Category Archives: IAEA RADIATION TECHNOLOGY SERIES No
Measuring techniques depend on the tracers’ radiological characteristics (beta, gamma or stable chemical ones), but all of them include some sample treatment prior to undertaking the measurement itself.
When counting a radioactive sample, it is well known that the instrument reading is a measure of the sample activity plus the background activity. The latter must be subtracted in order to evaluate the actual net sample activity. The background activity is usually taken to be the activity measured by using the sample taken before the injection (blank sample). If, however, the tracer does not appear immediately (there are no canalizations), a more representative value of the background activity is obtained by measuring several of the samples and averaging the results, taking into account that the more samples, the lower the background’s variation coefficient will be. The variation coefficient is the ratio between the standard deviation and the mean value.
Radioactive decay is an inherently random phenomenon that follows, strictly speaking, the binomial distribution. Nevertheless, the Poisson distribution is an excellent approach that takes into account some of the radioactive decay characteristics (the random event ‘disintegration’ is repeated many times and the individual probability of an atom disintegrating is very low).
Poisson distribution depends on just one parameter, generally symbolized by the Greek letter a and the distribution mean value and the variance are both equal to a This property is very useful in radioactive measurements. Once the count rate has been determined, its numerical value can be used as the expected average value and its square root as the standard deviation.
Furthermore, in the case that the number of events approaches infinity, the binomial distribution and the Poisson distribution converge towards anothers statistical distribution known as the normal distribution or Gaussian distribution, which is continuous and symmetrical around its mean value. In a normal distribution, the probability for the random variable to take values close to the mean value is very high while it approaches, asymptotically, zero for large values located in the positive and negative distribution ‘tails’.
As a ‘rule of thumb’ it is common to require that all random variable values fall within a 2.0 standard deviation interval around the mean value. In such a case, a confidence level of 95% for the measurement is established. This means that there is a 5% probability that the ‘true’ mean value is outside the range given by the measured mean value by +2.0 and -2.0 standard deviations. Consequently, two measurements may be said to belong to different populations when their measured mean values differ by at least five standard deviations. This criterion is also applied to determine whether a sample is active or not, namely, that a sample has some radioactivity of its own when its count rate is five standard deviations greater that the background.
Generally, the following condition is established to calculate the lower detection limit (minimal detectable concentration), LD, on the basis of the instrument background:
RN > 2 SRB (5)
This means that the sample count rate should be at least twice its own standard deviation in order to be distinguished from the background. The standard deviation is given by the following expression:
<rRN is the standard deviation for the net count rate RN (cps);
Rg is the gross count rate (cps);
tc is the counting time (s);
RB is the background count rate (cps).
After some operations and approximations the following expression is obtained for LD:
Ld is the lower detection limit (or minimum detectable activity concentration)
RB is the background count rate (cps);
tc is the counting time;
є is the detection efficiency (counts per disintegration);
V is the sample volume.
As a consequence of statistical dispersion, some preprocessing of experimental data is usually needed in addition to the subtraction of background values in order to filter noise and smooth the response curves.
Finally, radioactive decay correction is needed, although in the case of tritium its half-life is long enough to avoid this kind of correction when the sampling periods last only a few months. In a general situation, an interwell study implies more than a year of sampling and tritium decays at a rate of 0.45% per month.
The Tongonan-1 sector of the Leyte geothermal production field (LGPF) in the Philippines (Fig. 67) has been experiencing declines in output in some of its production wells, which have been mainly attributed to injection returns from brine injected into one of the wells situated near the production area. Routine production well chemistry monitoring indicated physicochemical changes in the production wells attributed to migration of the reinjected brine back to the production sector since 2001, when the injection load from South Sambaloran was transferred to Tongonan-1.
TABLE 16. CONCENTRATION OF 125I IN PRODUCER WELLS (net cpm/g)
ЮООО 8000 ■
6000 4000 — 2000 — n —
FIG. 66. Tracer experimental response curve obtained at the production well HCE-28A.
FIG. 67. Location map of the LGPF showing tracks of wells drilled (in red lines). Inset delineates the area where the tracer test was conducted. Well 1R8D is the tritium injector well.
Each sample arriving in the laboratory should be properly registered. A suitable sample registration form (consisting of sample identification code, description of sampling well/station with its location, name of project, date of sampling, date of sample receipt in laboratory and name of receiving person with their signature) should be used to maintain the sample record.
Check that bottles are not leaking. Store the bottles in proper conditions away from heat sources and direct sunlight. Samples should not be stored in rooms/buildings where artificial tritium compounds are, or have been, handled (contamination risk).
Tracer technology plays an important role in oilfield development and operation. Interwell tracer testing is an important reservoir engineering tool for the secondary and tertiary recovery of oil. Most of the oilfields in many developing countries are in the stage of secondary recovery. Moreover, the oil industry remains a priority in these countries. Interwell tracer testing is also used in geothermal reservoirs to gain better understanding of reservoir geology and to optimize production and re-injection programmes. Today, the use of tracers for interwell communication studies is an established technique.
The IAEA facilitates the transfer of technology, and an important part of this process is the provision of relevant literature that may be used for reference purposes or as an aid to teaching. This publication aims to provide not only an extensive description of what can be achieved by the application of radiotracer techniques in interwell investigations in onshore and offshore fields, but also sound and experience based guidance on all aspects of the design and implementation of experiments and the interpretation of results. It describes the principles and the state of the art of radiotracer techniques for interwell investigations.
The publication contains guidance on the technical steps of interwell tracer testing, as well as input from participants of the coordinated research project (CRP) on Validation of Tracers and Software for Interwell Investigations. The major achievements of the CRP and novel developments in tracer methodologies and technologies as applied to interwell investigations are also included. The unedited reports of the CRP participants, as presented at the final research coordination meeting, and software for interwell data interpretation are included as support materials on the accompanying CD-ROM. The publication has been prepared with contributions from all CRP participants. The IAEA gratefully acknowledges all contributors to this publication, especially T. Bjornstad for compiling and reviewing it.
The IAEA officer responsible for this publication was Joon-Ha Jin of the Division of Physical and Chemical Sciences.
The concentration versus time curves (experimental response curves) are analysed to measure the main characteristics of flow in the group of wells under study (Fig. 30).
For this, 370 GBq of HTO was injected into well CnE-241. The experimental response curves give transit times and allow quantification of the injected water produced in different directions. Basic interpretation or first level interpretation allows to acquisition of qualitative and semi-quantitative information that in many cases is the only type required by the end users. Tracer sampling and measurement showed that tracer came up at three out of five production wells around the injection well. The arrows represent the fraction of the injected tracer recovered in each well.
Figure 31 presents the experimental response and cumulative curves obtained at production well CnE-324.
The experimental response curve obtained in production well CnE-324 indicates:
FIG. 32. Extrapolated response curve for well CnE-324.
It is easy to observe that tracer sampling was interrupted before the concentration reaches background level; consequently, for more accurate results, extrapolation of the curve is required.
For this purpose, it is recommended that an exponential function be used and parameters obtained from a least squares fit using the last points in the tail of the experimental data as reference. However, extrapolation has to be used carefully and based upon the general knowledge of each particular situation in order to avoid speculation and unexpected results. Figure 32 shows the extrapolated curve generated by the Anduril software.
The new values for the parameters are:
• Breakthrough time is 117 d.
• Time for the peak is 159 d.
• Mean residence time is 217 d.
• Tracer recovery is 7.7%.
A similar treatment has been applied to other wells in the pattern shown in Fig. 30 with the following experimental results.
Table 21 shows the summary of reported results.
TABLE 20. INJECTION WATER (PROCESSED BRINE) COMPOSITIONS
TABLE 21. REPORTED RESULTS
IV1.2.1. Case study 1
Figure 96 shows a well pattern of a reservoir sited in southern Argentina, where an interwell study by means of radiotracers was performed some years ago.
Since HTO was the selected tracer, the liquid scintillation technique was used for measurement. Because of operative limitations in the laboratory measurement, samples were not distilled before counting and, in addition, a short count time was used. For that reason the detection limit was much higher than usual. The detection parameters were:
• Background = 20 cpm;
• Efficiency = 0.28 (counts/disintegration);
• Count time = 10 min;
• Volume of the sample = 8 mL;
• Detection limit = 29.5 Bq/L.
From the detection limit, the mean output concentration was fixed at ten times this value (295 Bq/L), which leads to an activity of 167 GBq (4.5 Ci). In fact, 10 Ci of HTO was injected into well K-22 using the bypass device. Figure 97 shows an example of the tracer concentration and cumulative response curves for well K-329, whose output was followed during a full year and belongs to the K-22 pattern.
The first information obtained in production well K-329 from a quick analysis of the response curve using Anduril software was:
FIG. 97. Instantaneous (left hand scale) and cumulative (right hand scale) response curves (well K-329).
• Breakthrough = 86 d;
• Mean residence time = 193 d;
• Final time = 321 d;
• Tracer recovery = 9.2%.
The distance between wells K-22 and K-329 is 251 m, thus, the calculated minimum, medium and maximum water velocities are 0.78 m/d, 1.3 m/d and 2.9 m/d, respectively. Permeability was also evaluated by Anduril software using Darcy’s law. A value of 282 mD was obtained, which appeared reasonable to reservoir engineers.
Anduril software was used to model the experimental response curve obtained by sampling in production well K-329. A radial dispersion model was applied. Figure 98 indicates a good fit.
The model gives the following parameter values for the dynamics of tracer movement between injection well K-22 and production well K-329:
• Breakthrough = 75 d;
• Mean residence time = 210 d;
• Final time = 410 d;
• Tracer recovery = 9.1%.
Cumulative injected volume (m5)
The volumetric response curve for well K-329 appears in Fig. 99, given in terms of the cumulative injected volume. Reservoir information extracted from the tracer response curve in well K-329 is:
• Breakthrough = 86 d;
• Mean volume = 19.073 m3;
• Swept volume = 1.775 m3.
The last value is the pore volume swept from the injector to the production well (K-329) and equals the mean volume multiplied by the recuperation factor (0.092).
IV 1.2.2. Case study 2
Complex response curves can be obtained in some cases due to the mixed response from different layers. Anduril software can be used to decompose a complex response curve into several simple curves. These simple curves are supposed to represent tracer movement in a unique layer. Modelling each of the simple curves provides the mean residence times and the quantity of tracer recovered from each layer.
An example of this methodology is shown in Fig. 100 in which the complex response of a production well was approached using four simple functions (Fig. 101) based on the radial model. A possible explanation of the tracer behaviour could be that it reached the production well by following four paths of different permeability belonging to a unique layer.
The parameters of each function are given in Table 26.
TABLE 26. COMPLEX RESPONSE MODELLED BY FOUR FUNCTIONS
Some 46% of the total quantity of tracer recovered in this well related to the injected activity. The parameter f is the contribution of each path expressed as a fraction of that percentage and was evaluated from the area under each curve. Dispersivities may be calculated by multiplying the ratio D1lvx by the distance between the injection and the production wells.
The sampling frequency and procedure should be well planned. An inadequately planned and prepared sampling programme may ruin the whole tracer project. Important aspects to consider are the following:
• Is each individual production well accessible for sampling? In most land based reservoirs this is the case. In offshore wells where the wellhead is ‘dry’, i. e. placed on a platform above sea level, this may also be the case. However, for subsea completion where the well heads are placed on the sea bed, each individual well may not normally and easily be accessible for frequent sampling. In this situation, the well flows from several wells come together and are co-mingled in one transportation pipeline from the bottom hub to the receiver installation either on an offshore platform or onshore. Individual well sampling may not be possible in this case. However, in order to enable production testing of individual wells, the bottom hub is normally equipped with valves on the production line from each individual well. Thus, by closing these valves according to a certain procedure, the production from one well may be increased at the expense of the others. In this way, the production of a certain tracer may be associated with a specific well (or a limited selection of wells) each time a production logging operation is carried out.
• Cross-contamination: When each individual well is accessible for sampling and the sampling is performed directly in the flow line, there is a very limited possibility for cross-contamination of the collected samples, i. e. that waters from various wells mix in the same sample. However, human error and erroneous labelling of the collected sample may happen and lead to confusion in the results. Most often, the collection of water samples, even from individual wells on a platform, is performed on a test separator which is common for several wells. In this case, cross-contamination can more easily happen. The remedy is to ensure that all water from testing of the previous well is swept out before sampling the next well.
• Sampling procedure: Regarding concrete sampling procedures some questions must be considered, Will it be discontinuous sampling involving personnel for each individual operation? In which case, who is going to do the job? Are there special requirements for sampling containers or stabilization additives to the sampled fluids (e. g. for preventing microbial degradation of the tracer during transportation and sampling)? Is it possible to adapt some form of automatic batch sampling so that personnel are not so involved at each operation? These are all questions that must be clarified upfront.
• Sampling frequency: How often should samples be collected? This question should be answered on the basis of a best estimate of expected tracer breakthrough. As a general rule, the sampling frequency should be relatively high in the beginning, starting shortly (a few days) after injection. All samples are stored safely. As an example, each fifth sample is sent for tracer analysis. After the first tracer detection, the previous four samples are analysed with priority reversed in order to determine more precisely the tracer breakthrough time. After tracer breakthrough, the high sampling frequency should be maintained and each sample analysed until the tracer production peak lies on a decaying slope. On the tail of the curve, the sampling can be less frequent.
The sampling frequency (samples/d) depends on the geometrical size of the tested field section and on the flow rates involved. In an average sized reservoir with well distances in the range of 500-1000 m, a sampling frequency in the beginning can be one sample each 2-4 d, followed by the same frequency after tracer breakthrough. After having passed the peak maximum, the frequency can be reduced to one sample per week or per two weeks and later even to one sample per month. It is, however, recommended that sampling and analysis be continued as long as possible, since much information from a tracer test lies in the tail of the curve. The LD value for the particular tracer sets a practical limit to the sampling period.
3.2.1. !. Solvent extraction method
The procedure was as follows: A series of 5 mL samples of seawater were doped with tracer quantities of 60Co labelled [Co(CN)6]3- and pH adjusted in the range pH1-7 by addition of HCl. Each sample was then contacted in a separator funnel with 5 mL of organic phase (kerosene) containing the extraction agent. The funnel was then vigorously shaken by a mechanical shaking machine for a predetermined time of 3 min at room temperature, which is sufficient to reach extraction equilibrium. After phase separation (by gravity), 1 mL samples were extracted from both phases and measured for radioactivity.
For the high gamma energies of 60Co (1173 keV and 1332 keV), correction for differences in counting efficiency due to sample density is not needed. In addition, when equal volumes of the two phases are used in the counting procedures, the distribution ratio (D value) can be determined from the raw counting rates by the formula:
D = -^ (21)
where Rorg is the net counting rate of the organic sample and Raq is the net counting rate of the aqueous sample.
The quantity extracted from one extraction operation of aqueous and organic phases is expressed as E(%) by the formula:
%E = —D—— 100% (22)
+ D V
where Vaq and Vorg are the volumes of the aqueous and organic solutions, respectively.
For high values of D, it is possible to obtain a substantial volume reduction and enrichment factor by using a volume ratio VqVog = 10 or higher. For equal volumes of the organic and aqueous phases, the formula simplifies to:
% E = -^~ -100% (23)
1 + D
For even further enrichment, it is necessary to strip the activity back to an aqueous phase with a high stripping efficiency and perform a new extraction sequence on this strip solution with a volume ratio Vaq/Vorg >> 1. As an example, for D = 10 and Vaq/Vorg = 10, it is possible to obtain a volume reduction of 10 and an enrichment factor of 5 after one extraction. After a stripping with 100% efficiency and a new extraction with the same volume ratio, a volume reduction of 100 and an enrichment factor of 25 will be obtained. For higher D values, the enrichment factor will be higher.
Therefore, the stripping efficiency was also screened with a few stripping agents.
The extractions performed here are based on so-called ion pair formation and the extraction agents are all amines of different kind, which, as a rule, have been conditioned with 0.1M H2SO4. Data for extraction and stripping systems are found in Table 7.
TABLE 7. EXTRACTION AND STRIPPING SYSTEMS
a Without pretreatment with 0.1M H2SO4.
(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  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 . 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.