, and then washing with water. In this way the resins are converted into the Na +, K + or NH4 + form.

From Int. J. Appi. Rad. Isotopes 1978, Vol. 29, pages 91 to 96, it is known that the resin "Dowex 50W-X8" in Na + form can be used for the separation of 90Y from 90Sr. Under the reaction conditions described in this article, namely in the presence of a small amount of EDTA, the influence of pH on the adsorption of 90Sr was determined. The results show that 90Sr is adsorbed on «Dowex 50» resin at a pH of 1.5 to 5.5 but not at a pH of 7.0. The concentration of EDTA had no influence on the adsorption of 90Sr. These results lead to the assumption

that «Dowex 50» resin is not suitable for adsorbing 90Sr from a solution suitable for pharmaceutical use, namely an almost neutral physiological saline solution. Contrary to expectations, however, it was found that a cation exchange resin, in particular a strongly acidic cation exchange resin, such as "Dowex" or "Bio-Rad 50W-X8", which is converted into the Na +, K + or NH4 + form, is particularly suitable to purify "mTC generated from cleaved Mo" so that a solution containing 99mTc and suitable for radiopharmaceutical use is obtained in an exceptionally high chemical, radiochemical and radionuclide purity.

From the above-mentioned US patent it is known that aluminum oxide, which contains wholly or partly hydrated manganese dioxide, is an adsorbent for the mother isotope "Mo". This material is also excellently suitable as an adsorbent for the completely or nearly molybdenum carrier-free , "Mo. This is not a matter of course, because the latter concerns extremely small amounts of adsorbed molybdenum, which also contains undesirable impurities. The desired optimal elution yield is highly dependent on the type and amount of material to be eluted and the adsorbed material present, and it is well known that small differences in these matters can easily upset the subtle balance, which in turn results in either a poorer yield or lead to an undesirable elution pattern.

From the above it can be seen that the method according to the present invention is preferably carried out in a device as defined in claim 9. The device contains an eluent reservoir equipped with an eluent drain. The reservoir is surrounded by a generator housing. Such a system is often referred to as a “cow”. The device for carrying out the method described above comprises a reservoir which has a feed element for the eluent and a discharge element for the eluate, and in which the adsorbent for the mother isotope is present. A device called such a generator is known, e.g. from the above U.S. Patent.

However, the device according to the present invention also comprises a radioisotope generated by cleavage and a cation exchange material. Because the radioisotope produced by cleavage is completely or almost carrier-free, a small amount of adsorbent is completely sufficient for the mother isotope. As a result, the di-

5

10th

15

20th

25th

30th

35

40

45

50

55

60

65

661215

Dimensions of the device are greatly reduced, so that the device is easier to handle, both in use (in hospitals or clinical laboratories, the device must be replaced regularly) and after assembly by the manufacturer. It is of great advantage that the cation exchange material is also present in the device according to the invention. As a result of this fact, the eluate can be cleaned in the device itself, so that the liquid drawn off from the device, which contains the radioactive daughter isotope, has a higher chemical and radionuclide purity and is therefore suitable for radiopharmaceutical use. An additional cleaning of the eluas, i.e. after the eluate has left the device, is unnecessary.

Such additional purification is generally even impossible or at least undesirable because the daughter isotope obtained usually has a half-life too short to be able to survive such post-treatment and also because post-treatment in a hospital or clinical laboratory , where suitable aids for this purpose are missing, is out of the question for security reasons.

It is common to enclose the adsorbent for the mother isotope in the reservoir between two filters. In order to load the adsorbent with the mother radioactive isotope, a solution of this isotope is permitted on one side of the reservoir. Glass wool or glass beads are often used as filter material on this side. However, glass beads create channels in the adsorbent and thereby an ineffective loading and a non-uniform distribution of the mother isotope over the adsorbent. Glass wool, in turn, often prevents loading due to the excessive resistance and also tends to, like synthetic resins, e.g. Polyethylene to adsorb some mother isotope. However, this is: undesirable because after elution, the amount of mother isotope not adsorbed on the adsorbent contaminates the eluate.

As a special aspect of the present invention, it has now been found that the disadvantages mentioned above can be eliminated by producing the filter on the side of the reservoir on which the solution of the mother isotope is permitted, from sintered glass. It has been found that if such a filter is used, which of course can also be used in previously known isotope generators, an effective and homogeneous loading of the adsorbent can be achieved very easily without the mother isotope being adsorbed by the filter .

The device is preferably constructed such that both the cation exchange resin and the adsorbent for the mother isotope are present in the same reservoir. In this embodiment, in which the dimensions of the device can be reduced and an optimum purity of the radiopharmaceutical preparation can be achieved, the advantages mentioned above are emphasized even more, while the production costs can also be kept as low as possible.

In a further preferred embodiment, the reservoir, which contains both the cation exchange resin and the adsorbent for the mother isotope, is divided into two compartments which are separated from one another by a filter, the periphery of which reaches the inner wall of the reservoir. A compartment of the reservoir contains a feed member for the eluent, and the adsorbent for the mother isotope is between the feed member and the separating filter, the adsorbent being sandwiched between the above-mentioned sintered glass filter and the separating filter. The other compartment of the reservoir contains a discharge device for the eluate. The cation exchange material is located between the separating filter and the outflow element, and the space between the adsorbent particles and between the particles of the ion exchange resin is filled with a physiological solution. A suitable separating filter for this purpose consists of two filter disks, which completely or almost completely overlap, the disk which touches the adsorbent being made of glass fiber paper, e.g. a Millipor pre-filter AP 200, while the disc, which touches the ion exchange material, is made of porous polyethylene.

Finally, the invention relates to a reservoir for the above-mentioned device, which reservoir contains both the cation exchange material and the adsorbent for the mother isotope. It has been found that such a reservoir, which is loaded and sterilized, can be stored in the uncooled state for more than 3 months and can be incorporated into the device at any desired time during this period without any pretreatment. The reservoir can then be used to give a radioactive daughter isotope containing eluate in high yield. This is advantageous because the reservoirs can be manufactured in larger quantities and can be sent to the supplier of devices, who can use a reservoir for his devices at any desired time without any pretreatment; this means significant cost savings.

The invention is explained in more detail below with reference to the following specific examples.

The drawing is a cross section through an embodiment of the reservoir.

A practically cylindrical reservoir 4 made of a suitable inert material, e.g. Glass or a polymeric material, preferably borosilicate glass, is expanded at each end and provided with a flange 10.13. The openings at the two ends of the reservoir are by means of rubber pins 2.14, which have a flange 11.15 and a jacket part 12.16; the flange of the pin engages the flange of the reservoir, while the jacket part of the pin fits into the opening of the reservoir. The flange of the pin and the reservoir are connected to each other by a metal cover, e.g. an aluminum folding lid 1, 17. The reservoir contains a slurry of the adsorbent 5 in a solution of 0.9% NaCl in water. This adsorbent consists of AtC ^ particles, which are completely or partially covered with a layer of fully or partially hydrated manganese dioxide. In the reservoir, the adsorbent is enclosed between a filter made of sintered glass 3 of average porosity and a filter disc made of glass fiber paper 6, namely a Millipor pre-filter AP 200. The reservoir also contains a slurry of the resin "Bio-Rad 50W-X8" in the Na + -Form 8 in a solution of 0.9% NaCl in water. The resin was converted into the Na + form by treatment with NaOH and then washed with water. This resin is enclosed between a filter disk 7 made of porous polyethylene, which is in contact with the filter disk 6 and a filter disk 18, also made of porous polyethylene, which rests on a polycarbonate spacer ring 9.

example 1

Ten of the reservoirs described above were stored for 3 months and then used for the following experiment.

4th

5

10th

15

20th

25th

30th

35

40

45

50

55

60

65

5

661 215

Each reservoir was loaded with cleaved carrier-free molybdenum "Mo" in the form of sodium molybdate (pH 1.5 to 10) by perforating the pegs at the ends of the reservoir to provide inlet and outlet openings, and then a solution of the radioactive sodium molybdate flowed into the reservoir through the inlet opening (at A.) After washing and sterilizing in an autoclave at 121 ° C. for 30 minutes, the product thus obtained was placed in a device. The radioactivity was 1000 mCi.

In use, the eluent was delivered through the inlet at one end of the reservoir (at A) while the eluate was withdrawn through the outlet at the opposite end of the reservoir. The devices were eluted with sterile isotonic saline solutions (0.9% w / v NaCl in water) in amounts of 4.6 or 15 ml and the average elution yields obtained are shown in Table I below.

Table I

Properties of 4.6 ml and 15 ml "mTc-containing eluates

Eluate volume: 4.6 ml

Eluate volume: 15 ml

Eluate average

Eluate average

Elution yield

Elution yield

(%)

(%)

1

89.4

1

90.8

2nd

89.9

2nd

94.1

3rd

89.1

3rd

93.9

4th

88.6

4th

92.4

5 •

89.6

5

93.3

6

89.6

6

94.2

7

87.2

7

90.9

8th

89.3

8th

93.1

9

89.6

9

92.7

10th

88.5

10th

91.9

Continuation of table I

Analyzes

Analyzes pH: 6

radiochemical purity> 99%

Mn ++: <0.3 | xg / ml Al + ++: <0.5 | xg / ml Labeling effectiveness: -EHDP 97.8% -Sb2S3 97.5%

pH: 6

radiochemical purity> 99%

Mn ++: <0.3 ng / ml Al + ++: <0.5 | ig / ml labeling effectiveness: -EHDP 98.4% -Sb2S3 97.0%

radionuclide purity: -99Mo <1 nCi / mCi 99mTc -131I <1 nCi / mCi 99mTc -103Ru <1 nCi / mCi 99mTc 5 sterility: sterile radionuclide purity: -99Mo <2 nCi / mCi 99mTc -131mCi <i ni -103Ru <1 nCi / mCi 99mTc Sterility: sterile

Apyrogenicity: pyrogen-free Apyrogenicity: pyrogen-free

The elution yields shown in Table I above show the average percentages of the ten devices of theoretically possible 99mTc activity per elution, the devices being eluted every ten times on consecutive days. A high elution yield was achieved, namely an average of 15 89.06% + 2.15 and 92.17% + 1.64 of the theoretically possible 99mTc activity after elution with 4.6 ml and 15 ml of the eluent.

The table above also shows the average results of the eluate analysis. The radiochemical purity was determined using a paper chromatography method; the concentration of Mn + + and Al + + + was determined using a spectrophotometric method or a colorimetric method (as quinalizarin complex). The labeling effectiveness 25 of ethylene hydroxydiphosphonate (EHDP) and Sb2S3 proves that the 99mTc obtained is extremely suitable for the preparation of compounds labeled with 99mTC and that it can therefore be used for all desired applications. The radionuclide purity was determined using a gamma analyzer. At most traces of radioisotopic "Mo, l31I and I03Ru could be found; no other radionuclide contaminants were found in the eluate.

It can be seen from the results shown that the long storage of the filled reservoirs had no harmful effect on the elution yield and on the purity of the eluate.

In another experiment, the same device could be eluted 15 times with 20 ml of a physiological saline solution without changing the elution yield or without contaminating the resulting technetium eluate with cationic contaminants.

Example 2

45 Ten additional columns, substantially similar to those used in Example 1 but with less dead volume, were used to obtain 99mTc. The reduction in dead volume was achieved by omitting the spacer ring and using a cylindrical, flangeless reservoir in the manufacture of the column. The results are summarized in Tables II to V below.

Table II

Elution yields (% of possible 99mTc activity)

Device mCi "Mo loaded

Elution volume

1.E1.

2.E1.

3.E1.

4.E1.

5.El.

6.E1.

7.El.

8.El.

9.El.

10.E1.

1

768

15 ml

92.36

92.03

92.41

94.25

93.41

89.66

87.48

92.61

90.54

89.52

3rd

1161

4.6 ml

90.45

91.09

89.82

89.61

91.14

88.14

87.17

90.70

89.32

88.96

4th

1103

4.6 ml

90.25

91.15

89.65

89.09

91.30

88.45

85.69

94.15

88.62

88.59

5

1833

4.6 ml

92.02

91.03

90.19

89.76

88.99

80.74

86.24

91.93

89.06

87.12

6

1089

15 ml

92.50

93.27

91.82

92.10

91.47

89.41

88.05

94.18

90.46

89.40

7

1826

4.6 ml

90.44

89.91

90.21

90.15

89.91

91.69

86.36

93.22

89.21

87.49

9

1810

15 ml

92.24 '

93.50

91.95

92.41

91.93

92.34

88.81

93.68

91.42

88.86

10th

1815

4.6 ml

91.52

90.89

90.45

89.91

91.40

89.80

86.63

92.01

89.98

88.01

661 215

6

Table III Chemical purity of the eluate

Device - 1st eluate 10th eluate device No.

pH

(ig Mn / ml

Hg Al / ml pH

Hg Mn / ml

Hg Al / ml

1

5.8

<0.3

<1

5.8

0.55

<1

3rd

5.9

<0.3

<1

5.9

<0.30

<1

4th

5.8

<0.3

<1

5.8

<0.30

<1

5

5.9

<0.3

<1

6.0

<0.30

<1

6

5.8

<0.3

<1

5.9

0.35

<1

7

5.8

<0.3

<1

5.8

<0.30

<1

9

6.0

<0.3

<1

5.9

0.78

<1

10th

5.8

<0.3

<1

5.8

<0.30

<1

Radiochemical purity of the first eluate> 99%. Labeling properties of the first eluate Sb2S3 colloid. Labeled with 2 ml of the 4.6 ml eluate 96.0%

4 ml of the 4.6 ml eluate 93.9%

2 ml of the 15 ml eluate 100.0% 4 ml of the 15 ml eluate 99.8%.

Table IV

Preliminary experiment "Mo in the eluate detector: Nal / single channel analyzer

"Mo - breakthrough (nCi" Mo / mCi 99mTc)

Device l.EI. 2.E1. 3.E1. 4.E1. 5.E1. 6.E1. 7.E1. 8.E1. 9.E1. 10.E1. No.

1

1.78

1.47

1.72

0.91

1.47

0.78

0.31

1.24

0.90

1.88

3rd

1.80

1.70

2.00

1.52

1.23

1.41

2.01

2.01

0.34

1.76

4th

3.21

2.83

2.31

2.19

2.43

1.96

2.37

3.35

1.03

2.35

5

1.05

1.26

1.19

0.97

1.00

0.92

1.06

0.62

0.89

0j63

6

5.40

3.91

3.47

3.01

3.08

2.37

2.68

1.53

2.66

1.73

7

2.90

2.37

3.09

2.12

2.30

1.47

1.74

1.89

1.31

0.99

9

2.50

2.32

1.84

1.59

1.51

1.60

1.81

1.42

0.94

2.18

10th

1.82

1.42

1.61

1.33

1.15

1.25

1.66

1.35

1.03

1.53

definite try

Table V Radionuclide purity of the eluate detector: Ge (Li) / Nuclear Data System 4410 analyzer and ND 812 minicomputer

Gene. No. 1. Eluate 10. Eluate

131I nCi / mCi Tc 103Ru nCi / mCi Tc "Mo nCi / mCi Tc 131I nCi / mCi Tc

103 Ru nCi / mCi Tc "Mo nCi / mCi Tc

1

3rd

4th

5

6

7 9

10th

not detectable not detectable not detectable 0.005

0.004 sterility: sterile;

not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable

0.29

3.70

1.71 1.27

not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable not ascertainable ascertainable not ascertainable not ascertainable not ascertainable not ascertainable

Apyrogenicity: pyrogen free

1 sheet of drawings

Claims (16)

661 215 2nd PATENT CLAIMS 1. A method for producing a liquid containing a radioisotope for radiopharmaceutical applications by eluting a radioactive daughter isotope from a mother isotope which is adsorbed on an adsorbent with the aid of a physiological solution, characterized in that a mother isotope produced by cleavage is used and the eluate, which contains a daughter isotope, is purified with a cation exchange material. 2. The method according to claim 1, characterized in that molybdenum-99 is used as the mother isotope. 3. The method according to claim 2, characterized in that aluminum oxide, which contains wholly or partially hydrated manganese dioxide, is used as an adsorbent for the mother isotope. 4. The method according to any one of claims 1 to 3, characterized in that a cation exchange resin is used as the cation exchange material. 5. The method according to claim 4, characterized in that a strongly acidic cation exchange resin which has been neutralized is used as the resin. 6. The method according to claim 5, characterized in that a strongly acidic cation exchange resin, which has been converted into the Na +, K + or NH4 + form, is used as the resin. 7. The method according to any one of claims 4 to 6, characterized in that a cation exchange resin is used which has a particle size of 300 to 37 microns, preferably 150 to 74 microns. 8. The method according to any one of claims 1 to 7, characterized in that it is carried out in a device which comprises a reservoir which has a supply member for the eluent, an outlet member for the daughter isotope-containing eluate and the mother isotope has loaded adsorbent, and which device also contains a cation exchange material. 9. The device for carrying out the method according to claim 8, characterized in that it comprises a reservoir which has a feed member for the eluent and an outlet member for the eluate and the adsorbent with the mother isotope, the mother isotope entirely or approximately is carrier-free, and that a cation exchange material is present in the device. 10. The device according to claim 9, characterized in that the adsorbent is enclosed with the mother isotope in the reservoir between two filters, being on that side of the reservoir on which during the loading of the adsorbent, the solution of the mother isotope is let in Filter made of sintered glass. 11. The device according to claim 9 or 10, characterized in that the cation exchange material is also attached in the reservoir. 12. Vorrichtimg according to claim 11, characterized in that the reservoir is divided into two compartments, which are separated from each other by a filter as a separating filter, the periphery of which reaches the inner wall of the reservoir, being in a compartment which contains a supply element for the eluent , the absorbent for the mother isotope is present between the feed element and the separating filter, and the adsorbent is enclosed between a sintered glass filter and the separating filter, and wherein in the other compartment, which contains an outlet element for the eluate, the cation exchange material is present between the separating filter and the outlet element, and the space between the adsorbent particles and between the particles of the cation exchange material is filled with a physiological solution. 13. The device according to claim 12, characterized in that the separating filter consists of two disks, which completely or almost completely cover one another, the disk which contacts the adsorbent consists of glass fiber paper, while the disk which contacts the ion exchange material consists of porous Polyethylene exists. 14. Reservoir for carrying out the method according to claim 8, which has a feed element for the eluent, an outlet element for the eluate and the adsorbent for the mother isotope, the reservoir containing two filters, between which the adsorbent for the mother isotope is enclosed is characterized in that on that side of the reservoir on which the solution of the mother isotope is permitted during the loading of the adsorbent, the filter consists of sintered glass. 15. Reservoir for carrying out the method according to claim 8, which has a feed element for the eluent, an outlet element for the eluate and the adsorbent for the mother isotope, characterized in that the reservoir is both an adsorbent with the mother isotope and a cation exchange material contains. 16. Reservoir according to claim 15, characterized in that the reservoir is divided into two compartments, which are separated from one another by a filter, the periphery of which reaches the inner wall of the reservoir, wherein in a compartment which contains a supply element for the eluent, the Adsorbent is present with the mother isotope between the feed element and the separating filter and the adsorbent is enclosed between a sintered glass filter and the separating filter, and the other compartment, which contains an outlet element for the eluate, the cation exchange material between the separating filter and the outlet element and the space between the adsorbent particles and between the particles of the ion exchanger is filled with a physiological solution.