Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C13/00—Pressure vessels; Containment vessels; Containment in general
  • G21C13/02—Details
  • G21C13/032—Joints between tubes and vessel walls, e.g. taking into account thermal stresses
  • G21C13/036—Joints between tubes and vessel walls, e.g. taking into account thermal stresses the tube passing through the vessel wall, i.e. continuing on both sides of the wall
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C17/00—Monitoring; Testing ; Maintaining
  • G21C17/017—Inspection or maintenance of pipe-lines or tubes in nuclear installations
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
  • G21C19/20—Arrangements for introducing objects into the pressure vessel; Arrangements for handling objects within the pressure vessel; Arrangements for removing objects from the pressure vessel
  • G21C19/207—Assembling, maintenance or repair of reactor components
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C7/00—Control of nuclear reaction
  • G21C7/06—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section
  • G21C7/08—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods
  • G21C7/12—Means for moving control elements to desired position
  • G21C7/14—Mechanical drive arrangements
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
  • Y02E30/00—Energy generation of nuclear origin
  • Y02E30/30—Nuclear fission reactors

Definitions

• TITLE A thermal sleeve for a control rod drive mechanism and associated nuclear reactor head, nuclear reactor and method for replacing a damaged thermal sleeve
• the present invention concerns a thermal sleeve for a control rod drive mechanism of a nuclear reactor pressure vessel.
• the present invention concerns also a method for replacing a damaged thermal sleeve of a control rod drive mechanism of a nuclear reactor pressure vessel.
• Control rods are used in nuclear reactors to control the fission rate of the nuclear fuel in the reactor vessel.
• a control rod drive mechanism is designed to insert, withdraw or maintain the position of these control rods from which neutron absorbers are suspended via a drive shaft connecting the control rod drive mechanism and the associated control rods.
• the reactor vessel closure head comprises thermal sleeves, whose main function is to shield the control rod drive mechanism nozzle form thermal shocks when the control drive rod is withdrawn from the core.
• the thermal sleeves wear over time due to vibrations and eventually fail.
• the reactor vessel closure head is removed from the vessel and placed on a head stand, then the control rod drive mechanism is fully removed and the thermal sleeve is replaced.
• the document US 2019/025082 discloses a method for replacing a damaged thermal sleeve with a new thermal sleeve comprising several elastic leaves protruding from a tube. The leaves enable to support the thermal sleeve during all the operation of the nuclear plant by resting on the nozzle.
• One of the aims of the invention is therefore to provide a thermal sleeve for a control rod drive mechanism which improves the safety of the reactor vessel by presenting a better thermal protection, rigidity and fatigue resistance.
• the invention proposes a thermal sleeve for a control rod drive mechanism of a nuclear reactor pressure vessel the thermal sleeve comprising a tube comprising an upper end and a lower end, the tube comprising also an outer surface and an inner surface, the outer surface and the inner surface extending from the upper end to the lower end ; a plurality of protrusions disposed at the upper end of the tube, each protrusion extending radially outward the tube; and a collar protruding radially outward the outer surface of the tube and extending circumferentially all around the tube, away from the upper end.
• the thermal sleeve comprises one or several of the optional following features, taken individually or according to any technical feasible combination:
• protrusions are circumferentially spaced and separated from each other by circumferentially spaced slots;
• the collar is manufactured by machining the tube
• the collar is manufactured by a plastic expansion of the tube.
• the invention also concerns a nuclear reactor head comprising a vessel top head, a plurality of control rod drive mechanisms comprising each a control shaft, and a plurality of penetrations extending each through the vessel top head for allowing passage of one of the control shafts through the vessel top head, each penetration comprising a nozzle extending through the vessel top head and a thermal sleeve as defined previously extending inside the associated nozzle and coaxially with the nozzle.
• the collar is manufactured by a plastic expansion of the tube.
• the invention also concerns a nuclear reactor comprising a reactor vessel, the reactor vessel comprising a vessel shell and a nuclear reactor head as defined previously the vessel top head of the nuclear reactor head closing an upper opening of the vessel shell, a control rod guide tube located inside the reactor vessel, a control rod assembly guided inside the control rod guide tube, a control rod drive mechanism located outside the reactor vessel, comprising a control shaft extending through the penetration and connected to the control rod assembly for controlling movement of the control rod assembly inside the control rod guide tube.
• the nuclear reactor comprises one or several of the optional following features, taken individually or according to any technical feasible combination:
• the nuclear reactor comprises also a plurality of control guide tubes of the nuclear reactor allowing passage of one of the control shafts, said control guide tubes supporting the lower end of thermal sleeves;
• the nuclear reactor further comprises a spacer attached to the upper end of the control rod guide tube for maintaining minimal spacing with the lower end of each thermal sleeve;
• the invention also concerns a method for replacing a damaged thermal sleeve of a nuclear reactor head, the nuclear reactor head comprising a vessel top head, a plurality of control rod drive mechanisms comprising each a control shaft, and a plurality of penetrations extending each through the vessel top head for allowing passage of one of the control shafts through the vessel top head, each penetration comprising a nozzle extending through the vessel top head and the damaged thermal sleeve extending inside the associated nozzle and coaxially with the nozzle, the method comprising at least the following steps :
• the new thermal sleeve comprising a tube comprising an upper end and a lower end, the tube comprising also an outer surface and an inner surface, the outer surface and the inner surface extending from the upper end to the lower end, a plurality of protrusions disposed at the upper end of the tube, each protrusion extending radially outward the tube, and a collar protruding radially outward the outer surface of the tube and extending circumferentially all around the tube, away from the upper end.
• the method comprises one or several of the optional following features, taken individually or according to any technical feasible combination:
• the step of replacing the damaged sleeve by a new thermal sleeve comprises at least a sub-step of plastic deformation of the protrusions so that each protrusion extends radially outward the tube, the tube being inserted in the nozzle;
• the step of replacing the damaged sleeve by a new thermal sleeve comprises at least a sub-step of plastic deformation of the protrusions so that each protrusion extends radially outward the tube, the tube being inserted in the nozzle;
• the step of replacing the damaged sleeve by a new thermal sleeve comprises at least a sub-step of plastic deformation of the protrusions so that each protrusion extends radially outward the tube, the tube being inserted in the nozzle;
• the nuclear reactor comprises also a plurality of control guide tubes of the nuclear reactor allowing passage of one of the control shafts, the method comprising a step of attaching a spacer to the upper end of a control rod guide tube for maintaining a minimal spacing with the lower end of the corresponding thermal sleeve.
• - figure 1 schematically shows a cross-sectional view of a nuclear reactor pressure vessel
• - figure 2 shows a cross-sectional view of a control rod drive mechanism nozzle of figure 1 and an associated conventional thermal sleeve
• FIG. 3 shows a perspective view of a thermal sleeve according to the invention
• FIG. 4 shows a cross sectional view of a first embodiment of the thermal sleeve of figure 3;
• FIG. 5 shows a cross sectional view of a second embodiment of the thermal sleeve of figure 3.
• the nuclear reactor 2 of figure 1 comprises a nuclear reactor vessel 4 containing a reactor core 6.
• the nuclear reactor vessel 4 has a substantial vertical central axis A-A’.
• the terms “vertical”, “horizontal”, “upper”, “lower”, “bottom” and “top” are in reference to the vertical central axis A-A’ of the reactor vessel 4.
• the nuclear reactor vessel 4 comprises a substantially cylindrical vessel shell 8 extending along the central axis A and a vessel bottom head 10 closing the bottom end of the vessel shell 8.
• the vessel shell 8 has an opening at the top end thereof.
• the nuclear reactor vessel 4 comprises a nuclear reactor head 12 that includes a vessel top head 14 closing the top end of the vessel shell 8.
• the vessel top head 14 is removable and attached to the vessel shell 8 via screws 16.
• the reactor core 6 includes fissile material.
• the reactor core 6 is made of a plurality of nuclear fuel assemblies 18 that are arranged side-by-side inside the nuclear reactor vessel 4.
• Each nuclear fuel assembly 18 typically comprises a bundle of fuel rods supported by an armature in a spaced relationship with each fuel rod containing fissile material.
• the nuclear reactor vessel 4 has coolant fluid inlets 20, coolant flow outlets 22, and an internal sleeve 24 surrounding the reactor core 6.
• coolant fluid entering the coolant fluid inlets 20 flows downwardly, outside the internal sleeve 24, enters the internal sleeve 24 at a bottom end, and then flows vertically upward inside the internal sleeve 24 through the reactor core 6, before exiting the nuclear reactor vessel via the coolant flow outlets 22.
• the nuclear reactor 2 comprises control rods, not represented, that are movable inside and outside the reactor core 6 to adjust the reactivity of the reactor core 6.
• the control rods contain neutron-absorbing material.
• the control rods combine into control rod assemblies. Each control rod assembly comprises a bundle of control rods.
• the nuclear reactor 2 comprises control rod guiding tubes 30.
• Each control rod guiding tube 30 is located inside the nuclear reactor vessel 4 above the reactor core 6, and receives a respective control rod assembly that slides vertically inside the control rod guiding tube 30.
• Each control rod guiding tube 30 provides guidance for the corresponding control rod assembly to ensure that each control rod remains aligned with the corresponding space provided in the reactor core 6.
• the nuclear reactor 2 comprises control rod drive mechanisms 32 that are configured for moving the control rods, specifically the control rod assemblies. Each control drive mechanism 32 is associated to one respective control rod assembly.
• Each control rod drive mechanism 32 is located outside the reactor vessel 4 and drives a respective control rod assembly via a drive shaft that extends though the vessel top head 14. Each drive shaft connects a respective control rod drive mechanism 32 to the corresponding control rod assembly through the vessel top head 14.
• the nuclear reactor head 12 comprises penetrations 35 extending through the vessel top head 14. Each penetration 35 has the configuration to receive a respective drive shaft 34. Each drive shaft 34 is slidably received inside the corresponding penetration 35.
• Each penetration 35 extends through the vessel top head 14 along the vertical axis A-A’, and is composed of a nozzle 37 welded to the vessel top head 14 and a thermal sleeve 39 that is received inside the nozzle 37.
• Each penetration 35 protrudes longitudinally along the vertical axis A-A’ past the vessel top head 14 such that a first end of the assembly 36 said outer end 41 and a first end 43 of the thermal sleeve 39 are positioned outside of the pressure vessel 10.
• Each penetration 35 also protrudes longitudinally along the vertical axis A-A’ past the vessel top head 14 such that a second end said inner end 45 of the nozzle 37 and a second end 34 of the thermal sleeve 39 are positioned in the interior of the nuclear reactor vessel 4.
• the nozzle 37 includes an intermediate portion 47 extending from the outer end 41 to the inner end 45 through the vessel top head 14.
• the conventional thermal sleeve 39 includes an intermediate portion 38 extending from the first end 43 to the second end 34 through the vessel top head 14.
• the outer end 41 of the nozzle 37 extends vertically along the vertical axis A-A’ upward further than the first end 43 of the thermal sleeve 39 away from the vessel top head.
• the outer end 41 of the nozzle 37 includes a radially enlarged annular portion 40.
• the radially enlarged annular portion 40 is radially thicker than the intermediate portion 36 of the nozzle 37.
• the radially enlarged annular portion 40 has an outer circumferential surface 40a that is radially further away from the vertical axis A-A’ than an outer circumferential surface 36a of the intermediate portion 36.
• the radially enlarged annular portion 40 includes a lower section 42 having an inner circumferential surface 42a of a same diameter as an inner circumferential surface 36b of the intermediate portion 36.
• an inner diameter of the enlarged annular portion 40 defines a radially enlarged flared support section 44.
• the support section 44 is formed as an annular shoulder having a frustoconical inner circumferential support surface 44a extending radially away from the inner circumferential surface 42a while extending axially upward to join an inner circumferential surface 46a of an upper section 46 of the enlarged annular portion 40.
• the upper section 46 defines a top edge 46b of the nozzle 37.
• the first end 43 of the sleeve 39 includes a radially enlarged annular portion 48.
• the radially enlarged annular portion 48 is radially thicker than the intermediate portion 38 of the conventional thermal sleeve 26.
• the radially enlarged annular portion 48 has an outer circumferential surface 48a that is radially further away from the vertical axis A-A’ than an outer circumferential 38a of the intermediate portion 38.
• the radially enlarged annular portion 48 is supported by a support section 44 of the radially enlarged annular portion 40 of the nozzle 37. More specifically, the radially enlarged annular portion 48 includes a lower surface 48b that rests vertically on the support surface 44a. Over time, due to vibrations experienced by the conventional thermal sleeve 39, failure can occur at the radially enlarged portion 48.
• the second end 34 of the sleeve 39 extends vertically downward further than the inner end 45 of the nozzle 37 away from the vessel top head 14.
• the second end of sleeve 39 is formed by a funnel 50 that is fixed to the intermediate portion 38.
• the funnel 50 includes a cylindrical section 50a which is fixed to the outer circumferential surface 38a of the intermediate portion and a frustoconical section 50b extending downward from the cylindrical section 50a.
• the frustoconical section 50b enlarges radially as it extends downward vertically away from the intermediate portion 38.
• the inner end 45 of the nozzle 37 is substantially cylindrically shaped and surrounds a section of intermediate portion 38 of the sleeve 39.
• the second end 34 of the thermal sleeve 39 is in register with the upper end of the corresponding control rod guide tube 30.
• the drive shaft extends inside the thermal sleeve 39 as well as into the control rod guide tube 30.
• Figures 3 to 5 show a new thermal sleeve 60 according to the invention.
• the new thermal sleeve 60 comprises a tube 62, a plurality of protrusions 64, a collar 66 and a funnel 50.
• the tube 62 is advantageously composed of stainless steel type 304.
• the tube 62 comprises an upper end 68 and a lower end, not represented on the figures.
• the upper end 68 is positioned above the lower end with respect to the vertical axis
• the upper end 68 is positioned outside of the vessel shell 8 and the lower end is positioned in the interior of the vessel shell 8.
• the tube 62 also comprises an outer surface 70 and an inner surface 72.
• the outer surface 70 extends from the upper end 68 to the lower end of the tube 62, on the exterior of the tube 62.
• the inner surface 72 extends from the upper end 68 to the lower end of the tube 62, on the inner of the tube 62.
• Each protrusion 64 is disposed at the upper end 68 of the tube 62 and extends radially outward the tube 68.
• the protrusion 64 rest temporarily on the support surface 44 of the nozzle 37 to hold the new thermal sleeve 60 in position in the nozzle 37, as it will be explained below.
• Each protrusion 64 is away from the nozzle 37, and in particular from the support surface 44, during the operation of the nuclear reactor 2.
• Each protrusion 64 is obtained by plastic deformation so that each protrusion 64 moves from an initial position in which each protrusion 64 extends along the vertical axis A-A’ to a final position in which each protrusion 64 extends radially outward the tube 62.
• each protrusion 64 extends in a direction forming an angle with the vertical axis A-A’ comprised between 20 0 and 90 °, in partiular 45°.
• the outer diameter formed by the protrusions 64 advantageously more than 20 % greater than the outer diameter of the tube 62, in particular more than 50 % greater than the outer diameter of the tube 62.
• the protrusions 64 are circumferentially spaced and separated from each other by circumferentially spaced slots 74.
• the new thermal sleeve 60 comprises three protrusions 64 circumferentially spaced and separated by three slots 74.
• Each slot 74 extends on a length comprised between 30 mm and 60 mm along the vertical axis A-A’.
• the collar 66 protrudes radially outward the outer surface 70 of the tube 62 and extending circumferentially all around the tube 62, away from the upper end 68.
• the collar 66 has a hydraulic valve function in the sense that the new thermal sleeve 60 is able to lift in the nozzle 37 to allow fluid to flow from the inside of the vessel shell 8 to the inside of the control rod drive mechanisms 32 in the event of a rapid lowering of the control rod assembly in the core 6.
• the collar 66 is able to be moved from the lower section 42 of the nozzle 37 to the upper section 46 of the nozzle 37 to free the passage of the fluid towards the mechanisms 32.
• the collar 66 is disposed at a distance lower than 100 mm from the upper end 68 of the tube 62, in particular at a distance comprised between 30 mm and 50 mm. The distance between the upper end 68 and the collar 66 is measured between the lower end of the slots 74 and the upper end of the collar 66.
• the collar 66 protrudes radially from the outer surface 70 over a length comprised between 1 mm and 3 mm.
• the collar 66 defines with the nozzle 37 a gap greater than 0,1 mm.
• the gap is comprised between 0,1 mm and 1 mm.
• the collar 66 extends along the vertical axis A-A’ on a length comprised between 15 mm and 50 mm.
• the collar 66 is manufactured by machining the tube 62, as illustrated in figure 4.
• the tube 62 initially presents a larger diameter than the final piece shown in figure 4.
• the tube 62 is then cut into the desired final shape and sized by a controlled material-removal process.
• the tube 62 is manufactured before the insertion of the new thermal sleeve 60 in the nozzle 37.
• the inner surface 72 is, in the first embodiment, straight in a horizontal section as shown in figure 4 and the tube 62 presents a larger thickness at the level of the collar 66 than on the rest of the tube 62.
• the collar 66 is manufactured by a plastic expansion of the tube 62, as illustrated in figure 5.
• the inner surface 72 and the outer surface 70 are, initially, straight in a horizontal section.
• the tube 62 is then expanded by any process existing to radially expand the diameter of a tube such as, but not limited by, hydroforming, expansion mechanical tool etc, either on site or off site.
• the tube 62 is in particular expanded when the tube 62 is inserted in the nozzle 37.
• the inner surface 72 is, in the second embodiment, distorted at the level of the collar 66, as shown in figure 5.
• the tube 62 presents an equal or lower thickness at the level of the collar 66 than on the rest of the tube 62.
• the funnel 50 has a frustoconical portion fixed at the lower end of the tube 62.
• the funnel 50 of the new thermal sleeve 60 is similar to the funnel 50 of the conventional thermal sleeve 39 and will not be described again.
• the nuclear reactor 2 is stopped to perform maintenance operations.
• the vessel top head 14 is removed from the vessel shell 8 and placed on a head stand to perform the thermal sleeve replacement so that it is easy to access the inner end 32.
• the radially enlarged portion 48 of the conventional thermal sleeve 39 is broken apart such that the conventional sleeve 39 can be pulled downward through the nozzle 37 out of the inner end 45.
• the method comprises a step of machining the tube 62 to manufacture the collar 66.
• the new thermal sleeve 60 is then inserted into the nozzle 37 from the inner end 45 while on the head stand, the protrusions 64 being in the initial position, extending along the vertical axis A-A’.
• the new thermal sleeve 60 is slipped upwardly in the nozzle 37 until the protrusions 64 reach the support section 44 of the nozzle 37.
• the method comprises then a step of plastic deformation of the protrusions 64 so that each protrusion 64 extends radially outward the tube 62, as illustrated in figures 3 to 5.
• the protrusion 64 rest temporarily on the support surface 44 of the nozzle 37 to hold the new thermal sleeve 60 in position in the nozzle 37.
• the method comprises a step of plastic expansion of the tube 62 to manufacture the collar 66 while the tube 62 is inserted in the nozzle 37.
• the expansion of the tube 62 is done with an expansion tool received inside the tube 62.
• the vessel top head 14 is placed above the vessel shell 16 and the nuclear reactor 2 can be started again.
• the protrusions 64 rest on support surface 44 when the reactor vessel head 14 is not placed above the vessel shell 16.
• the thermal sleeves 60 are supported by the associated control guide tubes 30 and the protrusions 64 are away from support surfaces 44.
• the method comprises of attaching a spacer to the upper part of associated guide tubes 30 for maintaining, in operation, a minimal spacing between the thermal sleeve 60 and its associated guide tube 30.
• the new thermal sleeve 60 according to the invention provides a better thermal protection than the replacement thermal sleeves described in the state of the art.
• the collar 66 provides the hydraulic valve function and enables a better protection of the control rod drive mechanism nozzle form thermal shocks.
• the rigidity and the fatigue resistance of the new thermal sleeve 60 is reinforced thanks to the protrusions 64 being apart the nozzle 27 in operation of the reactor 2.
• the spacer provided at the lower end of the new thermal sleeve 60 limits the downward movement of the new thermal sleeve 60 relative to the nozzle 37, thus avoiding wear between the new thermal sleeve 60 and the nozzle 37. The risk that the protrusions 64 get damaged during the operating of the nuclear reactor 2 is then eliminated.
• the new thermal sleeve 60 is easy to manufacture.
• the manufacturing of the collar 66 by machining the tube 62 or by plastic deformation of the tube 62 is particularly simple.
• the method for replacing the damaged thermal sleeve 24 by the new thermal sleeve 60 according to the invention is easy and quick to implement.

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• Physics & Mathematics (AREA)
• Engineering & Computer Science (AREA)
• Plasma & Fusion (AREA)
• General Engineering & Computer Science (AREA)
• High Energy & Nuclear Physics (AREA)
• Monitoring And Testing Of Nuclear Reactors (AREA)

Applications Claiming Priority (1)

Publications (1)

Family

ID=71670301

Family Applications (1)

Country Status (2)

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• 2020
  • 2020-01-28 WO PCT/IB2020/000435 patent/WO2021152342A1/en unknown
  • 2020-01-28 EP EP20742837.6A patent/EP4097745A1/de active Pending

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