(a) Tritium label

HTO is water in which one of the atoms of the ordinary hydrogen isotope JH is replaced by an atom of another hydrogen isotope 3H, also known as tritium. This hydrogen isotope is radioactive with a half-life of 12.32 y.

Tritium disintegrates through a process where beta particles with energies up to 18 keV are emitted. Owing to its low energy, the penetrative power of beta
radiation from tritium is low: a sheet of paper can stop the particles. Thus, there will, in practice, be no radiation from a tritium tracer outside its containments, e. g. bottles, tubes and pipelines. Tritium does not emit gamma radiation. However, during the use of HTO as a water tracer, the injection process is often monitored by adding a modest quantity (10-100 kBq) of 131I- to the primary tracer mixture (i. e. a tracer in the tracer) in order to facilitate the monitoring of the injection process itself.

The only way of receiving a radiation dose from tritium is by intake, i. e. through mouth or by inhalation. As the tracer is kept in a closed system during the injection process, there is, generally, no possibility of HTO intake under normal conditions.

Tritium is common in the environment. It is produced continuously in the atmosphere from cosmic ray interaction with atmospheric molecules. Tritium is also generated in nuclear power production and in nuclear bomb tests. The global inventory of tritium is in the order of 5 x 1010 GBq. Approximately 99% of the tritium inventory is in the form of HTO. The concentration of 3H in sea water off the coast of northern Europe is in the order of 1 kBq/m3. The water volume of the North Sea is approximately 5 x 1013 m3, and the inventory of tritium in this sea volume is, therefore, in the order of 5 x 107 GBq.

For tritium labelled radiotracers, the D2 value applies. Table 1 [24] gives D2(3H) = 2 x 103 TBq, which is about 500 times higher than the upper estimate of applied quantities of HTO per injection in oil reservoirs. For tritiated methanol, CH2TOH, which is also a useful water tracer under certain conditions, the normally injected quantities are a factor of 10 lower than for HTO. Accordingly, it may be concluded that the tritium activities described here do not approach being defined as dangerous quantities.

(b) 14C and 35S labels

Typical water tracers labelled with 14C is thiocyanate, S14CN-, and cobalthexacyanoferrate, Co[(CN)514CN]3-, while a typical water tracer with the 35S label is 35SCN-. In water solutions, these are all anions and not volatile.

Carbon-14 is produced naturally in the upper atmosphere by the reaction of neutrons originating from cosmic rays with nitrogen and, to a lesser extent, with oxygen and carbon. The natural steady state inventory of 14C in the biosphere is about 1019 Bq, or 10 EBq (about 300 million Ci), most of which is in the oceans. Large quantities of 14C have also been released to the atmosphere as a result of nuclear weapon testing. Weapon testing through 1963 added about 3.5 x 1017 Bq, or 350 PBq (about 9.6 million Ci), an increase of 3% above natural steady state levels. Carbon-14 is also made commercially for use in medical, biological or technical tracer research described in this publication.

Carbon-14 is produced in nuclear reactors by the capture of neutrons by nitrogen, carbon, or oxygen present as components of the fuel, moderator, or structural hardware.

Carbon-14 is a pure beta emitter with a half-life of 5730 y and a maximum energy of Epmax = 156.4 keV. The range of these beta particles in air (20°C) is 22 cm and in stainless steel or Monel <50 pm. Hence, the beta particles do not penetrate the walls of the combined transport and injection container.

The only way of receiving a radiation dose from the 14C labelled molecules described above during the injection phase is by intake (e. g. through mouth) or by liquid spillage on the skin. As the tracer during the injection process is kept in a closed system, there is, generally, no possibility of tracer intake or human skin contamination under normal conditions.

For 14C on this basis the D2 value applies. This value is D2(14C) = 50 TBq (1.35 x 103 Ci) as compared with the actual injection quantities of 3.7-37 GBq (0.1—1.0 Ci) which is more than a factor of a 1000 lower. Hence, it may be concluded that the 14C activities described do not approach levels defined as dangerous.

For 35S labelled SCN—, the same evaluation and conclusion as for 14C is valid. Also, 35S is a pure beta emitter with a half-life of 87.4 d and with a maximum beta energy of Epmax = 167 keV, close to that of 14C.

The D2 value is also similar, D2(35S) = 60 TBq (1.62 x 103 Ci) as compared with the actual injection quantities of 3.7—37 GBq (0.1—1 Ci), which gives the same conclusion as for 14C above.

(c) Gamma emitting labels

Each of the labels 57Co, 58Co, 60Co, 125I and 131I will be treated separately.

Cobalt-57 is produced in charged particle reactions at accelerator facilities (for instance by the reactions 55Mn(a,2n), 56Fe(d, n) or 59Co(p,3n) plus beta decay) and there is no sizable production (or natural inventory) in either the biosphere or geosphere. It decays by 100% electron capture with a half-life of 271.74 d. The main gamma energies are low, the four strongest being 14.4 keV (9.16%), 122.1 keV (85.60%), 136.5 keV (10.68%) and 692.4 keV (0.15%).

Being a gamma emitter, special precautions have to be taken during injection operations. It is possible to apply the injection apparatus shown in Fig. 7 with some extra shielding on the injection container and eventually also on the injection tubing (see below). Alternatively, a method such as the one illustrated in Fig.10 is applicable.

Radiation dose may be received directly from the tracer container (external radiation) by spillage on skin and clothes and by oral intake of radioactive liquids. Given that the tracer during the injection process is kept in a closed system, there is, generally, no possibility, of oral intake of the tracer or of skin and clothing contamination under normal conditions.

The D1 and D2 values are different in this case. The proposed values are D1(57Co) = 7 x 101 TBq (about 20 Ci) and D2(57Co) = 4 x 102 TBq (about 1.1 x 103 Ci), respectively [24]. Considering that a typical injection quantity is 3.7-37 GBq (0.1—1.0 Ci), which is only a factor 20-200 lower than the given D1 value, measures should be taken to reduce the dose rate from the injection solution during handling and injection. Radiation dose can most effectively be minimized by passive shielding of the injection container, for instance, by lead. The linear attenuation coefficient for the most intense gamma ray at 122.1 keV in lead is calculated to be ^122keV(Pb) = 36.3 cm4. The half-thickness and the thicknesses needed to reduce the radiation intensity by factors of 100 and 1000, respectively, are x05 = 1.9 x 10-3 cm, x100 = 0.127 cm and x1000 = 0.190 cm.

For a dispersed source, however, where the D2 value applies, the injection quantity is a factor of more than 1000 lower.

Cobalt-58 is produced in charged particle reactions at accelerator facilities (for instance by the reactions 55Mn(a, n) or 57Fe(d, n) or by fast (14 MeV) neutron reactions (for instance 59Co(n,2n) or 58Ni(n, p)) and there is no sizeable production (or natural inventory) in either the biosphere or geosphere. It decays by 85% electron capture and 15% positron emission and has a half-life of 70.86 d. The main photon energies are 511 keV annihilation radiation (29.8%) and 810.76 keV (99.45%).

Being a gamma emitter with intermediate energies, the same general comments as given for 57Co above apply also for 58Co. Owing to somewhat different decay characteristics and higher gamma energies, the D values are lower at D1 = 7 x 10-2 TBq (about 2 Ci) and D2 = 7 x 101 TBq (about 2000 Ci), respectively.

Since a typical injection quantity is 3.7-37 GBq (0.1-1.0 Ci), which is only a factor 2-20 lower than the given D1 value, measures must also be taken to reduce the dose rate from the injection solution during handling and injection. For lead shielding, the linear attenuation coefficient for the most intense gamma ray at 810.76 keV is calculated to be ^810keV(Pb) = 0.94 cm4. The half-thickness and the thicknesses needed to reduce the radiation intensity by factors of 100 and 1000, respectively, are x05 = 0.74 cm, x100 = 4.9 cm and x1000 = 7.3 cm.

For a dispersed source, however, where the D2 value applies, the injection quantity is a factor of more than 1000 lower.

Cobalt-60 is produced in thermal neutron reactions in a nuclear reactor (59Co(nth, y)) or by fast (14 MeV) neutron reactions (e. g. 60Ni(n, p) or 63Cu(n, a)) and there is no sizable production (or natural inventory) in either the biosphere or geosphere. It decays by 100% beta emission and has a half-life of 5.27 y. The main gamma energies are 1173.2 keV (99.85%) and 1332.4 keV (99.98%).

Being a strong and relatively high energy gamma emitter, the same general comments as given for 57Co and 58Co above also apply for 60Co. The decay characteristics are different from the two gamma emitters described above and the D values are even lower than for 58Co: D1 = 3 x 10-2 TBq (about 0.8 Ci) and D2 = 3 x 101 TBq (about 800 Ci), respectively. Since a typical injection quantity is 3.7-37 GBq (0.1—1.0 Ci), which is in about the same region as the given D1 value, measures must also be taken to reduce the dose rate from the injection solution during handling and injection. For lead shielding, the linear attenuation coefficient for the most intense gamma ray at 1332.4 keV is calculated to be ^1332keV(Pb) = 0.72 cm-1. The half-thickness and the thicknesses needed to reduce the radiation intensity by factors of 100 and 1000, respectively, are x0 5 = 0.97 cm, x100 = 6.5 cm and x1000 = 9.7 cm.

For a dispersed source, however, where the D2 value applies, the injection quantity is a factor of about 800 lower.

Iodine-125 is produced in charged particle reactions at accelerator facilities (for instance by the reactions 123Sb(a,2n), 126Te(p,2n) or 127I(p,3n plus beta decay) or by thermal neutron reactions in a nuclear reactor (124Xe(nth, y) plus beta decay) and there is no sizeable production (or natural inventory) in either the biosphere or geosphere. It decays by 100% electron capture and has a half-life of 59.4 d and the main photon energies are the tellurium X rays Ka2 = 27.2 keV (40.1 %), Ka1 = 27.4 keV (74.0%), Kp3 = 30.9 keV (6.83%) and Kp1 = 31.0 keV (13.2%) and the gamma ray at 35.5 keV (6.68%).

Owing to the low photon energies, external radiation from a source of 125I is relatively low and it is easily shielded. The D value for a closed source is D1 = 10 TBq (270 Ci). However, because of the biological effect of iodine (for instance accumulation of I- in the thyroid gland), the D2 value is much lower than for the radionuclides previously discussed: D2 = 0.2 TBq (5.4 Ci).

The external radiation dose is easily shielded by a modest amount of shielding material. For lead shielding, the linear attenuation coefficient for the most intense gamma ray at around 30 keV is calculated to be ^0keV(Pb) = 204 cm-1. The half-thickness and the thicknesses needed to reduce the radiation intensity by factors of 100 and 1000, respectively, are x05 = 3.4 x 10-3 cm, x100 = 2.3 x 10-2 cm and x1000 = 3.4 x 10-2 cm. Dose may also be received by spillage on skin and clothes and by oral intake of radioactive liquids or inhalation of iodine in elemental (I2) form (I — may be easily oxidized in the environment). Therefore, it is especially important to ensure no liquid leakage occurs during the handling and injection processes.

Iodine-131 is mainly produced by thermal fission of 235U or by reactions induced by thermal neutrons in a nuclear reactor (130Te(nth, y) plus beta decay), and there is no sizeable production (or natural inventory) in either the biosphere or geosphere. It decays by 100% beta emission and has a half-life of 8.02 d and the main gamma energies are 284.3 keV (6.22%), 364.5 keV (81.5%) and 637.0 keV (7.16%).

The chemistry and the physiological processes and reactions of 131I — are the same as for 125I-. Owing to the higher gamma energies, the D values are relatively low: D1 = D2 = 0.2 TBq (5.4 Ci). Since normal injected quantitites are in the range 3.7-37 GBq (0.1—1.0 Ci), strict measures must be taken to reduce any risk of excessive doses.

External radiation dose may be reduced by proper shielding. For the most intense gamma ray at 364.5 keV, the linear attenuation coefficient is calculated to be ^364keV(Pb) = 2.8 cm-1. The half-thickness and the thicknesses needed to reduce the radiation intensity by factors of 100 and 1000, respectively, are x05 = 0.25 cm, x100 = 1.6 cm and x1000 = 2.4 cm.