Production of 123I, 77Br and 87Y with the Eindhoven A.V.F
cyclotron
Citation for published version (APA):
van den Bosch, R. L. P. (1979). Production of 123I, 77Br and 87Y with the Eindhoven A.V.F cyclotron.
Technische Hogeschool Eindhoven. https://doi.org/10.6100/IR123784
DOI:
10.6100/IR123784
Document status and date:
Published: 01/01/1979
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PRODUCTION OF
123
I
77
Br AND
87
y
'
WITH THE EINDHOVEN A.V.F. CYCLOTRON
PROEFSCHRIFT
TER VERKRIJGING VAN DE GRAAD VAN DOCTOR IN DE TECHNISCHE WETENSCHAPPEN AAN DE TECHNISCHE HOGESCHOOL EINDHOVEN, OP GEZAG VAN DE RECTOR MAGNIFICUS, PROF. IR. J. ERKELENS, VOOR EEN COMMISSIE AANGEWEZEN DOOR HET COLLEGE VAN DEKANEN IN HET OPENBAAR TE VERDEDIGEN OP
DINSDAG 9 OKTOBER 1979 TE 16.00 UUR
DOOR
ROBERTUS LUCAS PETRUS VAN DEN BOSCH
DIT PROEFSCHRIFT IS GOEDGEKEURD DOOR DE PROMOTOREN
PROF. DR. IR. H.L.HAGEDOORN EN
Contents
0 Scope of this study 1 INTRODUCTION
1 3
1.1 General remarks 3
1.2 Ideal radionuclides .for use in nuclear medicine 4 1.3 Cyclotrons as radionuclide production machines 5 1.4 Introduetion to the present study 7 2 IODINE RADIOISOTOPES IN NUCLEAR MEDICINE 11
2.1 Introduetion 2.2 Nuclear data
2.3 Uptake, retentien and distribution of sodium iodide in the human body
2.4 Absorbed-dose estimates iDduced by radioactive iodine, administered as iodide
2.5 Absorbed-dose estimates induced by radioactive iodine labelled compounds
3 PRODUCTION METHOOS FOR IODINE-123 3.1 Introduetion
3.2 Target yield
3.3 Various reactions for 123 I production 3.4 Proton activatien of tellurium
3.4.1 Yield curves
3.4.2 The effects of the impurity level on the radiation dose and the imaging quality 3 4 3 . . Degree o f enr~c ment o . h f 124 · Te
4 TARGET CONSTRUCTION AND RADIOCHEM.ICAL SEPARATION FOR 11 12 17 19 21 25 25 26 29 43 44 46 50
123 I PRODUCTION VIA THE 124Te(p,2n) REACTION 53
4.1 Introduetion 53
4.2 Target material for the dry-distillation
procedure 58
4.2.1 Elemental tellurium 58
4.2.2 Tellurium dioxide 60
4.3 Separation by the dry-distillation methad 63 4.3.1 Volatilization yield of 123 I from Te02
under routine production conditions 63 4.3.2 Separation of 18F from 123 I "64 4.3.3 Further purification and trapping of
4.3.4 Finalset-up of the radiochemical
separa-tion apparatus 66
4.3.5 Loss of Teo 2 from the target and
contami-nation of 123r 68
4.3.6 Chemical farms of the radioactive iodine 69 4.4 Target construction and handling 74 4.5 Evaluation of the experience obtained in routine
production
5 THE PRODUCTION OF BROMINE-77 5.1 Introduetion 5.2 Nuclear reaction 5.3 Radiochemical separation 5.3.1 General aspects 5.3.2 Target material 5.3.3 Dry-distillation separation 5.4 Evaluation 76 79 79 82 88 88 89 91 94 6 PRODUCTION OF YTTRIUM-87 FOR AN
YTTRIUM-87/STRONTIUM-87m GENERATOR 97
6.1 Introduetion 97
6.2 Nuclear reaction and target material 99 6.3 Target construction and handling 103 6.4 Radionuclide purity
6.5 Discussion 7 CONCLUDING REMARKS
ADDENDUM RADlATION HAZARDS AND SAFETY MEASURES IN THE 104 109 113
PRODUCTION OF IODINE-123 115
A.1 Introduetion 115
A.2 Radiation risks at the irradiation facility 118 A.2.1 External radiation exposure during
bombardment 118
A.2.2 External radiation exposure after
bombardment 121
A.2.3 Internal radiation. exposure after
bombardment 122
A.2.4 Aspects of system failures 123 A.3 Radiation risks in the radiochemical facility 124
A.3.1 External radiation exposure during
A.3.2 Internal radiation exposure during separation
A.3.3 Aspectsof system failures A.4 The multisphere method
A.4.1 Introduetion A.4.2 Detection method A.4.3 Calibration
A.4.4 The unfolding procedure
A.4.5 Results for an 241Am/Be souree and a 252
ct
soureeA.4.6 Concluding remarks A.5 Evaluation
Appendix I Radiation emitted by 123 I and 131 I Appendix II Radiation quantities and units Appendix III Estimation of stopping-power values
interpolation References Summary Samenvatting Nawoord Levensloop 125 127 12 7 127 128 130 133 134 141 141 143 147 via 15 3 157 165 167 171 175
0 SCOPE OF THE STUDY
At the Eindhoven University of Technology (E.U.T.) the Cyclotron Applications Group of the Department of Physics and the Instrumental Analysis Group of the Department of Ch~mistry participated in a study entitled: "Investigation of Pröduction Methods for Short-Lived Radionuclides to be used in Nuclear Medicine" . ··
For this study we had the Eindhoven A.V.F.-cyclotron (Sch73) at our disposal. The external beam delivered by the cyclotron has the following properties:
-Energy range of accelerated.prptons: EP= 3-29.6 MeV. For ether particles wi th charge z1e and rnass number Ai the eer-responding energies E.
'
ar~:
E.~
z.?E /A ..l l . l p l
-Maximum beam current for protons ·and deuterons: 100 )JA; maximum beam current for 3He.,.particles and o.-particles:
50 )JA. .;.
The aim of this study has been 'to develop processes for radionuclide production that'fuifil the following main re-quirements:
'
-production of large quarttities of short-lived radionuclide~ in a short irradiation pe~iod, in order. to avoid as much as possible interference· with o·ther experiments carried out with the multi-purpose cyclotron ':
-simplicity, especially with respeç:t to target preparation, target recovery and radiochemical separation procedures in order to make routine production possible
-~eliability and low casts in order to make routine appli-cation possible in laboratorles for nuclear medicine.
Production methods for the following radionuclides were in-vestigated and developed:
123 . 124 123
- I (t~ = 13.2 h); nuclear react~on Te(p,2n) I;
chemical separation between the tellurium and iodine via a dry distillation procedure using Teo 2 as target material with 96.2% abundance of 124Te
- 77Br
(t~
= 56 h); nuclear~eaction
78 se(p,2n) 77Br; chemical separation between the selenium and bromine via a dry-dis-tillation procedure using Na2seo 3 . 88 .0.39Na2o as target material with 97.9% abundance of 78se87 . 88 87
- Y (t~ = 80.3 h); nuclear react~on Sr(p,2n) Y; the radioisotape 87Y is used for the production of a 87msr
(t~ = 2.8 h) generator; strontium roetal with natural iso-topic composition serves as target material.
Because of the radiation risks involved in the production of radionuclides in large quantities, attention is also paid to various health-physics aspects of the methods.
1 INTRODUCTION
Some general remarks on radionuclides for use in medicine are given. The properties of an "ideal" radionuclide for nuclear medicine and some requirements regarding its final state are summarized. An introduetion into the aim and philosophy of this study is presented.
1.1 General remarks
Medical applications of cyclotrons include neutron therapy, in-vivo activatien analysis, proton-induced X-ray emission analysis (PIXE) and radionuclide production for medical diagnosis. Bath PIXE (Hei79) and radionuclide production are under study in our laboratory. This thesis is limited to radionuclide production.
Radionuclides or the compounds into which they are built ("radiopharmaceuticals") may be administered to patients for medical diagnosis (in-vivo application) . Organs or lesions, in which the radionuclides are accumulated may be "visualized" by detection of the y-radiation emitted. The distribution of the radionuclides over the human body as a function of time can give information about the kinetics of the radio-pharmaceuticals concerned or their roetabelites and hence about the functioning of the tissues involved.
The demand for short-lived radionuclides in nuclear me -dicine is rapidly growing for two reasons: due to a short half-life, a·higher-activity dose can be administered, so that the interpretation of observations is based on better
At the previous page is an i~~ustration of the bunker now in use for radionuclide production (see also Fig. A.l).
statistics and a reduction of the radiation dose can be achieved. After the diagnostic information has been obtained the radioactivity should disappear as quickly as possible out of the body. The disappearance occurs via biological clearance and via radioactive decay of the nuclide. The ef-fective half-life (t~ eff) of the radionuclide in the body depends on the biological half-life (t~ biol) and the phys-ical half-life
(t~
phys)(t~
eff)- 1=
(t~
biol)-l +(t~
phys)- 1 . (N.B.: for simplicity i t is assumed here that the clearance of the radionuclide/radiopharmaceutical may be described by only one half-life). Reduction of the effec-tive half-life can be achieved by a short-lived radionuclide, especially if long-biological half-lives are involved.The present investigation was undertaken because existing production techniques for short-lived radionuclides are nat always adequate.
1.2 Ideal radionuclides for use in nuclear medicine For in-vivo use, the properties of radionuclides in their final farm should fulfil the following requirements:
-No emission of S-particles, conversion electrans ar Auger electrons. They do not contribute to the detection but only to the radiation dose.
-The physical half-life should be of the same order as the time involved in the physiological process of interest. -Emission of y-quanta between about 100 and about 300 keV.
These quanta penetrate tissues adequately, can be detected with relatively high efficiencies and permit the use of collimators with thin septa.
-Suitable chemical properties. The radionuclide may be used for preparatien of radiopharmaceuticals, by the labelling of organic molecules. In such cases the chemical properties of the radionuclide should permit a chemical bond with the organic molecule. If the radionuclide is used as such, the requirements may be even more stringent. For example, for certain thyroid studies only radioisotapes of iodine can be used.
-High-speci~ic activity (Bq*/kg or Ci*/g) . This means that
stable isotapes of the same element should be absent as much as possible. These products may be added to biological systems without a significant change of the metabolism of the stable element already present.
-High radionuclidic purity. This implies that a minimum of other radionuclides should be present. This requirement does not generally give rise to problems, except when isotopic impurities are involved. In the latter case these impurities cannot be removed chemically and sametimes sub-stantially increase the radiation dose to patients, because of their less ideal properties.
-High radiochemical purity. This means that the chemical state of the radionuclide should be such that i t ends up
in the organ under study. For instance, iodide is the re-quired chemical state for thyroid investigations.
-Sterility and apyrogenicity. Especially for short-lived radionuclides this requirement is important, because of the short time available for quality control and quality im-provement befare administration to patients.
-High chemica! purity. This means that no elements or com-pounds should be present in the product which are taxie or otherwise give unwanted effects.
-Chemica! and physical state should be such that a variety of labelling procedures and/or applications to patients are possible with a high radioactive concentratien of the radio-nuclide (Bq/liter or Ci/liter).
1.3 Cyclotrons as radionuclide production machines
Radionuclides for medical diagnostics are mostly produced in nuclear reactors or in cyclotrons~ In the first case neutron-excess nuclides are generally produced. In the latter case neutron-deficient nuclides are produced. Neutron-excess radionuclides decay byemission of 8--particles (hardly de-tectable outside the body) and neutron-deficient
radionu-*
Various physical quantities and units in the field of radiation andradioactivity are used in this thesis. Definitions are given in Appen-dix II. Quantities are mostly expressed in S.I.-units. The traditional units are also given.
elides by electron-capture or by emission of s+-particlesl resulting in annihilation radiation. The neutron-deficient
radionuclides generally fulfil the in Section 1.2 mentioned
requirements better than the neutron-excess radionuclides.
Cyclotrons have a large flexibility1 because the energy and the type of incident particles may be adjusted to give op-timum production of the radionuclide required. Charged par-tiele reactions may lead to radionuclides which are non -isotopic with the target material1 so that a carrier-free
(i.e. no dilution with the element isotopic with the radio-nuclide of interest) production is feasible. Because of these arguments i t is not surprising that the number of cyclotrons dedicated to radionuclide-production is increas-ing rapidly at the present time (Sil79). Cyclotrons can cover a wide range of energies. Table 1.1 lists some cyclotron parameters important for radionuclide production.
Table 1.1 Cyclotron parameters with regard to radionuclide production.
Parameters Maximum energy External beam Production capabi1ity Type - for protons (MeV) current+particles with reasonab1e yield
Gow energy 6-10 100-200JJA 1C, 13N, 150 , lBF 3
p,d, He,CI.
Medium energy 20-40 60-100JJA 11C, 13N, 150 , 18F p,d,3He,CI. f8 Mg, 43 K, 52 Fe, Ga, 67
~7Br,B1Rb,B5mKr,B7Y, 111In,1231 ,127Xe,201Tl, 203Pb and other radio-!rJUclides
ifiigh energy > 40 20-lOOJJA Muc;h wider range 3
p,d, He,a. of radionuclides
Low-energy cyclotrons are well suited for the production
of 11
C(t~
=
20 min) 1 13N(t~
=
10 min) 1 15o(t~
=
2 min) and 18F(t
~
=
110 min). These radionuclides emit B+-particlesl leading to annihilation radiat ion. Because of the short half-life of the radionuclides, the radiation dose to patients can be considerably reduced. The first three of the above radionuclides belong to elements which are present on a large scale in the hurnan body, so that many organs or lesions may be investigated. The radionuclide 18F may also be used as a label for various organic molecules. Because of the very short half-life of these radionuclides, the cyclotrons used to produce them should be placed at the user's site.
The mediurn-energy cyclotrons, like the Eindhoven Univer-sity of Technology cyclotron (Sch73), are more complex than the low-energy cyclotrons and require trained persennel for maintenance and operation, especially if they are designed for multi-purpose applications. These machines can produce a great variety of radionuclides for nuclear medicine~ They may fulfil a regional function, provided a distribution system and a pharmaceutical control organization is present.
The high-energy cyclotrons can be effici~ntly used only in large research institutes because they require substan-tial capital investment and extensive technical and financial support for maintenance and operation. The production possi-hilities of these accelerators is considerable. These cyclo-trons may also fulfil a regional function.
1.4 Introduetion to the present study
In December 1974 a project was started at the Eindhoven University of Technology, entitled: "Investigation of Pro-duction Methods for Short-Lived Radionuclides to be used in Nuclear Medicine". This project is a cooperative effort of the Physics Department (the Cyclotron Applications group) and the Chemistry Department (the Instrumental Analysis group, in particular the Radiochemistry section). The A.V.F.-cyclotron of the University was made available for this study.
The aim of the work is to develop methods for the pro-duction of radionuclides fulfilling the following require-ments:
-production of large quantities of short-lived radionuclides in a short irradiation period, in order to avoid as much as
possible interference with other experiments on the multi-purpose cyclotron
-simplicity, especially with respect to target preparation, target recovery and radiochemical separation procedures, to make routine production possible
-reliability and low casts in order to permit routine appli-cation in laboratories for nuclear medicine.
Investigations of labelling procedures with the radioisotapes produced or the investigation of applications in nuclear me-dicine are beyond the scope of this study. Only in the few cases where labelling procedures and applications in nuclear medicine set requirements as to the product itself any
at-tention is paid to these aspects. For example, the deelsion how to collect 123r and 77Br in their final form for deliv-ery (cf. Section 4.3.3 and Section 5.3.3) is based upon re-quirements set by labelling procedures and/or by direct ap-plications of the radionuclides in nuclear medicine; for the choice of the nuclear reaction for the production of 123 I, the quality of the scintiphotos was taken into account (cf. Section 3.4.2). A few problems in the field of target pre-paration and chemical sepre-paration, which we have met in our investigations, were solved phenomenologically. We felt that a fundamental and thorough research regarding these problems did not lie within the scope of the present study. For in-stance, phenomenological solutions were found for the re-moval of 18F (cf. Section 4.3.2) and for the impravement of
the chemical composition of 123 I (cf. Section 4.3.6).
. 123 77
Investigations on the product1on methods for I, Br and 87Y are described. Safety measures to reduceradiation hazards during the production of large quantities of radio-nuelides did form a part of the study. For this purpose a technique was developed for detection of neutrons, generated during proton bombardment.
Chapter 2 contains estimates of the radiation doses to patients induced by iodine radioisotopes. In Chapter 3 the nuclear reactions for the production of 123 I are considered. The 124Te(p,2n) 123
r
reactlon has been chosen. Thedescribed in Chapter 4. The methad for the production of 77 Br is given in Chapter 5. Chapter 6 deals with the investiga-tions on the production process for 87Y. Some concluding re-marks are made in Chapter 7. The radiation risks involved in the production of large quantities of 123
r
and the applied neutron-deteetien technique are discussed in the Addendum.2 IODINE RADIOISOTOPES IN NUCLEAR MEDICINE
In this chapter the most interesting radioactive iodine isotapes for
nuc~ear medicine are considered. The radiation dose of these radio -isotopes, administered as sodium iodide, to the thyroid is estimated. AZso the radiation doses invoZved with appZioations of some other radioactive iodine compounds are given.
2.1 Introduetion
Iodine occurs in a "standard man" (I.C.RiP.-2) to the extent of about 10- 6% (approximately 1 mg) . It is concentrated mainly in the thyroid. In 1938 Hertz-Roberts and Evans tried to investigate the iodine metabolism with
128
radioactiva iodine. They used I (t~ = 25 min) produced with the cyclotron of the Lawrence Radiation Labaratory
(Her75). However, this radioisotape has too short a half-life to trace the iodine metabolism adequately.
The first scintigraphic image of a thyroid was made by Cassen in 1951 (Cas51). He used the radioisotape 131 I
(t~
=
8.05 d) which was produced with a nuclear reactor via the reaction 130Te(n,y) 131Te(8-,t~
= 25 min) 131 I.The physical properties of 131 I (see Table 2.1) differ considerably from the properties required for an ideal nu-clide in medical diagnostics, mentioned in Sectien 1.2. The emission of high energetic B-particles (87% with 0.61 MeV) and the long half-life of 8.05 d increase undesirably the radiation exposure of patients. In addition, photons of
At ·the head of this page is the symboZ of the Society of NucZear Medicine.
different and relatively high energies necessitate the use of a collirnator with thick septa for irnaging. Generally this results not only in a decrease of detection efficiency but also in a decrease of spatial resolution. In spite of these unwanted properties, 131 I has been used for quite a long time, rnainly because its relatively low casts and com-mercial availability in large quantities.
In addition to 131 I, the isotapes 125 I
(t~
= 60 d) and 132 I(t~
= 2.4 h) have sametirnes been used for the exarnina-tion of the iodine rnetabolisrn. Although 125 I ernits no unde-sirable S-particles, i t is no langer applied for in~ivo rneasurernents because the y-energy is too low (36 keV) for adequate tissue penetration and detection, and also because of the too long half-life. The radioisotape 132 I has the advantage that i t can be produced by the longer-livinggen-132 132
erator systern Te (t~ = 78 h)/ I (t~ = 2.4 h). However, i t ernits S-particles and y-rays with very high energies
(Ey .5_ 2 MeV).
A n 10 1ne ra 101sotope o . d' d' . f . 1ncreas1ng 1nterest 1s . . . 123I
(t~
=
13.2 h). Since this radioisotape is neutron deficient, charged-particle bombardment is the obvious way for produc-tion. With the availability nowadays of rnany srnall and larger cyclotrons, sorne of thern specially constructed for radionuclide production, the dernand of 123 I, is rapidly in-creasing.2.2 Nuclear data
A total of 27 iodine radioisotapes are known now. Only a few of these are of interest for nuclear diagnostics. The
half-140 7 129
lives range frorn 0.8 s for I to 1.7 x 10 y for I. The radioisotapes with rnass nurnbers A = 117 to 126 are neu-tron deficient and decay by s+-ernission or, ,in the case of 123 I and 125 I, cornpletely by electron capture. The
radio-isotapes with A> 128 are neutron abundant and decay by S--ernission.
Camparing the properties of the iodine radioisotapes with those of an "ideal" radionuclide for rnedical diagnos-tics, i t follows that 123I is the most prornising one because
TabZe 2.1 NucZear dataforsome radioisotapes of iodine.*
Radionuclide 1231 1241 1251 1261 1301 1311 . 1321
Physical half-life 13.2 h 4.18 d 60.2 d 13.0 d 12.5 h 8.05 d 2.38 h Mode of decay E.C. (100) E.C. (77) E.C. (100) s- (44) s-(looJ 8 (100) s-(loo)
s+<23l E.C. (55) s+ < 1) Energy released per
disintegration by .172 1.1 .042 .418 2. 14 .381 2.28
photons (MeV). Energy released per
disintegration by . 0281 .194 .0194 .161 .297 .192 .494
betas and electrens (MeV). Principal y-rays .159(82.8) .603(61.0) .035(66.6) .666 (30) .418(34.2) .284(6.06) ;506( 5. 01) (Me V) .723(10.1) . 389 (32) .536(99.0) .365(81.2) .523(16.1) 1.691 (10.6) .669(96.1) .637( 7.27) .630(13.7) .734(82.3) .668(98.7) 1.157(11.3) .672( 5.2) .773(76.2) .812( 5.63) .955(18.1) 123Te("' 124 125Te 126 130 131 xe(98.2l 132
Daughter-nuclide 100) Te 126Te Xe Xe
t~ "' 1o13 Y (stable) (stable) Xe (stable) (stable) (.stable)
123!11Te(5.10-3) (stable) 131mXe(1.2)
t~ = 119.7 d
t~ = 11.9 d
~Abundance
is given between brackets in percent. The X-rays, y-rays and the annihilation radiation are taken in-te account for the energy released by photons. Only gamma-rays with abundances higher than 5% are given. The ~ energy released by daughter nuclidesis negligible. The data are taken from Dilman (Dil78).of the absence of B-particles, its appropriate y-energy of 159 keV and its favourable short half-life.
In Table 3.2 the reactions which may lead to the forma-t ion of this radioisotape are given. Since other radioi so-topes are produced simultaneously as impurities, attention is also given to 124 I, 125 I, 126 I and 130 I. The nuclear data of these iodine isotapes are given in Table 2.1. For comparison we have also included the data of 131 I and 132 I.
Since 131 I is the most commonly used radioisotape of iodine, we will campare the radioisotapes 123 I and 131 I to each other. In Fig, 2.1 and Fig. 2.2 the simplified decay
123 131 .
schemes of I and I are shown. In Append1x I a complete listing of all types of radiation emitted by bath radio iso-topes is presented. . 736 f'1eV . S29 He V . 03 % .08 % . 440 nev . 36 i'-. 539 He V . 27 1 . SOS l!eV . 26 % , . - - - -13.2 h E.C 100 ~~~~I
~
"--J f'-:347 !leV . 10 % .03 ' . l l % . 30 % l . 19 % . 30 % . 35 % 119.7 d. 9 7. 7 %f.
159 "•" 99. J ~ 1.2 x 1013 yBeta emission and
h
a
~
f-
~i
f
e
of 123I and 131 I Two important remarks must be made:Fig. 2.1
Simp~ified decay scheme of 123I. The gamma ray hlith the highest energy and gamma rays of which the abundance is higher than 0.1% are given
(Nuc76).
-Iodine-123 decays solely by electron capture. It emits no B-particles. However, i t emits conversion electrans and Auger electrons, which lead to an energy absorbed in human tissue per disintegration equal to 28.1 keV. Iodine-131 decays solely by B--emission, leading to an energy absorbed in human tissue per disintegration equal to 192 keV. De-tailed information is collected in Appendix I.
-The 13.2 h half-life of 123I is of the same order as the time required for the study of most of the relevant bio
lo-gical processes in the thyroid. The 8.05 d half-life of 131
I prolengs unnecessarily the radiation exposure to pa-tients, since the release of iodine from the thyroid is characterized by a long biological half-life (Eph72). The total number of disintegrations per unit of act ivity
ad-131 ministered to a patient is about 15-times larger for · I
than for 123
r.
Both arguments cause important factors for the reduction of the radiation exposure to patients when 131I is replaced by 1231. Moreover, the short half-life of 1231 offers a second advantage: fast physiological processes or changes induced pharmacologically can be evaluated. Thyroid studies may be repeated after 3 days since then only a few percent of the previous activity dose is present.
E8max .248 l-1eV 2.13 % E~max = .304 MeV .62 1. E!!max .334 HeV 7.36 % E~max = .607 r.leV 89.4 % E.:ma:.l .630 MeV .06 % E2max = .807 HeV .42 %
~
\
\
\\
r-
.
.
722 MeV l . 8 % . 284 He V 6. 4 ' .637 MeV 7. 3 \ Fig. 2.2Simplified decay scheme of 131
I. The gamma rays
feed-. 503 r1e
v ing the lJlmXe Level and . 36 %
. 17 . 33
the gamma rays of which the abundance is higher than 1% are given (Nuc76). 7 nev \ 1': 364 I 11.9 d. •ev 83 % • 7
F
a
Ne V 4 % 1~!xe (stable) 15123 131 Gamma-rays and X-rays of I and I
Following electron capture, the main transition of 1231 has an energy of 159 keV. 1ts abundance is 98.4%, consisting of
82.8% y-rays and 15.6% conversion electrons. The
half-thick-f f h · · 1 123 1 . 47 (W b79)
ness o water or t e pr~nc~pa y-ray ~s mm e .
This allows a sufficient penetratien even from parts of organs lying deep
ing 8 -emission, the main transition
in the human body, under the skin.
Follow-131
of I has an energy
of 364 keV. Its abundance is 83% consisting of 81.2% y-rays and 1.8% conversion electrons.
The principal y-line of 1231 has a number of advantages compared to the one of 131 1. The half-thickness of lead for the 123 I y-ray is less than 0.4 mm (Web79). For the y-line of 131 1, this half-thickness is 2.4 mm. Consequently, thin-ner septa can be used in collimators for 123 I, causing an increased detection efficiency and an improved resolution
(Gut77). Moreover, the small half-thickness of the 159 keV-line decreases the shielding problems in transport and
handling of the radioactive material.
1
Further, thedetec-tion efficiency of the 12.7 mm sodium-iodide crystals, com-monly in use for gamma-cameras, for the 123 I y-ray is a factor-of about 3 larger than the efficiency forthe princi-pal 131 I y-ray and is equal to 89% (Mye74).
The small half-thickness of lead and the relatively high detection efficiency in sodium-iodide crystals of the 159 keV-line permit small 123 I activity doses to be administered to the patient compared with 131 I. Conversely, when the same dosage is used for both nuclides, the scanning time may be shortened for 123 I. This may be an important consideration for medical diagnosis. However, if uptake studies are
neces-sarily performed 24 h after administration to the patient,
the advantages of 1231 are somewhat invalidated because of
its short half-life.
In addition to the 159 keV y-line of 123 1 13 other y-lines are emitted, with energies ranging from 184 to 784 keV. Their total abundance is 2-3%. The most important one is the 529 keV-line with an abundance of 1.4%. Apart from the 364 keV y-ray 131 I emits two important lines of higher energy, viz.
637 keV with an abundance of 7.3% and 723 keV with an abun-dance of 1.8%. These two lines have a high penetratien in lead, resulting in thick collimator septa,
In about 86% of the disintegrations of 123 r, K X-ray quanta, with energies between 27.2 keV and 31.8 keV are emitted. The half-thickness of water for these X-rays is in the order of 1.7 cm (Web79). The large difference in absorp-tion between the X-rays and the y-rays may enable a more exact localization of 123
r
under the skin.2.3 Uptake, retentien and distribution of sodiurn iodide in the hurnan body.
There are many applications of radioactive iodine in nuclear
medicine. It can be used either as sodiurn iodide for thyroid investigation or labelled to organic compounds for examina-tien of ether organs.
It is of interest to know the uptake, retentien and
dis-tribution of iodine compounds in the human body, not only for medical and radiation-detection reasens but also for
radiation-dose estimations. The chemical state of the iodine may be a dominant factor in the iodine metabolism. For the
thyroid, the relevant state is iodide as free anion because in this chemical form iodine is taken up from the blood-stream. Therefore, radioactive iodine must be administered to patients in the iodide form for thyroid investigations
(Ned78). The biological half-life of metabolized iodide in the thyroid is about 60 days (Mir75). Especially in this
I I-IOUR 6 HOURS 241-!0U~S 20 OAYS 80 OAVS
EXCRETED 87.1 EXCRE TEO 93.7 EXCRE TEO
Fig. 2.3
Estimated percentage of
iodine in tissues at var
-ious times after a single intravenous administration of iodide. Maximum thyroid uptake is assumed to be 15%
(Mir?5). ECEV =extra
-ceZZuZar-extravasauZar
space.
TabZe 2.2 Distribution fractions ah. and biological disappearance constants À. for various organs in case of
J J .
a single intravenous administration of sodium iodide to an euthyroid adult (maximum thyroid uptake 15%); parameters to be inserted in equation 2.1. Index h refers to organs. Data taken from (Mir75).
Souree organs ~1 À 1 ~2 À2 ~3 À3 ~4 À4 (h-1) (h-l) .. h""'1 h-1 Intestine 0.169 0.0994 0.000982 0.0488 -0.0000669 0.00492 0.000152 0.000498 Liver 0.0159 0.0994 -0.00130 0.0488 -0.00313 0.00492 0.00408 0.000498 Stomach 0.149 0.0994 0.000882 0.0488 -0.0000629 0.00492 0.000140 0.000498 Thyroid -0.154 0.0994 0.154 0.000498 Total body 0. 8'36 0.0994 0.164 0.000498
case, a radionuclide with a short physical half-life is effective for minimizing the radiation dose.
The fraction ~h (t) in an organ h of iodine administered as sodium iodide in one single intravenous dose as a tunetion of time (uptake and retention) can be described by a linear combination of j exponential terms:
-À.t
J ( 2. 1)
The factor ~h. is the distribution fraction for organ h; the factor
À.~h-
1
)
is the biological disappearance constant.J
The values for some relevant organs and the total body in case of a maximum uptake of 15% in the thyroid are given in Table 2.2; a good approximation may be obtained with a com-bination of 4 terros (j = 4). The histogram shown in Fig. 2.3 has been calculated by the MIRD-committee (Mir75) using the values given in Table 2.2 and summarizes the biological distribution of iodide in the body of an euthyroid person
(thyroid uptake 15%) at five different times. In a normal person the maximum thyroid uptake varies between 5 and 25%. The model does not apply to hypothyroid and hyperthyroid pa-tients or to iodine-deficient papa-tients. Also after adminis-tration of some diagnostic and therapeutic medicaments, the iodine metabolism may deviate from normal.
The radioactivity in an organ h is given by
A
0
·
~
h(t)·e-À\
where À is the physical decay constant of the radionuclide and A0 is the administered amount of activity. Knowing theactivity for each organ, the radiation exposure of a patient can be calculated.
2.4 Absorbed-dose estimates induced by radioactive iodine, administered as iodide
The radiation exposure may be expressed in the dose equiv-alent, which is equal to the absorbed dose* multiplied with
li'In this thesis different radiation quantities and units are used·. Definitions are given in Appendix II. The quantities wi11 be mostly expressed in the S.I.-units. The older units are given between brackets.
the so-called quality factor (Q). This dose equivalent has
the unit sievert (1Sv 100 rem; cf. Appendix II). The Q
factor equals 1 for all types of radiation relevant for this section, except for electrans below 20 keV. For these
low-energy electrans various ~ values are quoted in
litera-ture (Rui76), indicating a lack of substantial agreement as to the effect of this type of radiation (cf. Appendix II).
For this reason, we will only consider the absorbed dose
(i.e. Q
=
1).The absorbed dose rate in Gy/s (= 100 rad/s) of gamma quanta and X-ray quanta in an organ is given by:
D y ( 2. 2)
and of B-particles and electrans by:
A Ee f K/M. (2. 3)
The following symbols are used:
~Y absorbed dose rate for gamma/X-ray quanta (Gy/s). De absorbed dose rate for electrans and beta particles
A f x M K (Gy/s).
energy of the emitted X-rays or gamma quanta (MeV). mean energy of the emitted B-, S+-particles or energy of the emitted electrans (MeV).
activi ty (Bq).
abundance of tne emission of the relevant radiation.
linear absorption coefficient (m- 1 ). average radius of the target organ (m) . mass of the target organ (kg).
conversion factor from MeV to Joules (1.6 10- 13 J/MeV). The effect of annihilation radiation must be added to the
exposure rate of s+-emission.
The total exposure (dose commitment) of an organ h is given by:
(Ö
+D )e-ÀtdtY
e ' (2. 4)radio-nuclide. The values of ah(t) as a function of time can be calculated (cf. Sectien 2.3) for several organs. Data about the masses and the average radii of the various organs are given by the I.C.R.P. (I.C.R.P.-2). The values of the linear absarptien coefficients are listed by Weber (Web79). For the absorbed dose calculations, the radioactivity in each organ was assumed to be uniformly distributed.
A summary of the absorbed doses per unit activity admin-istered, expressed in Gy/Bq and in rad/mCi for pure 123
r
after a single intraveneus administration as sodium iodide to an euthyroid person with a maximum thyroid uptake of 15%, is given in Table 2.3 fora number of organs.TabZe 2.3 Estimated doses per unit of activity absorbed by a number f f . 1 • adm. . . f 123I d. o organs a ter a s~ng~e ~ntravenous ~n~strat~on o as so ~um
iod~de to an euthyroid adult, with a maximum thyroid uptake of 15%. Data are taken from (Mir75).
Absorbed dose per unit of activity Target organ
nGy/Bq (rad/mCi)
Liver 7.6 10- 3 (2.8 10- 2 ) Ovaries 9.2 10- 3 ( 3. 4 10- 2 ) Red marrow 8.1 10- 3 ( 3. 0 10- 2 ) Stomach wall 6.2 10- 2 ( 2. 3 10- 1 ) Testes 3.2 10- 3 ( 1. 2 10- 2 ) Thyroid 2.0 ( 7. 5) Total body 7.3 10- 3 ( 2. 7 10- 2 )
We have calculated the absorbed doses to the thyroid for 123I, l24I, 125I, 126I, 130I, 131I and 132I after a single intraveneus administration of sodium iodide for euthyroid persons. The results are given in Table 2.4.
2.5 Absorbed-dose estimates induced by radioactive iodine labelled compounds
The radioisotape 123
r
can be used to replace 131r
for ether studies besides thyroid investigation. After incorporationabl · · f h
·a
b ba a
·
f . . d 123 124 125 126 130 131 T e 2.4 Est~mat~on o t yro~ a sor e oses per un~t o act~v~ty ue to I, I, I, I, I, I and 132I after a single intravenous administration of sodium iodide to an euthyroid adult with maximum thyroid uptakes of 5%, 15% and 25%. The mass of the thyroid used for the calculation is 19.6 10-3kg. The average radius assumed for the thyroid is 30 mm. Data are in agreement with (Mir75).Absorb.ed thyroid dose per unit of activity in nGy/Bq (rad/mCi) Maximum thyroid 1231 1241 1251 1261 l30I 1311 1321 uptake 5 0.64(2.4) 48.6(180) 37.8(140) 86.4(320) 5.9 (22) 70.2(260) 0.62(2.3) 15 2.0 ( 7. 5) 143((530) 122 (450) 259 (960) 18.4 (68) 216 (800) 2.0 ( 7. 4) 25 3.5 ( 13) 241 (890) 214 (790) 43.2(1.600) 32.4(120) 351(1.300) 3.5(13.0)
Tah Ze 2. 5 Summary of estimated ahsorbed doses for "standard persons" to the kidneys due to sodium iodohippurate, to the Ziver due to radio -iodinated rose bengal and to the blood and fetus due to radioiodinated
human serum albumin (RIHSA) (Col76).
Radionuclide Absorbed dose per unit of activity in pGy/Bq (rad/mCi) for three radiopharmaceuticals and the corresponding
critical organ sodium rose
RIHSA iodohippurate bengal
kidneys liver blood fetus
123! 4.6 (1.7 10-2 ) 51 ( .19) 78 (0.29) 22.2 (8.2 10-2 ) 124! 27 (10 10-2 l 324 ( 1. 2) 2973 ( 11) 784 (2. 9) 125! 3.2 ( 1. 2 10-2 ) 38 (. 14) 2595 (9.6) 130 (. 48) 126! 17.0 (6. 3 10-2 ) 197 (. 73) 3514 ( 13) 892 (3. 3) 130! 43 (16 1o-2J 459 ( 1. 7) 622 (2. 3) 224 (. 83) 131! 19,7 (7.3 10-2 ) 219 (. 81) 3784 ( 14) 703 (2 .6)
of 123 I into an appropriate organic molecule ( 123 I-labelled radiopharmaceutical), various organs may be examined.
For instance: sodium iodohippurate may be used for the kid-neys, radioiodinated rose bengal for the liver and radio-iodinated human serum albumin (RIHSA) for vascular and extravascular pools (Col76). Using the calculation methods
as described in Sectien 2.3 and 2.4 the absorbed radiation
dose for these iodine-labelled radiopharmaceuticals can be estimated. The absorbed dose in the critical organ of a "standard person" are given in Table 2.5 for 123 I, 124
r,
125r,
126 I, 130 I and 131 I. It should be noted that in the calculation i t is assumed that the kidneys display a normalfunctioning. Kidney diseases may lead to langer residence
times of the iodinated compound. In such cases the advantage
123 131
factor of I compared to I may be substantially larger, up to a factor 70 (Ell78).
3 PRODUCTION METHOOS FOR IODINE-123
Methode for
cyc
~otron
production of 123Iwi~l
beeva~uated
,
invo~ving
various nuc~ear reactions. For the Eindhoven University of Techno~ogy
123 123 . . 124
cyc~otron the Te(p,n) I react~on
and
espec~a~~y the Te(p,2n)123I reaction seem to be most promising. The
y
ie~d
curves and the im -purity ~eve~s of both reactions were measured. The effect of the im -purities on the radiation dose and on the imaging qua~ity are eva~uate~3.1 Introduetion
A variety of nuclear reactions have been proposed for the production of 123I. In these processes, various natural or enriched targets and various bombarding particles with ener-gies ranging from a few MeV to a few tens MeV are involved. U f n ortunate y, 1 th e a van ageous d t propert~es . . o f 123 I w~ "th re-gard to the radiation dose and to the scintigraphic image quality can be partially undone by the presence of other iodine radioisotapes produced simultaneously with 123 I. The yield of the byproducts depends on the kind of nuclear re-action chosen for 123I and on the process parameters such as the energy of the incident beam, the target thickness and the degree of isotopic enrichment of the target material.
Various reactions which can produce 123I with n, p, d, 3He-or a-beams are shown in Table 3.1. As can be seen, the reactions lead to the formation of 123 I either directly or indirectly via the radionuclide 123xe. The indirect reac-tions do notproduce the medically undesired 124 I (cf.
Sec-At the head of thi8 page is an i~~ustration of the irradiation set-up,
showing the target ho~der, the target ho~derhousing and the target aoo~ing (see a~so Fig. 4.6).
Table 3.1 Several product~o. n metho s for d 123I w~. h t n, p, , d 3H e- or
d-particle beams.
Neutrons Protons Deuterons 3He-Particles Direct reactions 124 124 Xe(n,pn) Xe(p,2p) 126 Xe(p,a) 122 Te(p,y) 123T ( e p,n )a) 124Te (p, 2n) a,c) 125 Te(p,3n) 122 Te (d ,n )a) 122 Te He,pn (3 ) 123Te (d,2n) 124Te(d,3n) 123 3 Te( He,p2n) 121Sb(3He,n)a)
~ndirect
reaction via 123xe(
t~=2
.
1
h)a-Partic1es 120 Te(a,p) 122 Te(a,p2n) 121 Sb (Ct, 2n) a)
124Xe (n,2n) 127I (p,Sn)a,c) 127I (d,6n) a) 122Te(3He,2n) a) 120Te (a,n) 124xe (p, 2n) b,c) 124xe (d, 3n) b) 123Te ( 3He, 3n) a) 122Te (a, 3n) a)
124Te ( 3He ,4n) a) 127 I (a, Sn) a ,b)
127I(a,p7n)a)
a) Methods reported in literature, and discussed in sectien 3.3.
b) Indirect reactions via 123cs
(t~=5.9
m) and 123xe(t~=
2
.1
h). c) Theoretical1y determined cross sections reported in (Gra78).tion 3.3 and Chapter 2}, since 124xe is a stable nuclide. Several reactions mentioned in Table 3.1 have been investi-gated for production purposes by ether authors. In Sectien 3.3 we discuss these methods briefly.
3.2 Target yield
The production yield of radionuclides may be characterized by a reaction cross section. These cross sections denoted by a are generally expressed in barns (10-28
m
2 } inlitera-ture. The cross sections are a function of~the energy, which is continuously varying since the charged particles are
slowed down in irradiated material. In practice the use of cross section values is not convenient for the calculation of radionuclide production. The first reason is that the stopping power of the target for incident particles may vary for each nuclear reaction since the stopping power is a function of the type and energy of the incident particles as well as of the target composition. The second reason is that the isotopic composition of the target should be taken into account. A more convenient parameter for radionuclide production is the thin target yield Y(E), given in Bq/C MeV
(or mCi/~Ah MeV). It is defined as the activity per unit of electrical charge and per MeV energy loss for a given energy of th~ incident particle. This thin targetyield gives directly the production yield without using conversion fac-tors. The relation between the cross section and the thin target yield is given below.
The formation and decay of radionuclides during irradia-tion of a thin layer dx at penetratien depth x of the target is represented by:
dN(x)
~ <P n o (x) S dx - À N (x) •
The following symbols are used:
N number of the radionuclides produced -2 -1
partiele flux (m s . )
( 3 .1)
n
0
number of target nuclei per unit of weight (kg-1 ) nuclear cross sectien (m2 )
s
irradiated surface of the target (m 2 ) x penetratien depth in the target (kg/m ) ' 2decay constant of the radionuclide (s-1 ) t time(s).
With a constant partiele flux the number of radionuclides formed during time t in a thin layer dx at penetratien depth x becomes: N(x) -Àt ( 1 - e ) <P n o ( x ) S dx À · • (3.2) 27
In the thin layer dx, the activity A(x} (Bq} formed during time t is:
A(x} À N (x) <P n IJ ( x} S dx ( 1 - e -H ) . ( 3 . 3 }
The total activity produced in a target of thickness b be-comes: A b ( 1 - e-H) n <P S
J
IJ (x} dx . 0 (3. 4}The values of o(x} may be derived from these of IJ(E) and from range-energy relations.
Et
-1f
IJ (E) (dE)dx E dE, ( 3. 5)
where E0 and Ef are the energies of the particles, entering and leaving the target, respectively. The physical quantity dE/dx (MeV m2 /kg) is the stopping power. More information about the stopping power as a function of energy and target composition is given in Appendix III. After substitution of known values of constants, a formula for the target yield per MeV energy loss and for irradiation times much shorter than the half-life of the radionuclide, can be written as:
32 ÀIJ(E) Y(E)
=
37.6 x 10 M(dE/dx}E where Y(E) is given in GBq/C MeV, or as:Y(E) = 3.65 x 10 32 ÀIJ(E} M(dE/dx}E '
(3.6a}
(3.6b}
where Y(E) is given in mCi/~Ah MeV. The quantity Mis the
atomie mass (kg/mol} .
Using equations (3.6a) and (3.6b) and the values of dE/dx, we have converted the cross sectien functions for 123 r production given in literature to a more convenient expression for radionuclide formation.
3 3 . Var1ous react1ons . . f ar 1231 pro uct1on d .
I n T abl e 3 2 . t e react1ons h . f ar 123I pro uct1on reporte d . d . 1n the literature are summarized with their most important parameters. If target yields were nat avai1ab1e from the 1iterature, we have calculated them.
The choice of the most suitab1e reaction does nat on1y depend on physical parameters such as the production yield, the radionuclide impurity level and the maximal allowable beam current. Other important parameters are separation time, radiochemical yield, price and recovery yield of the target. However, thé choice of reaction is nat only a re-sult of technical considerations but also dictated by local circumstances. In this respect, the properties of the
available partiele accelerator and the organizational structure of the institute are important factors. For in-stance, if the partiele accelerator laboratory is a part öf a big hospital, the requirement regarding a high 123 I yield and a low impurity level of radionuclides with a long half-life will be less stringent, because of the short transport
123 time. On the other hand, if a cyclotron produces I for medical ins t i tutes a large di stance away, the cho.ice of the nuclear reaction is influenced by the transport facilities.
A discussion on the reactions listed in Table 3.2 with their advantages and disadvantages is given below.
Indirect high-energy reactions
Of the high energy nuclear reactions leading to the forma-tion of 123
r
via the 123xe precursor two reactions are suitable for the production of sizeable quantities of 123 I:127
I (p,Sn) 1123 Xe
127I (d,6n) t~ - 2.1 h
Use of these reactions have been reported by various authors (Wil75, Wei74, Cun76, Sym78, Fus72, Bic76, Paa76, Cun78, Lun79). The yield curves of these reactions are given in Fig. 3.1 and Fig. 3.2. We have calculated them with formula
(3.6) using the cross sectien functions given in (Wil75) and
Table 3.2 Iodine-123 pPoduction methods, with theiP most impoPtant paPametePs (see also next pages).
Indirect
e
e
llE Target Target Yield--+
re action n e Eo hickness costs (EOB+6. 64h) -(via 123 Xe) % % Me V Me V mg/cm 2 Hfl/cm2 mCi/].JAh 121I% 127I(p,5n) 100 100 64 16 2644
---
)i. 4 I 2.43 192 4.24 127 I(d,6n) 100 100 78 14 1643 --- 15.9 .0.91 122Te(a,3n) 2.5 90 40 8 88 3,000 1.5 (u) 122Te(3He,2n) 2.5 80 15 7 49 1,650 0.013 (u)123Te( 3He,3n) 0.89 87.5 36 14 161 12,700 0.60 (u)
124Te( 3He,4n) 4.62 96.2 60 22 366 6,740 1. 19 (u)
a) The half-life of 121I is 2.1 h. Via the production of 122xe
(t~ =
20.1 h) 122I(t~ =
3.6 min) may also be formed (Wei74, Sym78, Lun79). Directly produced radioactive iodine isotopes are not im-portant after the separation of xenon from the target. Indices re-fer to re-ferences given in the last column. (u) = unknown.
0n
=
natural abundance; 0e=
enrichment; E0=
energy of enteringpar-ticle; llE
=
energy absorption in the target.Impurities at EOB + 6.64 ha) ref. Re action 1251 Re action 131 1 Reaction
121 125 via 131 xe via Xe % via Xe % 127I(p,7n) 127I(p,3n) Wil75' 0. 1'
---
--- Vaa762 0.132 Sym78 3 Lam784 127 r(d,8n)o.
1 127 r(d,4n)---
--- Wei74 120Te(C1.,3n) (u) 122+x.re(CI., (u) 128Te(et.,n) Iod76[ x+l] n) 130 Te (a.,3n) x=0,1,2,3,4.
120Te( 3He,2n) (u) 123+xT (3 e He, (u) 130Te( 3He,2n) Iod76 [x+l]n)
x=O ,1,2.
120+xTe(3He, 0.01 123+xTe(3He, (u) 130Te( 3He,2n) Gui75
[x+2] n) [x+l]n)
x=0,2,3. x=0,1,2,3,5
120+xTe(3He, 0.015 123+xTe(3He, (u) 130 Te( He,2n) 3 Gui75
[x+2] n) [x+l]n)
x=O ,·2, 3, 4,5 ,6. x=O, 1 , 2 , 3, 5 , 7.
Table 3.2 (continued). Target Target Yield
___.
Directe
e
E LIE thick-casts (EOB) re action n e 0 ness % % Me V Me V mg/cm 2 Hfl/cm 2 jnCi/l.IAh 124! % 121 sb(Cl,2n) 57.3 57.3 25 2 16 --- 0.21 0.5 121 sb(Cl,2n) 57. 3 57.3 30 30 177---
1. 32 2.5 123sb(3He,3n)~ 42.7 42.7 26 8 74---
0.41 1. 23 122Te(d,n) 2.5 87.8 9 2.6 80 2,700 0.13 0.09 123T e ( p,n )c) 0.89 87.5 15 2 100 7,900 4. 1 0.5 124 c) Te(p,2n) 4.62 91.9 25 5 422 3,600 24. 3' 1. 08 16.72 124Te(p,2n)c) 4.62 96.2 25 5 422 7,800 24.7d) 0. 78 17.5 124Te(p,2n)c) 4.62 99.9 25 5 422 61,500 18.2 0.63b) The contribution of the 121 sb( 3He,n) 123 r reaction is negligible. cl The isotopic composition of the target material is given in Table
3.3. The produced impurities 128r
(t~
=
25 min) and 129r(t~
=
1.6 x 107y) are negligible.d) The value 24.7 has been derived from our yield curves given in Fig.
3.11. The value 17.5 is basedon our routine production during the last three years; this value is more in agreement with the curve of Kondb, given in the same figure.
Impurities at EOB ref Reactien 125I Reactien Other Reactien
% impurities
%
121Sb(a.,n) 0.7 123sb(a.,2n) 0.13 126I 123Sb(a.,n) Si172
123sb (a.,3n) Sed69
121Sb(a.,n) 0. 52 123sb(a.,2n) 0.24 126I 123Sb(a.,n) Wat73a
123sb(a.,3n) Ied76
123Sb( 3He,2n) (u) 123sb( 3He,n) 956 121I 12 \sb( 3He,3n) .Wat73a 123Te(d,n) 0.02 124 Te(d,n) 0.04 126I 125Te(d,n) Sed72 124Te(d,2n) 125Te(d,2n) 0.9 130I 130Te (d,2n)
0.09 131I 130Te(d,n) 124 .
Te(p,n) (u) 125 Te(p,n) 0.06 126I 126 Te(p,n). this 125
Te(p,2n) 126 Te (p,2n) 0.3 130I 130 Te(p,n) werk 124
Te(p,n) (u) 125 Te(p,n) (u) 126I 126 Te(p,n) Ace75a' 125
Te(p,2n) 126 Te (p,2n) 128 Te(p,3n) Ken77a2
126
Te(p,3n) (u) 130I 130 Te(p,n)
124
Te(p,n) 0.01 125 Te(p,n) 0.07 126I 126 Te(p,n) this 125
Te(p,2n) 126 Te(p,2n) 128 Te(p,3n) werk 126
· Te(p,3n) 0.02 l30I 130 Te(p,n)
124
Te(p,n) 0.001 125 Te(p,n) 0.001 126r 126 Te(p,n) Ken77a 125
Te(p,2n) 126 Te(p,2n) 128 Te(p,3n) Lam78 126
Te(p,3n) 0.003 130r 130 Te(p,n)
(u) unknewn.
102,-- - ---.-- - ---,-- - - --,-..,1Q3
~6-64h)
-2' ; ; - - --,_-1;,--- --f:--- --:f.:---110 40 50 60 70
proton energy (Me Vl
~ :::;: u 101
o-~ 60 70 80deuteron energy (MeV)
Fig. 3.1
Yie~d
curves for the 127r
123 123 .(p,Sn) Xe + I react~on,
calculated with formula (3.6)
using data given in (Wil75).
The estimated yields of 123I
and 125
r,
at the optimum sepa-ration moment (6.64 hafter
EOB), are a~so indicated.
Fig. 3.2
Yie~d
curves for the 127I123 123 .
(d,6n) Xe + I react~on,
ca~culated with formula (3.6)
using data given in (Wei74).
'-h . d . ' J
f
123T- e est~mate y~e&as o I
and 125I, at the optimum
sepa-ration moment (6.64 hafter
(\<Tei 7 4). In the same figures also .the yield curves for 125 xe leading to a 125 r impurity are given. The relevant reactions are:
127I (p,3n)l125 Xe
127 I (d, 4n)
t~
= 17t?
60 d).Separation of xenon nuclides .from the target is possible during ar after irradiation, using the volatility of xenon. The xenon is trapped and then allowed to decay for some hours after end of bombardment (EOB) . During this period a f ract~on . o f 125 Xe a sa ecays, . 1 d 1 ea d~ng . to t e ra h d. . ~o~sotop~c . impurity 125 I. The optimum decay time with respect to maxi-mization of the 123 I production is 6.64 h after EOB. Then the radioisotapes of iodine and xenon are separated. After
125 this chemical separation the impurity level due to I amounts to 0.2% and will grow because of its long half-life. The radiation dose to the thyroid per administered unit of
· · f 123 I f ' f . l b EOB
act~v~ty o as a unct~on o t~me apse etween and administration is given in Fig. 3.12 which shows that
. i f . f 125 I l . h ' h d' .
an ~ncreas ng ract~on o resu ts ~n a ~g er ra ~at~on dose to patients. For imaging purposes, the 125 I contamina-tion does not affect spatial resolucontamina-tion because all its emitted photons are less energetic than the 159 keV gamma
123 .
ray of I (see Table 2.1).
Advantages of the indirect, high-energy production methods are (Iod76):
-cheap target material (100% natural 127 I abundance)
-high yields, particularly for the (p,5n) reaction
-absence of iodine impurities, with high energy gamma quanta -only impurity: 125 I
(~
0.2% EOB)-simple chemical separation procedure. Disadvantages are:
-requirement of high energy cyclotrons with expensive eperation casts; these cyclotrons are not commonly avail-able
-complicated target construction, because of heat dissipa-tion and corrosive nature of liquid targets for on-line separation.
Other indirect, rarely used high energy reactions are 127I(a,Bn)123Cs ~ 123Xe ~ 123I, 127I(a,p?n)123Xe ~ 123I
(Ea = 102 MeV) (Lam76) or spallation reactions with 1 GeV protons on cesium, barium and lanthanum (Ale78) .
Indirect low or intermediate-energy reactions
The following are important indirect low or intermediate energy reactions: 122Te (a,3n) 122Te ( 3He,2n) 123Xe
s+
123I 123Te ( He,3n) 3 t~ = 2.1!?
124Te ( He,4n) 3Production methods, based on these reactions have been re-ported by various authors (Blu69, Leb71, Ter71, Sod73 and Gui75). The yield curves for the reactieris are given in the Figs. 3.3, 3.4, 3.5 and 3.6. We have calculated the curves of the first two reactions with formula (3.6) using the cross sectien ~unctions given in (Iod76). The curves of the latter _two reactions were derived from data given by (Gui75).
The yield curves for the formation of 12
~xe,
where re-ported in the literature, are given in the same figures.125
This radionuclide is the precursor of I and gives rise to an impurity level of 0.4% or less (6.64 h after EOB; cf. the indirect high energy reactions). If the 122Te ( 3He,2n)
. . d 125 . d d 1' . .
reactlon lS use , Xe lS pro uce on y Vla reactlons on tellurium isotapes ether than 122Te; thus 125
r
impurities may be minimized by irradiation of highly isotopically en-riched 122 Te targets. For the other reactions 125 xe is also produced via the enriched nuclide. Also here i t is possibl~ to separate radioactive xenon from the target during or after irradiation (Blu69, Lam72).The above mentioned reactions are nat very efficient be-cause of the rather low yield and, in the case of on-line separation because of the very vulnerable target conditions
.<::; <( .3 -1
u
10 E .r:. <(3
ü E 10 30 40 50O<.-particle energy <MeVl
123r <EOB+ 6.64hl
10 20 30
3He-partiele energy <MeVl
> "' ::;: u
1cfl
g
\2 -o a; ;;._ Fig. 3.3Yieûi curve for the 122Te
123 123 .
(a,3n) Xe ~ I react~on,
calculated with formula (3.6)
using data given in (Iod76),
for a 100% enriched 122Te
tar-get. The 123I yieZd, at the
optimum separation moment
(6.64 hafter EOB) is aZso
indicated.
Fig. 3.4
YieZd curve for the 122Te
3 123 123
-( He,2n) Xe~ I react~on,
caZcuZated with formuZa (3.6)
using data given in (Iod76)
for a 100% enriched 122Te
tar-get. The 123I yieûi, at the
optimum separation moment
(6.64 hafter EOE) is also
indicated.
---,---,-
-
,---,----,-,-
-
-,-,-
-
J
101 123Te
C
~e.3nl
123xe
~
J>
(l) 1Ö1 :::<: 123rcEOB+6.64hl L <( 2-u E 100 101 124Tec3He,4nl l23xe 100 ~ 123ICEOB+6.64hl :::<:>
L (l) <( :::<: 2- !2 ü CT E CD ~ -o Qi 1Ö1 >--3 10 30 40 50 603He partiele energy
Fig.3.5
Yieûi curves for the 123Te
3 123 123 .
( He,3n) Xe + I react~on,
caZcuZated using data given in (Gui75). The yieZds of 123I
d 125 h .
an I, at t e opt~mum
sepa-ration moment (6.64 hafter
EOB) are aZso indicated. The
degree of enrichment of 123Te
• % h 125I . . ~s 87.45,. T e ~mpur~ty . d d . h 123T ~s pro uce v~a t e e 3 125 . ( He,n) Xe react~on. Fig. 3.6 . ûi f h 124 Y~e curves or t e Te 3 123 123 . ( He,4n) Xe + I react~on,
caZcuiated using data given in
(Gui75). The yieûis of 123I
and 125I, at the optimum
sepa-ration moment (6.64 hafter EOB) are aZso indicated. The degree of enrichment of 124Te is 96.2%. T e h 125I ~mpur~ty . . . du d . h 124 ~s pro ce v~a t e Te 3 125 . ( He,2n) Xe react~on.
