Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C17/00—Monitoring; Testing ; Maintaining
  • G21C17/06—Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination
  • G21C17/063—Burn-up control
• • G—PHYSICS
  • G01—MEASURING; TESTING
  • G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
  • G01T3/00—Measuring neutron radiation
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21D—NUCLEAR POWER PLANT
  • G21D1/00—Details of nuclear power plant
  • G21D1/003—Nuclear facilities decommissioning arrangements

Definitions

• the present invention relates to a system, a device and a method for determining a quantity of fissile material in a facility, and more particularly in a facility to dismantle.
• the present invention applies to the non-destructive determination of the mass of fissile material in an installation or equipment, in an environment which may be very irradiating, more particularly before dismantling. This determination is based on the measurement of the in situ neutron flux which is of low amplitude under such conditions.
• the neutron flux is generally emitted by alpha a emitters such as Plutonium but also minor actinides such as Curium and Americium; the contribution of each of them must necessarily be taken into account when interpreting the measurement.
• the detectors implemented are most often cylindrical counters of the proportional type filled with Helium-3 under 4 bar which detect neutrons previously thermalized by a block of high density polyethylene HDPE.
• These detectors which are very sensitive to gamma radiation, must necessarily be protected by shielding, for example of lead, if the radiological environment exceeds 0.01 Gy/h. If the rate of gamma radiation becomes too high, it is possible to use boron deposit counters which can operate correctly up to about 1 Gy/h but at the expense of a significantly lower detection efficiency. For even higher dose rate values, uranium-235 fission chambers can be implemented, but in this case the thermal neutron detection efficiency is very low.
• Each detector must be wired to high voltage electronic amplification equipment, located remotely. Furthermore, the detection sensitivities (in general, of the order of a few tens of g of Plutonium) are very dependent on the environmental conditions, in particular the background noise linked to gamma irradiation, which can vary significantly from one checkpoint to another.
• the proportional neutron counters used in dismantling operations have relatively low detection sensitivity, sensitivity to gamma radiation and performance variability depending on the environment of the measurement point.
• these devices are expensive and complex to implement, especially in very irradiating environments.
• the object of the present invention is therefore to propose a system and a method for determining a quantity of fissile material in an installation remedying the aforementioned drawbacks, in particular by allowing an effective and low-cost determination of the masses of fissile material. in very irradiating areas.
• the present invention relates to a system for determining a quantity of fissile material in a facility, comprising: - a neutron activation device adapted to be placed in a zone of the facility for a predetermined irradiation time, the device for neutron activation comprising a matrix of neutron thermalizing material, preferably high density polyethylene, incorporating a first neutron activation target in a section between faces front and rear of said die and along a first predetermined distance from the rear face of the die,
• a gamma spectrometry device adapted to measure the activity of said first target outside said zone
• a computer device configured to calculate a neutron flux emitted by the fissile material from the measurement of the activity, and to determine the quantity of fissile material as a function of said neutron flux by also using predetermined data relating to the isotopic composition of said fissile material.
• the neutron thermalizer material is typically high density polyethylene (HDPE) but can also be another hydrogenated material such as paraffin, preference being given to HDPE.
• HDPE high density polyethylene
• This system allows efficient and low-cost determination of the masses of fissile material in very irradiating zones where techniques based on proportional counters are much more complex or even impossible to implement. Additionally, targets can be reused after a limited decay phase. In addition, a simultaneous implementation of several activation devices makes it possible to reduce the level of uncertainty during a check.
• the activation device comprises a second neutron activation target in the thermal range, said second target being integrated into said matrix in a plane parallel to that of the first target and according to a second predetermined distance from said first target, said second predetermined distance between the first and second targets being greater than the first predetermined distance.
• the ratio between the first and second predetermined distances is of the order of 1 to 4.
• each of said first and second targets consists of a material having a cross section of the isotope constituting the target in the thermal range, a radioactive period of the corresponding daughter isotope and an isotopic concentration of the isotope constituting the target .
• each of said first and second targets consists of a material selected from the following materials: Tantalum, Silver, Terbium, Cobalt, Copper, and Hafnium, and preferably Tantalum.
• the measurement of the activity of said first target and/or said second target is carried out by high resolution gamma spectrometry.
• the first and second targets are made of an identical material.
• the computing device is configured to determine the location of the fissile material according to the ratio of the activities measured on the first and second targets.
• the activation device comprises a third neutron activation target in the fast domain, said third target being integrated into said matrix in a plane parallel to those of the first and second targets.
• said second and third targets are integrated in juxtaposition on the front face of said matrix.
• the computing device is configured to determine the chemical form of the fissile material according to the ratio of the activities measured on the third and second targets.
• said third target consists of a material having a cross section of the isotope constituting the target in the fast range, a radioactive period of the corresponding daughter isotope and an isotopic concentration of the isotope constituting the target.
• said third target consists of a material selected from nickel and zinc and preferably nickel.
• the gamma spectrometry device is configured to measure the activity from each of the faces of the targets.
• the invention also relates to a neutron activation device, comprising a matrix made of a neutron thermalizing material, preferably high-density polyethylene, and a first neutron activation target in the thermal range consisting of a material selected from the materials following: Tantalum, Silver, Terbium, Cobalt, Copper, and Hafnium, and preferably Tantalum, integrated in a section between front and rear faces of said matrix and according to a first predetermined distance from the rear face of the matrix.
• a neutron activation device comprising a matrix made of a neutron thermalizing material, preferably high-density polyethylene, and a first neutron activation target in the thermal range consisting of a material selected from the materials following: Tantalum, Silver, Terbium, Cobalt, Copper, and Hafnium, and preferably Tantalum, integrated in a section between front and rear faces of said matrix and according to a first predetermined distance from the rear face of the matrix.
• the neutron activation device comprises a second neutron activation target in the thermal range consisting of a material selected from the following materials: Tantalum, Silver, Terbium, Cobalt, Copper, and Hafnium, and preferably Tantalum, integrated in a plane parallel to that of the first target and according to a second predetermined distance from said first target, said second predetermined distance between the first and second targets being greater than the first predetermined distance.
• the neutron activation device comprises a third neutron activation target in the fast range consisting of a material selected from Nickel and Zinc and preferably Nickel, said third target being integrated into said matrix in a parallel plane to those of the first and second targets.
• the invention also relates to a method for determining a quantity of fissile material in an installation, comprising the following successive steps: - positioning at least one neutron activation device in a zone of said installation for a predetermined irradiation period, the at least one neutron activation device comprising a matrix made of a neutron thermalizing material, preferably high density polyethylene, integrating a first neutron activation target in a section between front and rear faces of said matrix and along a first predetermined distance from the rear face of the die,
• FIG. 1 schematically illustrates a system for determining a quantity of fissile material in an installation, according to one embodiment of the invention
• Fig. 2 schematically illustrates a method for determining a quantity of fissile material in an installation in relation to FIG. 1, according to one embodiment of the invention
• Fig. 3 illustrates the gamma spectrum of a Cobalt target irradiated over a period of approximately two months by an actinide source, according to the invention
• Figs. 4A-4C schematically illustrate a neutron activation device, according to different embodiments of the invention.
• Fig. 5 illustrates an experimental neutron activation system suitable for calibrating the response of the device to the effect of the neutron source, according to one embodiment of the invention
• Figs. 6A and 6B illustrate the gamma spectra of the first and third targets relating to Tantalum and Nickel respectively, according to the invention.
• Figs. 7A and 7B illustrate the level of activation according to the depth in the first, third and second targets in mm for a source placed at the front and at the rear of the detection device respectively, according to the invention.
• the principle of the present invention is to position a neutron activation target in an area to be characterized in order to measure the neutron flux in situ, in an environment which may be very irradiating, for a predetermined irradiation period, and then to measure the activity of the target outside the zone to be characterized to deduce the quantity of fissile material in this zone.
• Fig. 1 schematically illustrates a system for determining a quantity of fissile material in an installation, according to one embodiment of the invention.
• This system 1 comprises at least one neutron activation device 3, a gamma spectrometry device 5 and a computing device 7.
• the neutron activation device 3 comprises a matrix made of a neutron thermalizer material, for example high-density polyethylene 9, called HDPE matrix, incorporating at least one neutron activation target 11.
• neutron activation target is meant a target 11 composed of a chemical element whose nucleus captures a neutron n and emits gamma radiation specific to this element to evacuate the excess energy.
• the spectrum of the gamma radiation emitted by the target 11 is thus representative of the neutron flux emitted by the source 12 of fissile material.
• the neutron activation target 11 is of the thermal type, that is to say, suitable for capturing thermal neutrons.
• This target is composed of a material chosen from the following materials: Tantalum Ta, Silver Ag, Terbium Tb, Cobalt Co, Copper Cu, and Hafnium Hf.
• the gamma spectrometry device 5 is advantageously a high-resolution gamma spectrometer comprising a detector 13 (for example, a Germanium HPGe detector) and a multichannel analyzer 15 suitable for analyzing the activity of the target 11 according to the recorded spectrum.
• the computing device 7 is for example a computer comprising in the usual way a processor 17, memories 19 and input 21 and output 23 units.
• the computing device 7 is coupled to the gamma spectrometry device 5 and comprises at least one software 25 configured to interpret and analyze the spectra from the multichannel analyzer 15.
• Fig. 2 schematically illustrates a method for determining a quantity of fissile material in an installation in relation to FIG. 1, according to one embodiment of the invention.
• step El at least one neutron activation device 3 is positioned in an area to be characterized 31 to integrate the neutron flux in situ for a predetermined irradiation time.
• This duration is of the order of a few days to a few months depending on the sensitivity sought.
• the activation device(s) 3 is (are) extracted from the zone to be characterized 31 in order to then measure, offline, the activity of each target 11 by gamma spectrometry.
• step E3 the activity A of each target 11 is measured outside the zone to be characterized by the gamma spectrometry device 5.
• This measurement is represented by a spectrum S comprising a set of lines R1, . .,Rn defining the number of counts according to the energy levels in keV.
• FIG. 3 illustrates the gamma S spectrum of a Cobalt target 11 irradiated over a period of about two months by a source 12 of AmBe actinide.
• the two lines RI, R2 of Cobalt-60 created by reaction (n, y) on Cobalt-59 appear clearly at energies of 1173 keV and 1332 keV.
• the amplitude of these last lines R3, R4 is much weaker because the cross sections of the reaction (n, p) in the fast domain are significantly weaker than those of the reactions (n, y) operating in the thermal domain.
• the activity A can be correlated with the neutron flux F having struck the target 11 and indirectly with the quantity of actinides according to the isotopic composition.
• the computing device 7 is configured to use the measurement of the activity A (expressed in Bq) in order to calculate a neutron flux F (expressed in neutrons per second n/s) emitted by the source of fissile material.
• the neutron flux F is defined according to the activity A of the target 11, the number / of isotopes of the target 11 created by neutron reaction (n, y) for a neutron emitted by the source, the period or half-life T of the daughter isotope of target 11, of the duration T1 of activation of target 11 (typically from 1 to 6 months) and of the waiting duration T 2 between the end of the period activation and the spectrometric measurement, according to the following formula:
• the number / of isotopes created by neutron reaction (n, 7) can be calculated by a three-dimensional simulation of the in situ activation configuration of the target 11.
• the computing device 7 uses, for example, a calculation model based on a known Monte Carlo technique for N-Particles (known as MCNP code).
• the model is qualified by experience to estimate the number / of isotopes created by interaction (n, 7) in the target 11 relative to a neutron emitted by the source 12 to be characterized.
• This calculation model integrates all the physical and chemical data available on the equipment to be checked to obtain good representativeness.
• the values of the period or half-life T and of the duration T1 are known according to the target 11 used.
• the period or half-life T of the daughter isotope Tal82 for a Tantalum-181 target 11 is 114.6 d and the activation duration T1 of this target 11 is typically of the order of 1 to 6 months.
• the computer device 7 determines the neutron flux F emitted by the source 12 of fissile material from the data relating to the above variables and parameters.
• the computer device 7 is configured to determine the quantity of fissile material present in the zone to be characterized, from the neutron flux F calculated in the previous step and the predetermined data relating to the chemical and isotopic composition of fissile material.
• the mass M of fissile material is equal to the ratio between the neutron flux F and the neutron fluence FO according to the following relationship:
• neutron fluence FO of a source 12 related to one gram of the source essentially depends on the chemical and isotopic composition of this source 12.
• the neutron fluence F0 is a value which can be determined beforehand in a manner known to the person skilled in the art.
• the neutron fluence FO can be determined using two methods.
• the first method consists in using values tabulated in the technical literature in the case where the chemical forms of the source 12 are simple, such as, for example, Pu0 2 or PuF 4 .
• the calculation takes into account the isotopic ratios of Plutonium and the fluence values relating to the given chemical composition.
• the second method is more suitable for complex chemical compositions, and consists of using specialized software such as SOURCE4C.
• the neutron activation device 3 used in the above method can advantageously be configured according to different embodiments.
• FIGs. 4A-4C schematically illustrate a neutron activation device, according to different embodiments of the invention.
• Fig. 4A illustrates a neutron activation device 3 according to a first embodiment, comprising a first neutron activation target 111 giving rise to (n, y) type reactions.
• This device 3 comprises an HDPE matrix 9 forming a block which may have a cylindrical, parallelepipedic shape or another solid shape comprising front 33 and rear 35 faces (for example, planes).
• the HDPE 9 matrix increases detection sensitivity.
• the first target 111 is integrated in a section between the front 33 and rear 35 faces of the HDPE matrix 9 knowing that this section has a surface at least equal to that of the target 111.
• the target 111 is arranged in a first predetermined distance d1 from the rear face 35 of the HDPE matrix.
• this first predetermined distance is of the order of one fifth of the distance d between the front 33 and rear 35 faces of the HDPE matrix.
• the HDPE matrix 9 is a parallelepiped block whose dimensions are of the order of 10 cm ⁇ 10 cm ⁇ 5 cm, the depth between the front 33 and rear 35 square faces of the block being 5 cm.
• the first target 111 has dimensions of the order of 5 cm ⁇ 5 cm ⁇ 0.3 cm and is arranged in a section between the front 33 and rear 35 faces of the HDPE block 9 at a distance d1 of 1 cm of the rear face 35.
• the HDPE block is composed of a stack of five sub-blocks or modules, each of which has dimensions of the order of 5 cm ⁇ 5 cm ⁇ 1 cm.
• the target 111 can be inserted between four sub-blocks on one side and one sub-block on the other side so that the free square faces of the stack form the front 33 and rear 35 faces of the block.
• Tantalum Ta Silver Ag, Terbium Tb, Cobalt Co, Copper Cu, and Hafnium Hf.
• Tantalum, Silver, and Terbium are advantageous to use in areas requiring remote manipulation operations of significant duration, in particular complex access areas.
• each room or installation to be characterized is subjected to a certain level of irradiation and contamination.
• the positioning of the activation device 3 in the zone 31 to be characterized may require tele-operation and/or robotization means.
• these means can advantageously be pooled, which limits the costs and facilitates the installation of several control points making it possible to carry out a neutron “mapping” of the installation.
• Fig. 4B illustrates a neutron activation device, according to a second embodiment where the neutron activation device comprises first and second neutron activation targets.
• the first neutron activation target 111 is integrated in a section between the front 33 and rear 35 faces of the HDPE matrix 9 according to a first predetermined distance from the rear face 35 of the matrix 9 in a manner similar to the embodiment of FIG. . 4A.
• the second neutron activation target 112 is integrated into the HDPE matrix in a plane parallel to that of the first target 111 and according to a second predetermined distance d2 from this first target 111.
• the second predetermined distance d2 between the first 111 and second target 112 is greater than the first predetermined distance d1.
• the ratio between the first d1 and second d2 predetermined distances is of the order of 1 to 4. Using the same example as that of Fig.
• the second target 112 has the same dimensions (ie 5 cm x 5 cm x 0.3 cm) as the first target 111 and is arranged on the front face 33 of the block.
• the distance between the first 111 and second 112 targets is of the order of 4 cm.
• first 111 and second 112 targets made of the same material (for example, Tantalum-181) and according to the configuration above, the ratio of the activities measured on the two targets is representative of the location of the source 12.
• the computing device 7 is configured to determine the location of the dominant neutron source 12 according to the ratio of the activities measured on the first 111 and second 112 targets and taking into account the first d1 and second d2 predetermined distances.
• a ratio close to 1 indicates a location of the source 12 facing the HDPE block 9 while a ratio greater than 4 is characteristic of a location at the rear of the HDPE block 9.
• Fig. 4C illustrates a neutron activation device, according to a third embodiment where the neutron activation device comprises three neutron activation targets.
• the activation device 3 comprises a third neutron activation target 113.
• the third target 113 is a fast target made of a material selected from Nickel and Zinc and preferably Nickel. These materials exhibit significant cross sections for activation in the fast domain. In particular, Nickel is preferred because it has a higher cross section than Zinc.
• the third target 113 is integrated into the HDPE matrix 9 in a plane parallel to those of the first 111 and second 112 targets.
• the second 112 and third 113 targets are integrated in juxtaposition on the front face 33 of the HDPE matrix 9.
• the third target 113 has the same dimensions (i.e. 5 cm x 5 cm x 0.3 cm) as the first 111 and second 112 targets and is arranged below or above the second target 112 on the front face 33 of the block.
• the third target 113 (for example in Nickel) in addition to the first 111 and second 112 targets (for example, in Tantalum) makes it possible to determine the chemical form of the fissile material of the source 12 according to the ratio of the activities measured on the third 113 and second 112 targets. Indeed, the ratio of the activities of these two targets measuring the thermal and fast fluxes makes it possible to determine a characteristic indicator of the chemistry of the source 12 and in particular, of its hydrogen content.
• a second tantalum target 112 on the front face 33 of block 9 makes it possible to measure the thermal component of the spectrum by means of the activation reaction T a 181 (n, g) Ta 182 .
• a third nickel target 113 makes it possible to measure the fast component of the spectrum by means of the Ni 58 (n,p)Co 58 threshold reaction.
• the measurement of the Cobalt-58 activity then the calculation of the activity ratio formed on the two Tantalum and Nickel targets make it possible to identify a signature of the neutron spectrum of the source 12 characteristic of its chemical form.
• the oxide and fluorinated forms of Plutonium, Pu0 2 and PUF 4 present significantly different spectra and are thus easily discernible by this means.
• Fig. 5 illustrates an experimental neutron activation system making it possible to calibrate the response of the activation device to the effect of the neutron source, according to one embodiment of the invention.
• This system comprises a source 12 of Pu0 2 and a neutron activation device 3 similar to that illustrated in FIG. 4C.
• the neutron activation device 3 is a parallelepipedic block 9 in HDPE whose dimensions are of the order of 10 cm ⁇ 10 cm ⁇ 5 cm.
• the area of each of the front 33 and rear 35 faces of block 9 is 10 cm ⁇ 10 cm.
• each of the first 111, second 112 and third 113 targets has dimensions of the order of 5 cm ⁇ 5 cm ⁇ 0.3 cm.
• the three targets are arranged in sections parallel to the front 33 and rear 35 faces of the HDPE block.
• the first target 111 is arranged at a distance of 1 cm from the rear face 35 of the HDPE block while the second 112 and third 113 targets are arranged in juxtaposition on the front face 33 of the HDPE block 9.
• the first 111 and second 112 targets are made of Tantalum while the third target 113 is made of Nickel.
• Figs. 6A and 6B illustrate the gamma spectra of the second 112 and third 113 targets relating to Tantalum and Nickel respectively, recorded after activation by the Pu0 2 source. These spectra show the count number as a function of energy levels in keV.
• the ratio of the activities measured on the basis of the spectra of FIGS. 6A and 6B on the second target 112 and the third target 113 is very sensitive to the chemical composition of the source making it possible, for example, to distinguish between a source of PuF 4 or of Pu0 2 .
• the source 12 is placed practically in contact with the second 112 and third 113 targets at the level of the front face 33 of the block 9.
• the source 12 can be placed at the level of the rear face 35 of the HDPE block.
• Figs. 7A and 7B illustrate the level of activation in the first 111, second 112 and third 113 targets for a source placed at the front 33 and at the rear 35 of the block 9 respectively, according to the depth in each of the targets in mm.
• the source 12 is on the 0 mm side of the abscissa scale for FIG. 7A while it is on the 3 mm side of the abscissa scale for FIG. 7B.
• the location of the main neutron source 12 can then be determined by calculating the ratio of the activities A(III) and A(112) measured on the first target 111 and the second target 112 respectively.
• a source 12 placed facing the block leads to a ratio A(III)/A(112) between 1 and 2.
• a source 12 placed at the rear of the block leads to a ratio A(III)/ A(112) greater than 4.
• Figs. 7A and 7B show that the level of activation in the three targets is not homogeneous whether for a source positioned facing the block or at the rear of the block.
• the inhomogeneity results in a significant difference between the activity measurements of the two faces of each of the targets (5% to 10%), which makes it possible to confirm the location of the main source 12 with respect to the position of block 9.
• the counting of the two faces of the first target 111 allows the detection of an activation heterogeneity by an analysis of the ratio of peaks (high energy/low energy) which indirectly informs on the location of the main source.
• a first advantage is the possibility of measuring neutron fluxes in extremely irradiating environments.
• a second advantage is that the method according to the invention has an optimal sensitivity linked to a synergy between a target (for example, in Tantalum), a block in a neutron thermalizing material (for example in HDPE) as well as a spectrometric measurement by HPGe detector.
• a third advantage is the fact that the time constants of the selected targets are suitable for implementation in hard-to-reach areas leading to long intervention times.
• a fourth advantage is the possibility of estimating the localization of the dominant source by spectroscopic measurement of two targets (preferably in Tantalum) integrated in the same device, also by the spectroscopic measurement of each of the two faces of the first target.
• a fifth advantage is the possibility of identifying the chemical form of the source by a coupled implementation of two targets of different nature (for example, Tantalum and Nickel) measuring the thermal and fast fluxes.
• a sixth advantage is possible reuse of targets after a limited decay phase of about three years.
• a seventh advantage is the relatively low cost of the detection method.
• An eighth advantage is the possibility of using several blocks simultaneously in different locations, making it possible to reduce the level of uncertainty during a check.

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• 2021
  • 2021-05-20 FR FR2105247A patent/FR3123128B1/fr active Active
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