XA IAEA-TECDOC-1051

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1 XA IAEA-TECDOC-1051 Management

2 The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies

3 The originating Section of this publication in the IAEA was: Waste Technology Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria MANAGEMENT OF RADIOACTIVE WASTE FROM "Mo PRODUCTION IAEA, VIENNA, 1998 IAEA-TECDOC-1051 ISSN IAEA, 1998 Printed by the IAEA in Austria November 1998

4 FOREWORD 99m Tc used for labelling different pharmaceuticals is the most important radionuclide in nuclear medicine practice,

5 EDITORIAL NOTE In preparing this publication for press, staff of the IAEA have made up the pages from the original manuscript(s). The views expressed do not necessarily reflect those of the IAEA, the governments of the nominating Member States or the nominating organizations. Throughout

6 CONTENTS 1. INTRODUCTION BASIC PROPERTIES OF MOLYBDENUM-99 AND TECHNETIUM-99m Nuclear properties Decay scheme of "Mo External irradiation...

7 Waste from the Karlsruhe silicide process Waste generated from

8 Molybdenum-99 ( 99 Mo) 1. INTRODUCTION

9 still the only proven process for the production of large quantities of 99 Mo with very high specific activity.

10 99m 43 T, O-1427 CM4O T c i O-O FIG. 99m-, 99 j

11 2.2. DECAY SCHEME OF 99 Mo Figure 2 shows the decay scheme for the 99 Mo: 99m Tc: 99 Tc: 99 Ru system with respective half-lives and branching factors. The decay growth curve for the

12 2.4. INTERNAL CONTAMINATION Both

13 In 1966, Brookhaven National Laboratory withdrew from

14 Several countries have worked at developing discontinued Mo production technology, but have

15 TABLE I. SUMMARY OF METHODS OF w 99 Mo A PRODUCTION Method Advantages Disadvantages 235 T 1. Nuclear fission of U using 235 HEU (>20%,99*

16 Irradiated targets off-gas treatment! Trap 133 Xe «, (molecular sieve) Declading Dissolve targets \ ' V 133 Xe Product

17 At Fleurus a process is installed for triple production of Mo, ' land "Xe from the irradiated targets [20]. A simplified flow diagram of the isotope production facility is shown in Fig. 5. The irradiated targets are dissolved in 3M NaOH and 4M NaNC>3. Radioxenon is released from

18 Dissolution of targets NaOH-NaNOs Xe-133 Trapping Filtration Mo Acidification by cone. HNOs 1-131

19 Mo production

20 The acidified molybdenum solution first enters a column filled with an ion-exchange resin (SM-7)

21 Purification of

22 All technological steps of Mo recovery process are carried out under strict control to ensure high

23 The aqueous homogeneous nuclear reactor as a prospective

24 1.

25 250 g Uranium Dioxide (UOa) Magnesium Oxide (MgO) UO2 (Minor MgO) 8/W Nitric Acid "Rocket" can irradiated in HI

26 Chimney [ Irradiation F. E. Production FIG.

27 A procedure was proposed to dissolve uranium silicide targets based on the results of research and development work [27-31]. A conceptual flow diagram of the dissolution process is shown in Fig. 12. Initially, the cladding and the aluminum in the irradiated target are dissolved in 3M sodium hydroxide. Sodium nitrate (NaNC>3) is added to the system to minimize hydrogen production during the dissolution. A gas sparge is used during the dissolution

28 Off-gas Treatment Irradiated Target Dissolver NaOH& NaNO 3 30 wt% H 2 C and NaOH Dissolving Solution from the 2nd dissolution step Filtration Cladding Solutions (from the 1st dissolution step) U0 3,Si

29 addition to the solvent extraction system, before the introduction of fission based

30 Solvent extraction and sublimation systems have proved practical and economical, and capable of providing 99m Tc solutions of high activity and quality for the labelling of radiopharmaceuticals kits. Medium

31 Second,

32 TABLE r 235 T

33 TABLE III. MAIN INITIAL ACTIVATION PRODUCTS IN THE IRRADIATED TARGET Radionuclides T 1/2 (h) Activity (MBq) % Total: 2.03E A1 3.73E E Mn 2.58E E Cu 1.27E E Zn 9.50E E Cu 8.50E E H Ag 6.83E E m Co 1.75E E S 2.62E E H 1.08E±5 1.60E Li 2.34E E Be 5.56E E ' 8 Ag 3.95E E After

34 Calculation of actinides On the basis of the mentioned uranium composition, a list is obtained of 129 nuclides representing

35 TABLE

36 TABLE VII. COMPOSITION

37 TABLE VIII. CHARACTERISTICS

38 targets liberates xenon, which is used for commercial applications (see Section 3.2.2). During the acidification iodine is released and also used for commercial applications. From the alumina column (Woelm A, Super 1) the fission products are eluted and fixed in the effluents (ILLW) and in the subsequent washes of IN HNCh, H

39 TABLE

40 TABLE XI. AVERAGE RADIOLOGICAL COMPOSITION (THE MOST IMPORTANT ISOTOPES) OF LOW- AND INTERMEDIATE LEVEL WASTE RECEIVED

41 Solid waste is generated mainly from spent ion exchange resins and the alumina columns. Another type

42 99* The Mo purification process generates only small volumes of liquid wastes with relatively low activity level and consist mainly of aqueous solutions or solutions of low concentration of acidic or alkaline media. Solid waste is generated mainly from spent ion exchange resins and absorber columns. Other types of solid waste are filter material from off-gas cleaning, such as activated charcoal filters or absolute filters from the hot cell. Small amounts of solid waste arise from the replacement

43 The remaining U and fission products are retained in the dissolver by Bio-Rex-5 containing

44 Waste generation by

45 TABLE XVI. CHARACTERISTIC

46 TABLE. XVH. CHARACTERISTICS

47 TABLE XVIII. TYPICAL COMPOSITION OF ANSTO INTERMEDIATE LEVEL LIQUID WASTE Chemical composition Component U Mg Fe NH 4 + HNO-, Concentration

48 TABLE XIX. POTENTIAL WASTE STREAMS

50 and distilled water.

51 On-site treatment

52 Hydrogen which is evolved during target plate dissolution is either oxidized by hot copper oxide,

53 release into the in-cell area, all process steps of

54 charcoal. To prevent release of iodine and (3 particulate aerosol to the environment additional absolute filters

55 Intermediate level aqueous waste formed during other process steps of 99 Mo production

56 and The intermediate level non-fissile liquid waste is temporarily stored in a storage tank

58 and

59 TABLE XXIV.

60 For the treatment of La Celeine containers filled with medium and high level waste the following options

61 Solid waste If

62 the route for treatment, conditioning and disposal that is available for this kind of waste. More information

63 continuously enforcing more strict requirements concerning the management of radioactive waste.

64 Finally, total waste volumes

65 waste types to collection, transport, treatment, conditioning, interim storage and final disposal. These facilities could be (or are) used for processing of waste generated during 99 Mo production. The treatment and elimination of gaseous waste can be solved by implementing an efficient it off-gas system, and for decay or commercial use.

67 REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Fission Molybdenum for Medical Use, IAEA-TECDOC-515, Vienna (1989). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Alternative Technologies for "Tc m Generators, IAEA-TECDOC-852, Vienna (1995). [3] HELUS, R, COLOMBETTI, L.G., Radionuclide Production, Vol. II, CRL Press, Inc. (1983) [4] INTERNATIONAL ATOMIC ENERGY AGENCY, Radioisotope Production and Quality Control, Technical Reports Series No. 128, IAEA, Vienna (1971) [5] TUCKER, W.D., GREEN, M.W., WEISS, A.J., MURRENHOFF, A.P., BNL 3746, American Nuclear Society Annual Meeting, Los Angeles, CA, Trans. Am. Nucl. Soc.,v.

68 [19] SALACZ,

69 [36] INTERNATIONAL ATOMIC ENERGY AGENCY, Advisory Material

70 Basmanov, CONTRIBUTORS TO DRAFTING AND REVIEW

ENERGY Defense Nuclear Nonproliferation

ENERGY Defense Nuclear Nonproliferation Argonne-Design Annular LEU Foil Target and Possible Use in Alkaline-Based Processes George F. Vandegrift Chemical Sciences and Engineering Division Argonne National Laboratory 1 Annular LEU-Foil Target