Neutron dosimeter

Abstract

Needle devices for accessing lymph nodes and the thoracic duct. Needle devices for insertion into lymph nodes that are secured after accessing the lymph nodes are described. The needle can aspirate or introduce fluids through a needle opening in the sidewall of the needle. The needle may be curved. The needle device may also have a channel for fluids. The channel may have an opening in the sidewall or may be a separate channel aligned with the lumen and may have a distal opening. Securing the needle can be achieved via a hook, a spiral element, a balloon, or an adhesive, which can be introduced through the open needle tip or the channel opening. The spiral element may be attached to a driver configured to rotate the spiral element.

Description

[Technical field]

[0001] The present invention relates to a radiation monitoring device. In particular, the present invention relates to a neutron dosimeter capable of operating in continuous, quasi-continuous and pulsed neutron fields and capable of detecting neutron radiation in the energy range from 10E-9 (thermal) to 10 MeV or 20 MeV. In particular, the dosimeter is a lightweight neutron dosimeter. The dosimeter can be used in industrial, medical and research facilities, among others. [Background technology]

[0002] Neutrons are produced as primary or secondary radiation in nuclear reactors, accelerator-based facilities, fusion plasmas and radioactive sources. Due to their relatively high biological effectiveness at some energies, neutrons can dominate the total radiation dose received by workers, patients and the public in different facilities. Therefore, detection techniques that provide adequate energy and time structure sensitivity of neutron radiation are required to assess potential risks and determine radiation protection measures inside exposed locations.

[0003] The International Commission on Radiological Protection (ICRP) recommends the use of the ambient dose equivalent (H*(10)) as a practical quantity for area monitoring. (or flux) Conversion factors from radiative to ambient dose equivalent have been compiled and published for various radiation types, including neutrons, in ICRP Publication 74[1].

[0004] Ambient (or environmental) neutron dosimeters are detectors capable of providing readings proportional to neutron fluence. To provide a direct conversion from fluence to H*(10), the neutron energy response of an ideal dosimeter should mimic the shape of the conversion factor curve as a function of neutron energy. In practice, most ambient dosimeters present a dosimetric response that fits the ambient dose equivalent with underestimation and overestimation in different energy regions. International standards organizations such as IEC and ANSI recommend tolerance limits by energy response, which are usually 50%-200% of the ICRP recommendations in different energy regions [2, 3]. Most neutron dosimeters are currently specified with sensitivity from thermal energies up to 20 MeV. Versions with extended energy sensitivity should be specified, at a minimum, up to 200 MeV, which is the current ICRP74 recommendation.

[0005] The first designs of ambient neutron dosimeters date back to the 1960s. These designs consist of a thermal neutron counter surrounded by a moderator, typically made of high density polyethylene. A single layer of neutron absorbing material is also used to compensate for the typical overestimation in thermal energy. The moderator geometries used by these designs are (a) a 21.6 cm diameter / 24.4 cm length cylindrical moderator with cylindrical neutron counter (Anderson-Brown type [4]), (b) a 20.8 cm diameter spherical moderator with spherical neutron counter (Leake type [5]), and (c) a 22.9 cm diameter spherical moderator with cylindrical neutron counter (Hankins type [6]). Extended energy dosimeters use neutron multiplication reactions such as (n,2n') and (n,3n') on a single layer of high-Z material to increase detector sensitivity beyond the evaporation peak (~10-20 MeV) [7]. These classic designs have been adopted by manufacturers with minor variations over time.

[0006] Portability is one of the drawbacks of most of the ambient neutron dosimeters of the prior art. With the exception of the Leake type (~5 kg), most of them have a weight of about 9 kg or more, which obviously imposes serious limitations on their transport, handling and operation as handheld devices. The portability limitation is addressed by the novel NSN3 detector [8], which is based on nuclear reactions based on a specific mixture of gases instead of slowing down and counting thermal neutrons. The last detector weighs only 2.2 kg, but its dosimetric response to epithermal and thermal neutron energies is insufficient. Currently, there is no solution for a dosimeter under 4 kg that works well over a wide neutron energy range (from thermal to vaporization peak).

[0007] Similarly, performance in pulsed neutron fields is also a shortcoming of prior art ambient neutron dosimeters. Typically, pulsed neutron fields are generated by radiation bursts with durations ranging from a few ns to about 1 ms. An intercomparison exercise in pulsed neutron fields [9] showed that most active commercial neutron dosimeters suffer severe underestimation at about 10 nSv / burst or more. The problem is related to the pile-up of strong instantaneous events inside proportional neutron counters, which makes it impossible to count single events and impairs the proportionality of the counters due to space charge accumulation effects

[10] . The shortage of pulsed neutron dosimeters is a major problem for radiation protection in practice, since the number of pulsed facilities, such as spallation sources, high power lasers or fusion plasmas, is increasing. Moreover, pulsed neutron fields are usually generated as a result of beam losses in accelerator-based facilities for research and medical applications.

[0008] A solution for neutron dosimetry in pulsed fields has been proposed in WO 2010 / 040330

[11] . The dosimeter claimed in this invention is based on silver activation technology and therefore provides a reliable estimation of the dose for a single neutron burst. The main drawback of this invention is the response time, which is typically several tens of seconds depending on the half-life of the activated silver isotope.

[0009] The LUPIN detector is another proposed solution for pulsed neutron dosimetry. The detector concept is based on the use of an innovative readout including correction factors, mitigating space charge accumulation effects by using a long proportional neutron counter

[12] . The resulting detector shows real-time monitoring capability with good performance up to about 500 nSv / burst without applying correction algorithms, but has a mass of 18 kg that limits the portability of the device.

[0010] In addition, energy sensitivity is most likely the main problem in prior art ambient neutron dosimeters. Modern facilities for research or medical applications use particle accelerators of extreme energies. For example, particle therapy facilities use proton beams up to 250 MeV and heavy ion beams up to several GeV, respectively. In these energy ranges, several reaction channels are open when the beams interact with the gantry materials or the patient itself. As a result, complex stray secondary neutron fields are generated with contributions ranging from thermal to beam energy. For proton therapy, the only dosimeter that provides a reliable readout in near-clinical conditions of secondary neutrons is the WENDI-II device (US Pat. No. 5,578,830

[13] ,

[14] ). This detector is based on the Anderson-Brown design and includes a single tungsten layer for neutron multiplication. Despite good energy sensitivity performance, the response of this dosimeter to quasi-continuous or pulsed neutron fields is rather (or rather) poor.

[0011] According to the International Commission on Radiation Units and Measurements (ICRU), the current recommendations on operational quantities for radiation protection present several conceptual and technical shortcomings. This motivated ICRU to propose a new set of operational quantities directly related to effective and absorbed doses for an extended range of particles and energy ranges up to 200 GeV

[15] . For neutrons, the ICRU proposal has significant effects for energies below 1 keV and above 20 MeV, respectively, compared to ICRP 74. In the lower energy range, ICRP 74 is very conservative, while in the high energy region, the current recommendations are very underestimated. A recent study concluded that a reoptimization of the non-extended dosimeter design is necessary to better match the requirements of the new proposal for operational quantities

[16] .

[0012] A dosimeter that seeks to overcome the abovementioned drawbacks of energy sensitivity is known from US Pat. No. 7,655,921

[17] , which discloses a dosimeter for detecting neutron radiation in the energy range from 0.025 eV to several hundred GeV, comprising a substantially spherical substrate used as a moderator and containing a hydrogen material, a detector element arranged in the center of the substrate, and a neutron converter surrounding the detector element. The neutron converter contains metal atoms that essentially convert the energy of the high-energy neutron radiation into neutrons in a suitable energy range. The substrate is provided with accesses that allow the detector element to be introduced into and removed from the neutron converter. The neutron converter is embodied in the form of a cylinder.

[0013] Similarly, WO 2009 / 063078

[18] provides an apparatus for detecting radiation, the apparatus comprising an inner core housing a neutron detector and an outer core including a neutron moderator material, the apparatus further comprising at least one elongated thermal neutron guide disposed within the outer core and having an inner end terminating proximate to the neutron detector. In use, the elongated thermal neutron guide directs thermal neutrons towards the neutron detector. The apparatus is lightweight and is intended to reduce the overestimation of low energy neutrons (0.4 eV to 100 keV) characteristic of other apparatus.

[0014] Other instruments for detecting radiation are provided / disclosed by US Pat. No. 6,930,311

[19] , US Pat. No. 10,281,600

[20] and Chinese Patent Application Publication No. 102,928,867

[21] . DISCLOSURE OF THEINVENTION [Means for solving the problem]

[0015] An embodiment of the present invention provides a neutron dosimeter suitable for both continuous and pulsed fields, comprising, as known in the art, a cylinder made of polyethylene material acting as a moderator, an active component acting as a neutron detector and located in the center of the cylinder, and a neutron filter to compensate for the sensitivity of the neutron detector in a given energy range.

[0016] The neutron filter is formed by two (can be three or more) crowns made of a material having a given thickness, and an upper and lower caps, the two crowns being arranged inside the polyethylene moderator, where a first of the two crowns is arranged adjacent to the active component and a second of the two crowns is arranged between the first crown and the inner wall surface of the polyethylene moderator.

[0017] The upper and lower caps are constructed / arranged to cover (at least) the open end of the first crown. The upper cap has an opening therein to allow for the introduction of an active component. The thickness of the material of the upper and lower caps is the same as the thickness of the material used for the two crowns. Further, each of the first and second crowns comprises a plurality of vertical ribbons, and the area covered by the plurality of vertical ribbons in each crown, excluding the upper and lower caps, is configured in the range of 50-95%.

[0018] The neutron dosimeter also includes a removable element disposed between the first crown and the active component for improving the performance of the neutron detector against high intensity neutron radiation fields, either pulsed or continuous. In particular, the removable element includes a hollow cylinder having an inner diameter that is 1.05 to 1.2 times the outer diameter of the active component.

[0019] The neutron dosimeter can be removably attached to the cylinder by pressure means. To that end, the neutron dosimeter can include an annular joint or an O-ring, and the cylinder includes a corresponding groove for the annular joint that is attached and compressed during assembly of the two elements. Alternatively, the attachment / fixing of the neutron dosimeter to the cylinder can be done using screws or fasteners. Furthermore, in another embodiment, the neutron dosimeter can be provided with threads for fixing to the cylinder.

[0020] In one embodiment, the two crowns are made of a cadmium material, in which case the predetermined thickness is comprised in the range of 0.5-1 mm. In another embodiment, the two crowns are made of a borated silicone material, in which case the predetermined thickness is comprised in the range of 3-5 mm.

[0021] In one embodiment, the inner diameter of the first crown is configured in the range of 2.0 to 3.0 cm, and the inner diameter of the second crown is 0.5 to 1.75 cm larger than the inner diameter of the first crown. Alternatively or complementary, the height of the neutron filter is configured in the range of 50 to 100 mm.

[0022] The hollow cylinder may be made of a polyethylene material and a neutron absorbing material such as cadmium, cadmium mixed with boron, polyethylene, polyethylene mixed with boron, or lithium, among others. The neutron absorbing material may in particular comprise a plurality of ribbons (or strips) of any shape, for example vertical, radial, spiral, etc. The neutron absorbing material may cover at least a part of the surface of the active volume (or sensitive volume) of the active component. It should be noted that in any case, the neutron absorbing material may cover the entire surface of the active volume, and may even exceed this surface. Alternatively, the hollow cylinder may be made of a boronated polyethylene material, the boronated polyethylene material being present in a proportion of 0.5-30% by weight.

[0023] In one embodiment, the neutron dosimeter also includes a gamma filter made of a solid lead metal cylinder having an inner diameter comprised in the range of 7-16 mm and a thickness of 1-3 mm.

[0024] In one embodiment, particularly to provide a lightweight neutron dosimeter weighing less than 5 kilograms, the cylinder comprises an outer diameter Rm comprised in the range of 7-10 cm. The total height of the cylinder, i.e., Lm, is determined according to the following empirical formula: Lm(Rm,RA)=1.904·Rm+53.965·RA 2 It can be calculated using -138.65·RA+83.818 [cm], where Rm is given in cm and RA is the nominal anisotropy factor, defined as the ratio of the weighted neutron response of a neutron detector to an omnidirectional fast neutron field to the nominal response to a lateral irradiation of a cylinder containing the active components.

[0025] The neutron dosimeter may further include a beryllium metal or beryllium oxide element disposed between the second crown and the moderator. The beryllium metal or beryllium oxide element may be a third crown consisting of a hollow cylinder or a plurality of vertical ribbons or rods in a coaxial shape.

[0026] In particular, active components are limited to a mass percentage of up to 3%.3 A mixture of He and a quenching gas is included. Similarly, polyethylene materials have a density of 0.90 to 0.98 g / cm. 3 The polyethylene material has a density ranging from 0.1 to 0.5.

[0027] In yet another embodiment, the active component is configured to communicate with an electronic processing module to transmit electrical charges generated in the active component due to the detection of single or multiple neutron events. The electronic processing module can be mounted externally on the cylinder or can be located remotely from the cylinder.

[0028] Thus, the present invention provides a modern radiation sensor capable of overcoming the limitations of current market solutions in ambient dosimetry for continuous and pulsed neutron fields. In particular, the neutron dosimeter i) comprises a cylindrical detector, ii) comprises a removable element that fits with the cylinder, allowing scaling of the detection sensitivity without significantly changing the energy response, and iii) is lightweight (less than 5 kg).

[0029] The dosimeters can be used in nuclear reactors, radioactive sources, neutrons produced by DD fusion, proton therapy facilities, and the like. [Brief description of the drawings]

[0030] The foregoing and other advantages and features will be more fully understood from the following detailed description of the embodiments, taken in conjunction with the accompanying drawings, which should be considered in an illustrative and non-limiting manner. [Figure 1] FIG. 1 shows a schematic representation of the proposed neutron dosimeter according to one embodiment of the present invention. [Diagram 2] FIG. 2 shows a schematic representation of another embodiment of the proposed neutron dosimeter, in this case including an efficiency plug or removable element. [Diagram 3] FIG. 3 shows a schematic representation of another embodiment of the proposed neutron dosimeter, in this case including a power and electronic processing module. [Figure 4]FIG. 4 shows the dosimetric response obtained in Example 1. [Diagram 5] FIG. 5 shows the relationship between active component area coverage and neutron sensitivity scaling factor for Example 2. [Figure 6] Figure 6 shows the workplace performance. The upper and lower limits of commercial dosimeters reported by the EVIDOS project

[22] . [Figure 7] The performance of the pulsed field is shown in Figure 7. The data for the commercial dosimeters were extracted from

[24] . DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0031] The present invention provides a cylindrical neutron dosimeter 1, in particular of the Anderson-Brown type, capable of operating in continuous, quasi-continuous and pulsed neutron fields. The dosimeter can be used with electronic processing modules such as charge sensitive preamplifiers and readouts based on either analog or digital electronics. In some embodiments, the neutron dosimeter 1 has a maximum weight of less than 5 kilograms (excluding electronics), i.e. the dosimeter is lightweight.

[0032] The energy sensitivity of the neutron dosimeter 1 complies with recommendations ICRP74 or ICRU-RC26. Similarly, the neutron detector is designed for operating energies from 10E-9 (thermal) to 10 MeV or 20 MeV (the latter when using a neutron multiplier). The deviation from the average detection sensitivity is less than 20% from the nominal sensitivity for omnidirectional fast neutron fields or working fields.

[0033] Figures 1 and 2A-B show different embodiments of the proposed neutron dosimeter 1. The neutron dosimeter 1 comprises a cylindrical body 100 (or passive component) acting as a moderator, an active component 101 acting as a neutron detector and a neutron filter 102 compensating the sensitivity of the neutron detector in the thermal and epithermal energy range.

[0034] In one embodiment, the cylindrical body 100 is made of a high density polyethylene material (density 0.90 to 0.98 g / cm 3 ) The outer diameter (Rm) of the cylinder 100 can be from 7 cm to 9.5 or 10.0 cm. The total height (Lm) of the cylinder 100 can be calculated using the following formula: Lm(Rm,RA)=1.904·Rm+53.965·RA 2 -138.65·RA+83.818[cm] (Formula 1) where Rm is given in cm and RA is the nominal anisotropy factor. RA is defined as the ratio between the weighted neutron response of the neutron detector to an omnidirectional fast neutron field and the nominal response to lateral irradiation of the cylinder containing the active component 101. The aspect ratio (Lm / Rm) of the moderator 100 is between 1.2 and 1.8. Likewise, in particular, RA can be between 0.8 and 1.3.

[0035] Note that in other embodiments, the outside diameter (Rm) of the cylinder can be up to 16.0 cm, although not limited to a mass of less than 5 kilograms.

[0036] The active component 101 includes an active volume or length 111. The active component 101 includes: 3 It is filled with He gas and quenching gas (3% by mass or less). The outer diameter of the active component 101 may be in the range of 0.5 to 1.0 inches, and the sensitive volume 111 may be in the range of 30 to 70 mm, particularly 50 mm.

[0037] The neutron filter 102 may be made from cadmium foil having a thickness of 0.5-1 mm, or may be made from a borated silicone material having a thickness of 3-5 mm.

[0038] The neutron filter 102 is placed inside the cylinder 100 formed as two crowns 102A, 102B and upper and lower caps 102C, 102D. The first crown 102A is placed adjacent to the active component 101, and the second crown 102B is placed between the first crown 102A and the inner wall surface of the cylinder 100. The upper and lower caps 102C, 102D are included to cover the open end of the first crown 102A. The upper cap 102C has an opening that allows the active component 101 to be introduced therein. The material thickness of the upper and lower caps 102C, 102D is about 0.5 mm, the same as the material thickness of the two crowns 102A, 102B.

[0039] The first and second crowns 102A, 102B (not shown in the figures) are made from a discrete number of vertical ribbons. The area covered by the vertical ribbons of each crown is 50-95%, excluding the upper and lower caps 102C, 102D. According to a particular embodiment, the first crown 102A can be made from six vertical ribbons and the second crown 102B can be made from eight vertical ribbons. The height of the two crowns 102A, 102B can be between 50-100 mm.

[0040] Alternatively, in another embodiment not shown, the neutron filter 102 is disposed inside the cylinder 100 as a hollow coaxial cylinder. In this case where a single filter is used, the inner diameter may be 2.5-5 cm.

[0041] The neutron dosimeter 1 can also incorporate a gamma filter to reduce the sensitivity of the active component 101 to X-rays. In particular, the gamma filter consists of a solid lead metal cylinder. The inner diameter of said metal cylinder can vary from 7 to 16 mm and its thickness can be from 1 to 3 mm.

[0042] Furthermore, a metallic beryllium or beryllium oxide element acting as a neutron multiplier can be used to compensate for the under-sensitivity of the neutron detector for energies higher than 10 MeV. In particular, the metallic beryllium or beryllium oxide element is placed between the second crown 102B and the cylindrical body 100 and can be either in the form of a hollow cylinder or in the form of a third crown made of several vertical ribbons or rods in a coaxial shape. The inner diameter of the cylinder or crown can be between 4 and 7 cm. If vertical ribbons are used, the thickness of the material is between 3 and 12 mm. If rods are used, the diameter of the rods will be between 4 and 10 mm. If crowns are used, the total angle subtended by all ribbons or rods ranges from 72 degrees to 165 degrees.

[0043] 2A and 2B, in addition to the aforementioned elements, the neutron dosimeter 1 also includes a removable plug (or element) 104 inserted into the cylinder 100 to compensate for the undersensitivity of the neutron detector to energies higher than 10 MeV (the arrow at the bottom of FIG. 2A indicates the direction of insertion of the removable element 104 into the passive component 100). The removable element 104 comprises a hollow cylinder having an inner diameter that is 1.05 to 1.2 times the outer diameter of the active component 101.

[0044] The hollow cylinder may be made of neutron absorbing material 106 having multiple ribbons / strips and a polyethylene material (e.g., high density polyethylene material). The neutron absorbing material 106 may be made of different materials such as lithium, cadmium, polyethylene, boronized polyethylene, etc. The neutron absorbing material 106 may cover the active volume 111 of the active component 101. In particular, the neutron absorbing material 106 may cover the active volume 111 of the active component 101 except for the top and bottom ends, which are covered by the neutron absorbing material 106 within the hollow cylinder. 3The percentage of the He-filled tube area is 20% to 95%. Alternatively, the hollow cylinder can be made of boronated polyethylene material, which is present in an amount of 0.5 to 30% by weight. A gamma filter 105 is included as part of the removable plug 104. In this case, the gamma filter 105 has a thickness of 1 to 3 mm.

[0045] FIG. 3 shows an embodiment in which the neutron dosimeter 1 also includes an electronic processing module 108 mounted on top of the cylinder 100. The electronic processing module 108 communicates with the active component 101 and is operably connected to send readouts to different computing devices such as smartphones, computers, servers, etc. The electronic chain includes high and low voltage DC power supplies. The signal processing and readout can be based on either analog and / or digital electronics. When measuring continuous or quasi-continuous neutron fields, the neutron dosimeter 1 operates by using a rate meter or equivalent electronics capable of detecting single events (pulsed mode) and providing a dose rate in appropriate units. In case of pulsed neutron fields, the neutron dosimeter 1 operates in a charge integration mode including electronics capable of determining the total charge deposited inside the active component 101 by the detected events from a neutron burst and providing the integrated dose by the neutron burst (pulsed dose rate).

[0046] It should be noted that in other embodiments not shown, the electronic processing module can be located remotely from the cylinder 100 .

[0047] Different specific examples of the proposed neutron dosimeter 1 are detailed below.

[0048] Example 1 According to this example, the neutron dosimeter is lightweight with a total mass of less than 5 kg and has a specified energy range from thermal to 10 MeV. The active component 101 is 3It is a He-filled cylindrical neutron counter with a nominal pressure of 10 atm, an outer diameter of 1 inch, a total tube length of 74.6 mm, and a nominal length of detection volume of 50 mm. The cylinder 100 has a nominal density of 0.95 gr / cm 3 The neutron filter 102 is made of a high density polyethylene material with a thickness of 0.5 mm. The nominal anisotropy coefficient is chosen to be RA=1.1. The outer diameter of the moderator is chosen to be Rm=9 cm. The length of the moderator is therefore 13.7 cm. The neutron filter 102 is made of natural cadmium material with a thickness of 0.5 mm. The first crown 102A is formed of six vertical ribbons with a length of 75 mm including the top and bottom caps and an inner diameter of 2.25 cm. The arc subtended by each ribbon of the first crown 102A is 51 degrees. The second crown 102B is made of eight vertical ribbons with an inner diameter of 3.25 cm, each ribbon subtending an arc of 36 degrees. The active component 101 and the neutron filter 102 are arranged symmetrically with respect to the radial and axial axes of the cylindrical body 100. The total mass of the device including the elements 100-102 is 3.5 kg.

[0049] The reactions in this example were calculated by Monte Carlo simulations using the GEANT4 toolkit (https: / / geant4.web.cern.ch / ). The dose-response is 252 The dosimetric response is shown in Figure 4 and compared with similar calculations for commercial dosimeters with extended and non-extended energy ranges. The energy response provided by the lightweight dosimeter 1 of this example shows satisfactory performance and is similar to commercial devices.

[0050] The performance of the neutron spectrum in different workplaces, including nuclear reactors, transport casks, fuel processing nuclear facilities and fast neutron sources, was also studied. Reference data, as reported by the EVIDOS project

[22] and ISO recommendation 8529-2001, are provided in Table 1. The calculations of this example are shown in Figure 6 and compared with typical values. The lightweight neutron dosimeter 1 provides similar performance in the workplace as heavier commercial devices. TIFF0007601420000001.tif47169 Table 1: Reference neutron spectra for reactors, transport casks, fuel processing nuclear facilities and workplaces including fast neutron sources.

[0051] Example 2 The lightweight neutron dosimeter according to Example 1 is modified with a removable element 104 to adapt the neutron sensitivity to very intense neutron fields. The removable element 104 used in this example has a diameter of 0.95 gr / cm 3 The hollow cylinder is made of high density polyethylene material having a nominal density of 1.0 mm. The hollow cylinder has an inside diameter of 27.5 mm.

[0052] The general relationship between the coverage of the area of ​​the active component 101 and the magnification of the neutron sensitivity in the first embodiment is shown in Fig. 5. In this embodiment, the absorbing neutron material 106 is natural cadmium inserted inside the hollow cylinder in the form of a vertical ribbon covering 75% of the area of ​​the active component 101. The dosimeter sensitivity is therefore reduced by a factor of 0.46. The impact of this embodiment on the total mass is less than 100 gr.

[0053] 6 shows calculations of performance in different workplaces for Example 2. These calculations demonstrate that the removable element 104 can effectively scale neutron sensitivity without significantly compromising the overall energy response.

[0054] Example 3 The neutron dosimeters described in Examples 1 and 2 are used in charge integration mode with an appropriate electronic processing module to measure pulsed neutron fields. The specific response of the active component 101 described in Example 1 has been experimentally characterized in pulsed fields generated by a spallation neutron source. These data have been used to derive the dosimeter performance in pulsed neutron fields.

[0055] The performance of the pulsed neutron fields in Examples 1 and 2 is shown in Figure 7 and compared with reference data for commercial dosimeters

[23] . This lightweight neutron dosimeter in Example 3 is capable of measuring pulsed neutron doses higher than 1 uSv per bunch without being compromised by pile-ups of intense events. Furthermore, the use of a removable element 104 for the device in Example 2 provides additional flexibility to increase the dynamic range by a factor of two.

[0056] In other embodiments, not shown in this case, the proposed removable element or plug 104 may also be used in other detectors based on neutron moderators of any shape to compensate for inherent sensitivity of active components in order to reduce pile-ups or latency in counting rates. In this case, the body (or other element) of the neutron counting detector should have a configuration compatible with the removable element so that the removable element 104 can be fixed / attached thereto.

[0057] The present disclosure and / or some other examples have been described above. According to the above description, various modifications can be made. The topic of the present disclosure can be achieved in various forms and embodiments. All applications, modifications and changes that are required to be protected in the claims may be within the protection scope of the present disclosure.

[0058] The scope of the invention is defined in the following set of claims. The following is an invention described at the beginning of the present application. <Claim 1> a cylindrical body (100) made of a polyethylene material acting as a moderator; an active component (101) acting as a neutron detector, said active component (101) being located at the center of said cylindrical body (100); a neutron filter (102) configured to compensate for the sensitivity of the neutron detector in a given energy range, a neutron filter, the neutron filter (102) being constituted by at least two crowns (102A, 102B) made of a material having a predetermined thickness, the at least two crowns (102A, 102B) being arranged inside the cylindrical body (100), a first (102A) of the two crowns (102A, 102B) being arranged adjacent to the active component (101) and a second (102B) of the two crowns (102A, 102B) being arranged between the first crown (102A) and an inner wall surface of the cylindrical body (100); and upper and lower caps (102C, 102D) configured to cover an open end of the first crown (102A), the upper cap (102C) having an opening to allow introduction of the active component (101) therein, The thickness of the material of the upper and lower caps (102C, 102D) is the same as the thickness of the material of the two crowns (102A, 102B); The first and second crowns (102A, 102B) each include a plurality of vertical ribbons, and the area of ​​each crown covered by the plurality of vertical ribbons is in the range of 50 to 95%, excluding the upper and lower caps (102C, 102D); The neutron dosimeter a removable element (104) disposed between the first crown (102A) and the active component (101) for improving the performance of the neutron detector against intense neutron radiation fields, whether pulsed or continuous, the removable element (104) comprising a hollow cylinder having an inner diameter between 1.05 and 1.2 times the outer diameter of the active component (101); A neutron dosimeter comprising: <Claim 2> The neutron dosimeter of claim 1, wherein the two crowns (102A, 102B) are made of a cadmium material and have a predetermined thickness in the range of 0.5 to 1 mm. <Claim 3> The neutron dosimeter of claim 1, wherein the two crowns (102A, 102B) are made of a borated silicone material and have a predetermined thickness in the range of 3 to 5 mm. <Claim 4> The inner diameter of the first crown (102A) is in the range of 2.0 to 3.0 cm, and the inner diameter of the second crown (102B) is 0.5 to 1.75 cm larger than the inner diameter of the first crown (102A); and / or The height of the neutron filter (102) is in the range of 50 to 100 mm. 4. The neutron dosimeter according to claim 1. <Claim 5> 2. The neutron dosimeter of claim 1, wherein the hollow cylinder is made of a polyethylene material and a neutron absorbing material (106) having a plurality of ribbons / strips, the neutron absorbing material (106) covering at least a portion of a surface of an effective volume (111) of the active component (101). <Claim 6> 2. The neutron dosimeter according to claim 1, wherein the hollow cylinder is made of a boronized polyethylene material and the boronized polyethylene material is contained in an amount of 0.5 to 30% by weight. <Claim 7> 7. The neutron dosimeter according to claim 1, 5 or 6, further comprising a gamma filter (105) made of a solid lead metal cylinder having an inner diameter comprised in the range of 7-16 mm and a thickness of 1-3 mm. <Claim 8> The dosimeter is a lightweight dosimeter, and the cylindrical body (100) has an outer diameter Rm within a range of 7 to 10 cm and a radius Lm(Rm, RA)=1.904Rm+53.965RA 2 8. The neutron dosimeter according to claim 1, further comprising a total height Lm calculated by -138.65RA+83.818 [cm], where Rm is given in cm, and RA is the nominal anisotropy factor, defined as the ratio of the weighted neutron response of the neutron detector to an omnidirectional fast neutron field to the nominal response to a side irradiation of a cylinder containing the active component (101). <Claim 9> 9. The neutron dosimeter of claim 8, further comprising an element of beryllium metal or beryllium oxide disposed between the second crown (102B) and the cylindrical body (100). <Claim 10> 10. The neutron dosimeter of claim 9, wherein the beryllium metal or beryllium oxide element is a third crown consisting of a hollow cylinder or a plurality of vertical ribbons or rods in a coaxial shape. <Claim 11> Neutron dosimeter according to claims 1 to 10, wherein the active component (101) comprises a mixture of 3He and a quenching gas in a proportion of up to 3% by weight. <Claim 12> 12. The neutron dosimeter according to claim 1, wherein the polyethylene material is a high-density polyethylene material having a density in the range of 0.90 to 0.98 g / cm3. <Claim 13> 13. The neutron dosimeter according to claim 8, wherein the total dosimeter mass is less than 5 kilograms. <Claim 14> A neutron dosimeter as described in claims 1 to 13, wherein the active component (101) is configured to communicate with an electronic processing module (108) to transmit electric charges generated in the active component (101) due to detection of single or multiple neutron events. <Claim 15> The neutron dosimeter of claim 14, wherein the electronic processing module is mounted externally on the cylindrical body (100) or located remotely from the cylindrical body (100).

[0059] References: [1] ICRP, 1996. Conversion Coefficients for use in Radiological Protection against External Radiation. ICRP Publication 74. Ann. ICRP 26 (3-4). [2] International Electrotechnical Commission (IEC) 2014 Radiation protection instrumentation Neutron ambient dose equivalent (rate) meters IEC Report. [3] American National Standards Institute (ANSI) 2003 American national standard for performance specifications for health physics instrumentation-portable instrumentation for use in normal environmental conditions ANSI Report. [4] Andersson, I. Oe., & Braun, J. (1964). A Neutron Rem Counter (AE--132). Sweden. [5] J.W. Leake, Nuclear Instruments and Methods 63:3 (1968) 329-332. [6] Hankins, D. E. Health physics. 14(5) (1968) 518-20. [7] C. Birattari, A. Ferrari, C. Nuccetelli, M. Pelliccioni, M. Silari, NIMA 297 (1990) 250-257. [8] T. Nunomiya et al. Journal of Nuclear Science and Technology, 54:11 (2017) 1215-1222. [9] M. Caresana et al. NIMA 737 (2014) 203-213.

[10] I. Rios, J. Gonzalez, R.E. Mayer, Radiation Measurements 53-54 (2013) 31.

[11] “Neutron Dosimeter”. 2008

[12] C. Cassell et al. NIMA 804 (2015) 113-117.

[13] “Neutron dose equivalent meter”. 1995

[14] R. H. Olsher et al. Health Phys. 79:2 (2000) 170-81.

[15] T Otto et al. Radiation Protection Dosimetry 180 (2018) 10-16.

[16] J S Eakins et al. J. Radiol. Prot. 38 (2018) 688

[17] “Dosimeter for the detection of neutron radiation”. 2004

[18] “Radiation Detection”. 2007

[19] “Lightweight neutron remmeter”. 2003

[20] “Neutron Detector and Dose Rate Meter using Beryllium-Loaded Materials”. 2017

[21] “Compensation type neutron dose instrument”. 2012

[22] H. Schuhmacher et al, “Evaluation of individual dosimetry in mixed neutron and photon radiation fields”, PTB-N-49 report, 2006.

[23] M. Caresana et al., Nuclear Instruments and Methods in Physics Research A 737 (2014) 203-213.

[24] A. Tarifeno-Saldivia et al., Review of Scientific Instruments 85, 013502 (2014).

Claims

1. a cylindrical body (100) made of polyethylene material acting as a decelerator; an active component (101) acting as a neutron detector, said active component (101) being located at the center of said cylindrical body (100); A neutron filter (102) configured to compensate for the sensitivity of the neutron detector in a given energy range, The neutron filter (102) is composed of at least two crowns (102A, 102B) and upper and lower caps (102C, 102D) made of a material having a predetermined thickness, the at least two crowns (102A, 102B) are arranged inside the cylindrical body (100), a first (102A) of the two crowns (102A, 102B) is arranged adjacent to the active component (101), and a second (102B) of the two crowns (102A, 102B) is arranged between the first crown (102A) and the inner wall surface of the cylindrical body (100). A neutron dosimeter comprising: the upper and lower caps (102C, 102D) are configured to cover an open end of the first crown (102A), the upper cap (102C) having an opening therein to allow introduction of the active component (101); The thickness of the material of the upper and lower caps (102C, 102D) is the same as the thickness of the material of the two crowns (102A, 102B); the first and second crowns (102A, 102B) each comprise a plurality of vertical ribbons, and the area of ​​each crown covered by the plurality of vertical ribbons, excluding the upper and lower caps (102C, 102D), is in the range of 50% to 95%; The neutron dosimeter a removable element (104) disposed between the first crown (102A) and the active component (101) adapted to improve the performance of the neutron detector against high intensity neutron radiation fields, either pulsed or continuous, the removable element (104) comprising a hollow cylinder having an inner diameter between 1.05 and 1.2 times the outer diameter of the active component (101); A neutron dosimeter comprising:

2. The neutron dosimeter of claim 1, wherein the two crowns (102A, 102B) are made of cadmium material and said predetermined thickness is in the range of 0.5 to 1 mm.

3. The neutron dosimeter of claim 1, wherein the two crowns (102A, 102B) are made of borated silicone material, and said predetermined thickness is in the range of 3-5 mm.

4. the inner diameter of the first crown (102A) is in the range of 2.0-3.0 cm and the inner diameter of the second crown (102B) is 0.5-1.75 cm larger than the inner diameter of said first crown (102A); and / or The neutron dosimeter of claim 1, wherein the height of the neutron filter (102) is in the range of 50 to 100 mm.

5. 2. The neutron dosimeter of claim 1, wherein the hollow cylinder is made of a polyethylene material and a neutron absorbing material (106) having a plurality of ribbons / strips, the neutron absorbing material (106) covering at least a portion of a surface of an effective volume (111) of the active component (101).

6. 2. The neutron dosimeter of claim 1, wherein the hollow cylinder is made of a boronized polyethylene material and the boronized polyethylene material comprises 0.5-30% by weight.

7. The neutron dosimeter of claim 1, 5 or 6, further comprising a gamma filter (105) made of a solid lead metal cylinder having an inner diameter comprised in the range of 7-16 mm and a thickness of 1-3 mm.

8. The dosimeter is a lightweight dosimeter, and the cylindrical body (100) has an outer diameter Rm in the range of 7 to 10 cm and Lm(Rm, RA)=1.904·Rm+53.965·RA ２ Including the total height Lm calculated by -138.65 RA + 83.818 [cm], - Rm is given in cm; 2. The neutron dosimeter of claim 1, wherein -RA is a nominal anisotropy coefficient, defined as the ratio of a weighted neutron response of the neutron detector to an omnidirectional fast neutron field to a nominal response to a lateral irradiation of a cylinder containing the active component (101).

9. The neutron dosimeter of claim 8, further comprising an element of beryllium metal or beryllium oxide disposed between the second crown (102B) and the cylindrical body (100).

10. 10. The neutron dosimeter of claim 9, wherein the beryllium metal or beryllium oxide element is a third crown consisting of a hollow cylinder or a plurality of vertical ribbons or rods of coaxial shape.

11. The active component (101) is mixed with a quenching gas in a proportion of up to 3% by weight. ３ 2. The neutron dosimeter of claim 1 comprising a mixture of He.

12. The polyethylene material has a viscosity of 0.90 to 0.98 g / cm ３ 2. The neutron dosimeter of claim 1, wherein the material is a high density polyethylene material having a density in the range of 0.1 to 0.

5.

13. A neutron dosimeter according to claims 8 to 12, wherein the total mass of the dosimeter is less than 5 kilograms.

14. 2. The neutron dosimeter of claim 1, wherein the active component (101) is in communication with an electronic processing module (108) to transmit electrical charges generated in the active component (101) due to detection of single or multiple neutron events.

15. The neutron dosimeter of claim 14, wherein the electronic processing module is mounted externally on the cylindrical body (100) or located remotely from the cylindrical body (100).

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