METHOD FOR DETERMINING THE QUANTITY OF A RADIO ISOTOPE

Description

TECHNICAL FIELD

[0001] The present invention relates to a method for determining a quantity of a first radioisotope contained in a source term. In particular, the present invention uses a measurement by gamma spectroscopy for determining the mass of the first radioisotope, the first being in particular plutonium 241. PRIOR ART

[0002] Radioactive waste, and in particular packages of radioactive technological waste, are generally assembled in drums for storage.

[0003] This technological waste comes from the operation of nuclear industry workshops and factories, for example, nuclear fuel manufacturing workshops using nuclear materials such as uranium and / or plutonium, particularly for the manufacture of "MOX" (for "Mixed oxide") fuel. Thus, technological waste from such workshops, for example gloves, sleeves, or any other object made of organic polymer or more precisely polyethylene and used in glove boxes, is likely to contain small quantities of residual plutonium and other radioisotopes.

[0004] Also, depending on the mass and nature (level of radioactivity and half-life) of radioisotopes contained in the drums, the latter are stored either on the surface or at depth. To a certain extent, we can distinguish between drums "poor" and drums "rich" in radioisotopes, which can be stored, respectively, on the surface or at depth.

[0005] Since the cost of deep storage is much higher than that of surface storage, it is therefore necessary to optimize the assembly of waste packages in the drums in order to minimize their radioisotope content, and therefore first of all to know how to differentiate between packages that are "rich" and packages that are "poor" in radioisotopes.

[0006] Neutron counting measurements are then a method of choice for evaluating the mass of radioisotopes, and more particularly residual plutonium, in waste packages.

[0007] However, this type of measure is not satisfactory.

[0008] Indeed, in order to achieve acceptable accuracy, neutron measurements require a relatively long counting time (at least 900 seconds) per package. This affects the management of packages and their assembly into drums, both in terms of cycle time and cost.

[0009] Furthermore, neutron counting measurements are expensive.

[0010] Document US 6791093 discloses a method for determining a quantity of a first radioisotope of a source term contained in a package, said first radioisotope is likely to emit a first gamma line, a second gamma line is also likely to be emitted. The gamma lines are likely to be attenuated by the matrix of the package and the source term itself. A correction is applied to the intensity of the gamma lines emitted by energy.

[0011] WO00 / 42446A1 discloses various correction techniques known in the art, including the technique of calculating ratios with and without attenuation: "differential peak absorption based correction" based on a ratio of gamma emissions of one energy to gamma emissions of another energy in the absence of attenuation. The ratio is compared with the measured result (the actual ratio) to derive a correction factor for attenuation. A transmission source is also used to determine the attenuation effects.

[0012] An aim of the present invention is therefore to propose a method for evaluating the mass of radioisotopes, in particular those contained in a package, making it possible to optimize the design and management of waste drums.

[0013] Another aim of the present invention is to propose a method for evaluating the mass of radioisotopes having a lower cost than the methods known from the state of the art. STATEMENT OF THE INVENTION

[0014] The aims of the invention are, at least in part, achieved by a method for determining a quantity of a first radioisotope of a source term contained in a package, according to claim 1, said first radioisotope is capable of emitting a first gamma line, a second gamma line is also capable of being emitted either by the first radioisotope or by a second radioisotope of the source term, the first gamma line and the second gamma line are capable of being attenuated, each according to a different attenuation rate, by the matrix of the package and the source term itself, the method comprising the following steps: a) a step of measuring, by at least one gamma detector and according to two diametrically opposed positions relative to the package, the net counts associated with the first and second gamma lines emitted; b) knowing the ratio of the counts of the first and second gamma lines in the absence of attenuation and from the results of step a), a step of determining the attenuation rates of the first and second gamma lines by the matrix of the package; c) a step of determining, from the attenuation rate of one or other of the first and second gamma lines determined in step b), the quantity, advantageously the mass, of first radioisotope contained in the source term.

[0015] It is understood that the relative counts of the first and second gamma lines in the absence of attenuation in step b) are theoretical counts.

[0016] According to one embodiment, the second gamma line is emitted by the second radioisotope, and the method also comprises a step a1) of determining the theoretical ratio of the counts between the first and second gamma lines emitted by the first and second radioisotopes in the absence of attenuation.

[0017] Carrying out such a step a1) may require knowledge of the isotopic composition at the measurement date.

[0018] According to one mode of implementation, the ratios of the emissions of the first and second gamma lines are stable over time.

[0019] According to one mode of implementation, the emission lines considered can come from a single radioelement in which case the theoretical ratio of the counts in the absence of a screen will be constant over time.

[0020] According to one method of implementation, the second radioisotope is the result of the radioactive disintegration of the first radioisotope.

[0021] According to one method of implementation, the half-life time T2 of the second radioisotope is much lower than the half-life time T1 of the first radioisotope.

[0022] According to one embodiment, the source term further comprises a third radioisotope, of the same mass number as the first radioisotope, the isotopic composition between the first and second radioisotopes being known.

[0023] According to one embodiment, the method comprises a determination of the isotopic composition.

[0024] According to one embodiment, the method further comprises a step of determining a conversion coefficient (counts / second / gram) of one or other of the first and second radioisotopes for counting one or other of the first and second gamma lines.

[0025] According to one embodiment, the conversion coefficient is also determined from the distance D of the at least one gamma detector relative to the source term.

[0026] According to one embodiment, the attenuation rate of one or other of the first and second gamma lines is representative of an equivalent screen comprising a mixture of polyethylene and uranium characteristic of the matrix of the package, said equivalent screen being characterized by a thickness, called screen thickness, and a polyethylene content.

[0027] According to one embodiment, step a) is carried out with two gamma detectors so that two estimates of the quantity, advantageously the mass, of the first radioisotope of the source term are obtained at the end of step c).

[0028] According to one mode of implementation, as soon as the ratio of the two estimates reflects an offset of the source term in relation to the gamma detectors, the average of said estimates is corrected by a correction coefficient.

[0029] According to one embodiment, the first radioisotope comprises plutonium 241, the second radioisotope uranium 237, and the third radioisotope americium 241.

[0030] The invention also relates to a measuring device suitable for implementing the method for determining a quantity of a first radioisotope contained in a source term according to the present invention, according to claim 14, the device comprising: at least one gamma detector; a docking station for the source term, the docking station being provided with a scale; a calculator intended to process gamma line counts collected by the gamma detectors, said calculator also being provided with a computer program adapted to determine, from the collected gamma ray counts, the quantity of first radioisotope included in the source term. BRIEF DESCRIPTION OF THE DRAWINGS

[0031] Other characteristics and advantages will appear in the following description of the method for determining a quantity of a first radioisotope contained in a source term according to the invention, given as non-limiting examples, with reference to the appended drawings in which: there figure 1is a representation of a measuring device capable of being implemented for the determination of a quantity of a first radioisotope of a source term according to the present invention; figure 2 is a graphical representation of the attenuation rate (vertical axis, in cm -1< ) ​​of a mixture containing polyethylene and uranium as a function of the length of the gamma radiation (horizontal axis, in KeV), in particular, curves A, B and C represent the attenuation of the gamma radiation, respectively, of pure uranium, of a 50 / 50 uranium-polyethylene mixture, and of pure polyethylene; figure 3 is a schematic representation of a source term contained in a package and a gamma detector taken as an example for the determination of a corrective coefficient. DETAILED DESCRIPTION OF SPECIFIC EMBODIMENTS

[0032] The present invention implements a measurement by gamma spectroscopy for the estimation of a quantity of a first radioisotope of a source term contained in a package.

[0033] By "source term" we mean the nature and quantity of radioactive products.

[0034] The estimation of the quantity or mass of the first radioisotope is then based on a measurement of the relative intensities of two gamma lines emitted respectively by the first and second radioisotopes and corrected for an attenuation rate of said lines by the matrix formed by the package.

[0035] The invention is then particularly suitable for estimating the mass of plutonium 241 of a source term contained, in particular, in a package of waste containing polyethylene objects contaminated by a mixture of plutonium and uranium oxides.

[0036] To the figure 1, we can see a device 10 capable of being implemented within the framework of the present invention.

[0037] The measuring device 10 then comprises a docking station 11, possibly provided with a scale 12 intended to measure the mass of a package.

[0038] The measuring device 10 also comprises at least one gamma detector (also called a gamma radiation detector), for example two gamma detectors 13 and 14, arranged to measure gamma radiation emitted by a package containing a source term and positioned on the docking station 12.

[0039] The two gamma detectors 13 and 14 are, for example, positioned on either side of the source term (in other words diametrically opposed to the package).

[0040] A gamma detector 13, 14 can then comprise a plurality of sub-units (for example 4 sub-units) of CZT type detection.

[0041] The measuring device 10 may also be provided with a computer or calculator which includes a computer program.

[0042] The computer program is then intended to process gamma radiation counts collected by the gamma detector(s) so as to determine the quantity of first radioisotope included in the source term according to a method detailed in the remainder of this description.

[0043] The term source may be understood to mean a package or a drum of radioactive waste contaminated by radioactive elements, and in particular contaminated by the first radioisotope.

[0044] Contaminated waste may include gloves, sleeves, or any other waste made, for example, from organic polymer material considered throughout the rest of the statement to be polyethylene.

[0045] The package thus forms a matrix within which the organic polymer material and the radioactive elements are mixed.

[0046] According to a first embodiment of the present invention, the radioactive elements may comprise in particular the first radioisotope and, optionally, a second radioisotope.

[0047] A first gamma line is emitted by the first radioisotope.

[0048] A second line is also emitted either by the first radioisotope or by the second radioisotope whenever the latter is considered.

[0049] Throughout the remainder of the description, it will be considered that the second gamma line is emitted by the second radioisotope. A person skilled in the art, starting from the present description, will be able to adapt the latter to the case of the first and second gamma lines emitted by a single radioisotope, in this case the first radioisotope.

[0050] The counts of the first gamma line and the second gamma line are carried out by detectors 13 and 14.

[0051] The second radioisotope may be the result of the decay of the first radioisotope according to a first conversion rate (the conversion rate is also called the branching factor "Γ"). However, the scope of the invention should not be limited to this aspect and first and second radioisotopes without parentage may be considered.

[0052] Optionally, the radioactive waste may also comprise at least a third radioisotope, in particular a third radioisotope with the same mass number as the first radioisotope. The first and third radioisotopes thus form a mixture whose respective mass proportions, referred to in the remainder of the statement as "isotopic composition", are assumed to be known for the implementation of the method for determining the quantity of the first radioisotope.

[0053] The isotopic composition of the mixture consisting of the first and third radioisotopes can be determined by applying a law of radioactive decay of said radioisotopes and considering the age of the mixture.

[0054] It will be assumed throughout the remainder of the statement that the radioactive elements comprise the first, second and third radioisotopes, and that the isotopic composition between the first and third radioisotopes is known. However, the implementation of the present invention does not necessarily imply the presence of the third isotope.

[0055] For example, the first radioisotope is plutonium 241, which emits a first gamma line at 148 KeV, while the second radioisotope, obtained by α decay of the first radioisotope, is uranium 237, which emits a second gamma line at 208 KeV. In this example, plutonium 241 decays, by emitting β radiation, into americium 241, which forms the third radioisotope.

[0056] The counts of the first and second gamma lines collected by the gamma detector(s) theoretically make it possible to estimate the quantity of the first radioisotope contained in the source term.

[0057] However, the counts are generally attenuated by the package matrix and by more or less significant self-absorption phenomena depending on the mass and geometry of the source term. In other words, the package as a whole (the matrix and the source term) forms a screen against gamma radiation characterized, for each of the gamma radiation wavelengths, by an attenuation rate.

[0058] Also, in a particularly advantageous manner, the attenuation rates of the first and second gamma lines, called respectively, first screening rate and second screening rate, are different.

[0059] In this regard, the figure 2 is a graphical representation of the gamma radiation attenuation rate for a mixture consisting of polyethylene and uranium. In particular, positions 1 and 2 on the figure 2indicate attenuation rates for a given screen, respectively, of the first gamma line and the second gamma line which are different. Thus, knowledge of the densities of polyethylene and uranium, as well as their respective absorption coefficients of the first and second gamma lines, makes it possible to establish the attenuation rates of said lines as a function of the respective proportions of polyethylene and uranium and the screen thickness.

[0060] The method according to the present invention then comprises a step a1) of determining the theoretical ratio of the counts between the first and second gamma lines emitted by the first and second radioisotopes in the absence of screening.

[0061] It is understood that the counts are theoretical counts at the level of the detector(s), and take into account the efficiency of the detector at the energy considered, the number of detection sub-units, but also the isotopic composition therefore: that the third radioisotope is considered and / or that the emission lines taken into account come from two radioelements which are not related.

[0062] In particular, the efficiency and the number of detection subunits make it possible to determine a conversion coefficient c / s / g for one and / or the other of the first and second gamma lines and associated with the first radioisotope.

[0063] By "c / s / g conversion coefficient" we mean a coefficient counting the number of counts of a given gamma line per unit of time and unit of mass of the first radioisotope (therefore of the mass of the source term).

[0064] Therefore, since the corrected count of one or other of the first and second gamma lines is known, the mass of the first radioisotope corresponds to the corrected count of one or other of the first and second gamma lines divided by the conversion coefficient c / s / g of the gamma line considered and the measurement time by the gamma detector.

[0065] The term "corrected gamma line count" refers to the number of counts that would be returned by the detector in the absence of a screen. In other words, the corrected count does not take into account any attenuation by the screen.

[0066] The remainder of the description of the invention will be limited to the case of the second gamma line.

[0067] Knowledge of the relative proportions of the first and second radioisotopes makes it possible to weight the theoretical count per unit of time and per unit of mass, in the absence of screening, of the first and second gamma lines.

[0068] The relative proportions of the first and second radioisotopes are then accessible, for example, by application of a radioactive decay law involving the half-life times T1 and T2, respectively, of the first radioisotope and the second radioisotope.

[0069] Advantageously, since there is a relationship between the first and second radioisotopes and the radioactive period of the son is much lower than that of the father, their activities are similar in which case the theoretical ratio of the counts is constant over time.

[0070] The method according to the present invention also comprises a step a) of measuring by the at least one gamma detector, the net counts associated with the first and second gamma lines emitted, respectively, by the first and second radioisotopes contained in the source term. The measurement is advantageously carried out according to two diametrically opposite positions relative to the package.

[0071] It is understood that the net counts associated with the first and second gamma lines correspond to the actual attenuated counts.

[0072] Advantageously, step a) uses the two gamma detectors 13 and 14.

[0073] The measurement may include collecting the gamma radiation(s) for a predetermined collection time of, for example, between 45 seconds and 60 seconds, including 45 seconds.

[0074] The net count can be obtained from a gross count deduced from the background noise, actually measured by the gamma detector, adjusted for a dead time during which said gamma detector does not collect gamma radiation at its maximum capacity.

[0075] Step (a) may also include calculating the ratio of the net counts of the second and first gamma lines.

[0076] Step a) is then followed by step b) of determining an attenuation rate of the first and second gamma lines by the source term matrix.

[0077] In this regard, the net counting ratio, measured in step a), makes it possible to determine the attenuation rate of the screen formed by the matrix.

[0078] This operation, which can be automated at the computer program level, takes the parameters allowing the graph of the figure 2 .

[0079] Said parameters then include the respective densities of polyethylene, uranium, absorption coefficients of uranium and polyethylene at the respective energies of the first and second gamma lines.

[0080] The computer program may include a mathematical solver which, by an iterative method and from the ratio of net counts of the first and second gamma lines, makes it possible to determine the polyethylene content (in percentage) of the screen, the screen thickness and the attenuation rate for each of the two gamma lines considered.

[0081] This mathematical resolution step known to those skilled in the art is not described in the present invention.

[0082] Furthermore, the ratio of the net count of the second gamma line to the attenuation rate thus determined makes it possible to access the corrected count of said second gamma line previously defined.

[0083] Finally, the method according to the present invention comprises a step c) of determining the mass of the first radioisotope included in the source term.

[0084] This step c) may then initially include the calculation of the corrected count associated with the second gamma line. For example, the corrected count (of the attenuation by the equivalent screen) of the second gamma line corresponds to the net count of said second line by the detector divided by the attenuation exerted by the screen (divided by the attenuation rate at the energy considered).

[0085] Thus, the mass of the first radioisotope contained in the source term is then determined by dividing the corrected count associated with the second gamma line by the product formed by the measurement time and the conversion coefficient c / s / g.

[0086] In the case of a large package, the main "hot" point may be off-center relative to said source term, and generate an error in the estimation of the counting of the first and second gamma lines.

[0087] By "main hot spot" we mean the area of ​​the most radiative source term.

[0088] Also, since the method uses two gamma detectors 13 and 14, two estimates of the counting of the second gamma line are obtained. A deviation from 1 of the ratio of these two estimates reflects an eccentricity of the main hot spot.

[0089] To overcome this problem, the solver integrates a correction table of the average value of the estimated mass based on the ratios of the estimates of the mass of the first radioisotope for each of the detectors.

[0090] In this respect, the following table presents a method for determining the correction coefficient. In this example, the source term, here considered as punctual, is assumed to emit 100 counts, and two gamma detectors are arranged symmetrically on either side of the source term 15 and in alignment along an axis XX' ( figure 3 ). Average Report C coefficient min max d0° d180° Offset cm 100 1,03 1,000 99 101 151 149 1 100 1,05 1,001 97 103 152 148 2 100 1,08 1,001 96 104 153 147 3 100 1,11 1,002 95 106 154 146 4 100 1,14 1,003 94 107 155 145 5 100 1,17 1,005 92 109 156 144 6 101 1,21 1,007 91 110 157 143 7 101 1,24 1,009 90 112 158 142 8 101 1,27 1,011 89 113 159 141 9 101 1,31 1,013 88 115 160 140 10 102 1,34 1,016 87 116 161 139 11 102 1,38 1,019 86 118 162 138 12 102 1,42 1,023 85 120 163 137 13 103 1,45 1,027 84 122 164 136 14 103 1,49 1,031 83 123 165 135 15 103 1,53 1,035 82 125 166 134 16 104 1,58 1,039 81 127 167 133 17 104 1,62 1,044 80 129 168 132 18 105 1,66 1,049 79 131 169 131 19

[0091] The “cm offset” column corresponds to offsets of the main hot spot relative to the center of source term 15.

[0092] The columns “d0°” and “d180°” correspond to the distances of the main hot spot from the gamma detectors.

[0093] The "min" and "max" columns correspond, respectively, to the actual counts of each of the gamma detectors. The "Average" and "C Ratio" columns are, respectively, the averages and ratios of the "min" and "max" columns.

[0094] The “coef” column is the ratio of the data from the “Average” column and the corrected count (100 counts in this case), and corresponds to the corrective coefficient to be applied when the ratio of the estimates from the two gamma detectors reflects an offset of the main hot spot.

[0095] The estimation of the eccentricity can also use other models, for example by considering a linear source term, which the person skilled in the art, with his general knowledge, is able to implement.

[0096] The present invention then makes it possible to measure, in a reasonable time, the mass of the first radioisotope contained in a source term.

[0097] This measurement time then limits the exposure of people to radiation, but also makes it possible to shorten the cycle times necessary for managing and assembling source terms in waste drums.

[0098] The present invention also relates to the measuring device 10 adapted for implementation for determining a quantity of a first radioisotope contained in a source term, the device comprising: at least two gamma detectors 13 and 14; a docking station 11 of the source term, the docking station being provided with a scale 12; a calculator 15 intended to process gamma ray counts collected by the gamma detectors, said calculator also being provided with a computer program adapted to determine, from the collected gamma ray counts, the quantity of first radioisotope included in the source term.

Claims

1. Method for determining a quantity of a first radioisotope of a source term contained in a package, said first radioisotope is suitable for emitting a first gamma line, a second gamma line is also suitable for being emitted either by the first radioisotope or by a second radioisotope of the source term, the first gamma line and the second gamma line are suitable for being attenuated, each according to a different attenuation rate, by the matrix of the package and the source term per se, the method comprising the following steps: a) a step of measuring, by at least one gamma detector (13, 14) and according to two diametrically opposed positions in relation to the package, net counts associated with the first and second gamma lines emitted; b) in the knowledge of the ratio of the counts of the first and second gamma lines in the absence of attenuation and based on the results of step a), a step of determining the attenuation rates of the first and second gamma lines by the package matrix; c) a step of determining, based on the attenuation rate of one or the other of the first and second gamma lines determined in step b), the quantity, advantageously the mass, of first radioisotope contained in the source term.

2. Method according to claim 1, wherein the second gamma line is emitted by the second radioisotope, and the method also comprising a step a1) of determining the theoretical count ratio between the first and the second gamma lines emitted by the first and second radioisotopes in the absence of attenuation.

3. Method according to claim 2, wherein the emission ratios of the first and second gamma lines are stable over time.

4. Method according to claim 2 or 3, wherein the second radioisotope is the result of the radioactive disintegration of the first radioisotope.

5. Method according to one of claims 2 to 4, wherein the half-life T2 of the second radioisotope is substantially less than the half-life T1 of the first radioisotope.

6. Method according to one of claims 2 to 5, wherein the source term further comprises a third radioisotope, of the same mass number as the first radioisotope, the isotopic composition between the first and second radioisotopes being known.

7. Method according to claim 6, wherein the method comprises a determination of the isotopic composition.

8. Method according to claim 5 or 7, wherein the method further comprises a step of determining of a conversion coefficient (count / second / gram) of one or the other of the first and second radioisotopes for counting one or the other of the first and second gamma lines.

9. Method according to claim 8, wherein the conversion coefficient is also determined based on the distance D of the least one gamma detector in relation to the source term.

10. Method according to one of claims 1 to 9, wherein the attenuation rate of one or the other of first and second gamma lines is representative of an equivalent screen comprising a mixture of polyethylene and uranium characteristics of the package matrix, said equivalent screen being characterised by a thickness, referred to as screen thickness, and a polyethylene content.

11. Method according to one of steps 1 to 10, wherein step a) is carried out with two gamma detectors such that two estimations of the quantity, advantageously the mass, of first radioisotope of the source term are obtained following step c).

12. Method according to claim 11, wherein, once the radio of the two estimations conveys an offset of the source term in relation to the gamma detectors, the mean of said estimations is corrected by a correction coefficient.

13. Method according to one of claims 1 to 12 combined with claim 6, the first radioisotope comprises plutonium 241, the second radioisotope uranium 237, and the third radioisotope americium 241.

14. Measurement device (10) suitable for implementing the method for determining a quantity of a first radioisotope contained in a source term according to one of claims 1 to 13, the device comprising: - at least one gamma detector (13, 14); - a docking station (11) of the source term, the docking station being provided with a scales (12); - a computer intended to process gamma line counts collected by the gamma detectors, said computer being also provided with a computer program suitable for determining, based on the gamma ray counts, the quantity of first radioisotope comprised in the source term.