System, layout, and process for handling fuel for a nuclear reactor

Abstract

In a method for handling spent nuclear fuel assemblies, the spent nuclear fuel assemblies are immersed in water for a relatively short period of time compared to conventional methods. The spent nuclear fuel assemblies are removed from a nuclear reactor, an inert gas is applied to the fuel assemblies, the water content of the inert gas is gradually increased as the inert gas is applied to the fuel assemblies, and the fuel assemblies are immersed in water. The fuel assemblies are immersed relatively quickly, for approximately two hours or less. This improves safety and allows the fuel assemblies to be cared for using conventional processing and handling equipment. The fuel assemblies may then be loaded into casks for long-term storage and / or disposal.

Description

Detailed Description of the Invention

[0001] [Cross - Reference to Related Applications] This application claims the benefit of U.S. Provisional Patent Application No. 63 / 066,783, entitled "FUEL HANDLING SYSTEM, LAYOUT, AND PROCESS FOR NUCLEAR REACTOR", filed on August 17, 2020. The entire content thereof is incorporated herein by reference.

[0002] [Background] The field of the present disclosure relates to systems, layouts, and processes for fuel handling for one or more nuclear reactors. The removal of sodium from irradiated sodium - fast reactor core components has conventionally been a bottleneck in the disposal of irradiated fuel sodium - reactor core components and irradiated non - fuel sodium - reactor core components.

[0003] It would be advantageous if irradiated core components could be processed significantly more effectively and rapidly to improve safety and efficiency.

[0004] [Summary] According to some embodiments, a method of storing irradiated core components includes moving the irradiated core components to a pool immersion chamber, removing primary coolant from the outside of the irradiated core components by blowing an inert gas onto the irradiated core components, applying a moist inert gas to the irradiated core components, flooding the irradiated core components with water, and immersing the irradiated core components in a pool of water.

[0005] In some examples, the step of applying a moist inert gas is achieved by gradually increasing the moisture content in the inert gas.

[0006] In some cases, the step of applying the wet inert gas includes a step of gradually increasing the water content in the inert gas to 100%. In some embodiments, the primary coolant may be sodium. Sodium may be present on the irradiated core component, and the sodium may be reacted with water.

[0007] The method may further include a step of loading the irradiated core component into the cask while immersed in the pool of water. In some cases, the method is carried out for a time of less than about 2 hours.

[0008] In some embodiments, the method includes a step of washing the irradiated core component with water while immersed in the pool of water.

[0009] According to some embodiments, a method of handling spent nuclear fuel includes removing a spent fuel assembly from an in-reactor storage system in a reactor vessel, transferring the spent fuel assembly to a pool immersion cell, reacting residual sodium on the spent fuel assembly with water, and immersing the spent fuel assembly in a pool of water.

[0010] The method may further include a step of storing the spent fuel assembly in the pool of water for long-term decay. In some cases, the method includes a step of loading the spent fuel assembly into a cask.

[0011] In some embodiments, the step of loading into the cask is performed in the pool of water. Optionally, the step of reacting the residual sodium is performed by passing a gas having a positive water content over the spent fuel assembly. In some cases, the gas is an inert gas, and the gas may be argon.

[0012] The method may include a step of increasing the moisture content in the gas to, for example, up to about 100%. The step of increasing the moisture content may be performed gradually and may stop at any appropriate moisture content.

[0013] In some cases, the method includes a step of forming a passivation layer over the residual sodium. The passivation layer may be formed by reacting the sodium with a reactant (e.g., water) for a certain period of time. The passivation layer may be formed by applying water to the residual sodium to form a layer of sodium hydroxide.

[0014] The method may further include a step of blowing a gas through the spent fuel assembly and measuring the flow rate of the gas passing through the spent fuel assembly. In some cases, the gas may be measured for the presence of reaction products, fission products, the primary coolant, or the content of some other property.

[0015] [Brief Description of the Drawings] FIG. 1 is a schematic diagram of a nuclear facility showing a reactor building, a fuel handling facility, and a fuel storage facility according to some embodiments.

[0016] FIG. 2 is a perspective view of a schematic diagram of a nuclear facility according to some embodiments.

[0017] FIG. 3 is a schematic diagram of a nuclear facility having a plurality of reactor buildings, fuel handling, and fuel storage facilities according to some embodiments.

[0018] FIG. 4 is a schematic diagram of a nuclear facility showing two reactor buildings and a shared fuel exchange floor according to some embodiments.

[0019] FIG. 5 is a perspective view of a nuclear fuel assembly according to some embodiments.

[0020] FIG. 6 is a schematic perspective view of a core assembly inspection stand, a conditioning cell, and a jib hoist, according to some embodiments.

[0021] FIG. 7 is a schematic diagram of an EVHM attached to a rail, according to some embodiments.

[0022] FIGS. 8A and 8B are schematic diagrams of a pool immersion cell in plan view and elevation view, respectively, according to some embodiments.

[0023] FIG. 9 is a sample process diagram regarding the storage of irradiated core assemblies, according to some embodiments.

[0024] FIG. 10 is a sample process diagram regarding the storage of irradiated core assemblies, according to some embodiments.

[0025] [Detailed Description] The following detailed description provides a better understanding of the features and advantages of the invention described in this disclosure, according to the embodiments disclosed herein. This detailed description includes many specific embodiments, but these are provided merely as examples and should not be construed as limiting the scope of the invention disclosed herein.

[0026] In previous sodium removal efforts, steam inert gas was used, followed by a water flush. The assembly inlet was fitted into a semi-gas tight receptacle, and the sodium wetted component was placed into the pressure vessel. The pressure vessel was closed, and an inert gas flow was formed. Moist steam was introduced into the gas flow inlet in a controlled state. The hydrogen level in the gas flow exiting the pressure vessel was monitored, and the above control was input to increase the moisture level in the inert gas flow. When sodium reacted, the moisture level was increased until the moisture level in the gas reached 100 and there was no hydrogen. This process was usually carried out in a dedicated hot cell. The dedicated hot cell was designed to handle an increase in hydrogen pressure. Additionally, dedicated fuel handling equipment was required.

[0027] At that point, the pressure vessel was slowly flooded and a water flow was initiated using a conventional pump. The ion level in the water was monitored and exchanged as needed until the water flowing through the assembly had a low level of sodium ions. Next, the pressure vessel was drained and the processing system and assembly were dried. The cleaned assembly was removed from the pressure vessel. The processing time was approximately 18 - 24 hours for each assembly.

[0028] This process removed almost all sodium and sodium reaction products. Usually, stainless steel is used for fuel assemblies. It is generally considered beneficial to achieve high cleanliness standards regarding sodium hydroxide residues, especially in areas where stress acts.

[0029] In addition, in some cases, the irradiated assembly is sent directly to dry storage, post irradiation examination, or reprocessing. According to many regulatory standards, spent fuel disposal acceptance standards require that the reactivity of the metal be at a very low level.

[0030] According to some embodiments, by placing sodium-wetted components in a conventional spent fuel storage basin filled with water, a rapid method is described below for significantly improving efficiency, reducing the facilities and buildings required to handle irradiated assemblies, and reducing potential radiation exposure to facility workers and the public.

[0031] 〔Layout of the refueling system in the reactor building and fuel storage facility〕 Refer to FIGS. 1, 2, 3, and 4 according to some embodiments. A general layout of one or more reactor buildings 100 and a Fuel Storage Facility (FSF) 110 is shown. In some embodiments, this layout plan has the flexibility to support up to four or more reactor units in a quad arrangement. However, for the sake of efficiency, only two units, the first reactor building 100 and the second reactor building 112, are shown. In some cases, these buildings may be arranged as shown in FIG. 3 or FIG. 4. This allows for sharing of the fuel exchange facilities and also enables a larger crane envelope for the maintenance of the reactor building (RB) and the transfer of large component casks that may occur in the maintenance and fuel exchange space 302. As a result, the RB crane will be able to place large casks on the rails of the bottom-loading transfer cask for transfer to the FSF / maintenance facility 110. In some cases, the transfer hall 304 and the FSF / maintenance building 110 are separate structures, and only the fuel supply and maintenance facilities may have a separate overhead crane. On the other hand, in some embodiments, the transfer hall is part of the FSF / maintenance facility 110. The estimated fuel exchange and maintenance footprints according to some exemplary embodiments are described in Table 1 below.

[0032]

Table 1

[0033] Of course, other layouts, dimensions, configurations, and numbers of buildings are fully possible within the scope of the present disclosure. Any physical sizes, including dimensions, heights, sizes, weights, etc., are provided merely for illustration and do not limit the content of the present disclosure unless specified in the description of the appended claims.

[0034] In some cases, a residual heat removal (RHR) system 114 is provided to assist in decay heat removal, and one RHR system 114 may be associated with each reactor building 100, 112. One or more control building modules 116 may be provided for monitoring and / or operating the reactor within the reactor building. In some cases, auxiliary buildings such as a warehouse 118 space and a remote fuel storage area 120 may be provided.

[0035] As shown in FIG. 4, in some cases, the fuel handling system includes a central control facility that monitors the high-level operation of conditioning cells, a bottom loading transfer cask (BLTC) 402, an Ex-Vessel Storage Tank (EVST) 404, an Ex-vessel handling machine (EVHM) 406, a transfer adapter, a transfer station lift, an In-vessel Transfer Machine (IVTM), a rotating plug, a pool immersion cell 410, a pool cooling and cleanup system, a pool core assembly handling machine 412, and other components and systems. Local refueling control points may have a plant-based data and communication network that transmits information to a central refueling control room.

[0036] In some cases, a new core assembly arrives from the supplier and is erected in a transport container at FSF110. A jib hoist and grapple tool may be used to transfer the vertical (longitudinal) core assembly to an inspection table and then to a conditioning cell. Prior to an outage, the BLTC may be used to pick up multiple conditioned core assemblies from the conditioning cell and transfer them to the EVST in the reactor building. In some embodiments, the EVST may be a sodium EVST, an argon EVST, or may comprise another substance that can be inert. In some cases, during the outage, the EVHM may transfer the core assembly between the EVST and the reactor transfer adapter. The EVHM hoist may be used to transfer the core assembly through the transfer adapter to the cover gas area. In the cover gas area, the core assembly is passed to a transfer station lift below the fuel exchange port of the reactor. The transfer station lift may move the core assembly vertically between the cover gas area and the IVTM fuel exchange area above the top of the core. In some examples, the IVTM, in combination with the movement of the rotating plug, moves the core assembly between the transfer station lift, the core position, and the in-vessel storage (IVS) position for decay.

[0037] According to an exemplary method, a spent fuel assembly can be removed from an IVS position within a nuclear reactor by an IVTM, and the spent fuel assembly may be transferred to a transfer station lift. The transfer station lift may be used to raise a fuel core assembly or a non-fuel core assembly to a cover gas area. In the cover gas area, an EVHM hoist and grapple engage the assembly. Then, the EVHM may raise the core assembly, remove it from the nuclear reactor through a transfer adapter, and raise the core assembly into an EVHM cask on a fuel exchange floor. Once the EVHM cask is secured, the core assembly can be transferred to an EVST. The core assembly remains at the EVST during fuel exchange batch transfer.

[0038] The method may include additional steps. For example, following shutdown, the BLTC may pick up spent core assemblies and transfer them to an FSF to be packaged as waste or to be processed as spent fuel storage. In a wet cask loading method, the BLTC may transfer some or all of a plurality of core assemblies to a pool immersion cell where sodium reacts, the assembly may be immersed in water, and then the assembly may be transferred to a larger spent fuel pool. A pool handling machine may be used to move the fuel assembly to a storage rack for long-term decay (e.g., 10 to 15 years). As known to those skilled in the art, some or all of a plurality of core assemblies may ultimately be processed into a cask within a pool, such as in a typical light water reactor cask loading process. A completely dry, inerted, spent fuel cask may be transported to a site storage pad, and a waste (non-fuel) cask may be sent to a long-term waste disposal site.

[0039] [Description of a detailed fuel exchange process according to some embodiments:] 〔Arrival of a New Core Assembly (Assembly of Fuel, Control Rods, Shield, and Reflector)〕 Refer to FIG. 5. FIG. 5 shows a representative example of the core assembly 500. The new core assembly may be composed of a core assembly of drive fuel, control rods, shield, and reflector. In some cases, the core assembly 500 includes a duct 502 that defines a nozzle assembly 504 at a first end, an upper core load pad 506, and a handling socket 508 at a second end. In some cases, the core assembly 500 may have a consistent hexagonal outer configuration. Inside the core assembly 500, the fuel pin bundle assembly 510 is fixed within the duct 502. The fuel pin bundle assembly 510 may include any suitable number of fuel pins. A series of coolant inlet ports 512 allow coolant to enter the core assembly 500, flow through the duct 502, and absorb heat from the fuel pin bundle assembly 510.

[0040] In some cases, the core assembly 500 within the transport container arrives at the reactor site. Each container may accommodate four core assemblies. The transport container may be unloaded from the supplier's truck by an FSF overhead crane or forklift. The transport container may be stored in a safe area of the FSF until its inspection and conditioning are required for an outage. The transport container may be used for the protection and safety of the assembly before the outage. This is because the transport container has appropriate component supports, vibration monitors, FME protection, and environmental controls.

[0041] During initial reactor construction, the critical characteristics of the reactor and systems may be examined using full core loading with the addition of dummy core assemblies (e.g., core assemblies without fuel therein). Also, depending on the design, these dummy assemblies may be used to maintain the core configuration during subsequent refueling or maintenance operations. Since the dummy assemblies contain no fuel, only normal radioactive waste disposal management is required. Also, special core assemblies with startup neutron sources may be provided for initial reactor startup. Since these startup sources are the contents of radioactive sources, they can be handled as fuel for final disposal. All of these core assembly types may follow the normal core assembly handling procedures used for reactor refueling and disposal.

[0042] In some cases, the outage batch is estimated to be 18 months in length and to have 30 fuel assemblies and 10 to 15 control rods. In preparation for refueling outage, the transport cask containing the core assemblies may be moved to a vertical position by an uprighting mechanism so that the core assemblies stored therein can be grappled (gripped) by a top entry grapple tool. The cask may be opened under strict cleanliness control, for example, so as to avoid contamination.

[0043] Refer to FIG. 6. The jib hoist 600 and the top entry grapple tool 602 may engage each core assembly within the transport cask and transfer them to an inspection table. In some embodiments, the grapple tool 602 may have about three to four grapple fingers that engage the handling socket 508 of the core assembly for safe lifting. FIG. 6 further shows a potential layout of the jib hoist 600, the conditioning cell 604, and the inspection table 606. The emptied core assembly transport cask may be returned to the supplier for future batch reloads and is usually not part of the plant equipment.

[0044] In some cases, the inspection table 606 may be seismically qualified and may include a vertical elevator 608 for handling two (or more) core assemblies, and may be disposed within a pit 610 on the transfer hall floor near the conditioning cell 604. When an assembly is loaded into the table 606, the elevator 608 may raise the assembly past the operator for the full-length inspection process. The assembly may have recorded identification information, may be scanned for potential shipping damage, and may be verified to be clean and free of foreign objects. The identification information of the core assembly may be transferred to the plant's fuel replacement database for tracking of the core assembly. Such tracking can be maintained throughout the life of the core assembly and can also be maintained during disposal processing by the plant's radioactive inventory surveillance program. Also, a flow test may be performed on the core assembly at the inspection table as a final assembly go / no-go test.

[0045] 〔Storage and Conditioning of New Core Assemblies (Prior to Refueling Shutdown)〕 A clean and uncontaminated conditioning cell may be prepared for the core assembly, for example, by opening a floor valve and pulling or removing a floor access plug using a jib hoist 600 and a grapple tool 602. The hoist 600 transfers the floor plug to a storage location during loading. The inspected and recorded core assembly 500 may then be transferred into the conditioning cell 604 using the jib hoist 600 and the grapple tool 602. The position of each core assembly 500 within the conditioning cell may be filled by rotating a carousel to align it with the floor valve opening. When the carousel is filled and the assembly 500 is conditioned to the fuel change temperature in an inert argon environment, the valve may be closed. In some cases, the capacity of the carousel may be designed to hold one-third of the core batch (e.g., 15 assemblies in some embodiments) so that batch conditioning during intermediate BLTC transfer to the EVST is possible. The conditioning cell may not be intended to handle core assemblies with decay heat from the reactor.

[0046] According to some embodiments, conditioning starts with a feed and bleed process. By this feed and bleed process, the initial air and moisture content of the conditioning cell becomes a completely inert argon environment in a dry state. Then, the argon environment may be gradually raised to the refueling temperature of the reactor, about 400°F, by electrical heating and circulation. In some cases, other forms of heating may be utilized. For example, argon gas may be passed through a heat exchanger, or the heat generated in the reactor core may be used to heat the argon gas. The inspection, loading, and conditioning processes may be repeated until a batch (e.g., up to 45 core assemblies, or more core assemblies) is conditioned. This process may be performed prior to the refueling outage to ensure that the core assemblies are sufficiently conditioned and stored in the EVST before the reactor is shut down. Multiple reactors using the same conditioning cell and EVST may have staggered conditioning times and staggered shutdowns so that the conditioning cell and EVST can be utilized by multiple reactors.

[0047] An overhead crane may be used to install the floor isolation valve and an adapter associated with the isolation valve on the EVST. A shield plug handling cask may be fitted to the floor isolation valve such that the shield plug is withdrawn and removed, providing access to the carousel for loading. The floor isolation valve may be closed, and the shield plug may be removed and stored by the crane and the shield plug handling cask. The EVST may be put into an inactivated refueling state for receiving a new core assembly.

[0048] 〔Example: Transfer of BLTC to the EVST (before refueling outage)〕 In the case of refueling, BLTC may fit with the floor valve of the conditioning cell, and up to three or more core assemblies 500 may be sequentially gripped from the carousel and transferred into the cask. BLTC continues to transfer the core assemblies from the conditioning cell to the EVST carousel until a full refueling batch is staged for a stop. When each core assembly is placed in the EVST, the tracking database is updated, all the information necessary to establish the planned refueling stop sequence is verified, and uploaded to the refueling control unit. During the stop, EVHM may replace the new core assembly in the EVST with the spent core assembly coming from the reactor according to the established refueling sequence from the refueling control center.

[0049] In some cases, BLTC is a cask that is rail-mounted, self-standing, and seismic-qualified, with an isolation valve that translates vertically. BLTC may have heating and / or cooling capabilities to handle either new fuel core assemblies or spent fuel core assemblies. BLTC may have a centerline movement path that provides access to the conditioning cell, EVST, and / or pool immersion cell through a floor isolation valve.

[0050] 〔Example: Reactor Shutdown and Shutdown Preparation Method System〕 According to some embodiments, the reactor is prepared for refueling after shutdown. According to some embodiments, the forced flow pump may be fixed within the reactor. By natural circulation cooling, the reactor reaches a refueling temperature of about 400°F. This refueling temperature is provided as an example, and of course, other refueling temperatures may be applicable to other types of reactors. As used herein, the terms "about" and "approximately" may indicate variability with a maximum of ±5% of the combined numerical values in some examples, for example, variability with a maximum of ±2% or variability with a maximum of ±1%.

[0051] The ASME boundary flange may be removed from the fuel exchange port. Then, the fuel exchange port transfer adapter may be moved from its storage location on the fuel exchange floor to the fuel exchange port by the main hook of the RB overhead crane. The transfer adapter is installed on the fuel exchange port, and then the associated floor isolation valve on the fuel exchange floor may follow. Optionally, the transfer column assembly may be tested, deactivated, and / or heated to the fuel exchange temperature. Also, the transfer adapter may have a cooling capacity when the fuel assembly remains between the cask and the reactor. In some cases, the transfer adapter loading is supported by the civil structure of the fuel exchange floor of the reactor building rather than the reactor head. In some cases, the transfer adapter is shared among multiple reactors (e.g., two, three, four, or more reactors) by the storage position supporting an offset stop schedule.

[0052] In the EVST, the transfer cask may be fitted to the floor isolation valve of the EVST by the RB crane. The cask isolation valve and the floor isolation valve may be opened to access the EVST shield plug. The plug may be gripped by the cask hoist and raised into the cask. After raising the plug, both valves may be closed, and the EVST shield plug may be transported to a storage location within the EVST carousel or another floor structure. The floor isolation valve may be an inert EVST boundary for fuel exchange port plug transfer.

[0053] In a nuclear reactor, an inactivated shield plug transfer cask may be fitted to a transfer adapter and / or a floor valve by a crane RB. The cask isolation valve and the floor isolation valve may be opened to access the fuel exchange port plug of the nuclear reactor. The plug may be gripped by a cask hoist and lifted into the cask. Both valves may be closed, and the fuel exchange port plug may be transported to an EVST plug storage location or other location for portable glove box maintenance. Also, the shield plug cask may have a storage location on the RB fuel exchange floor and, in some cases, has seismic support.

[0054] The control rod drive line is separated from the control rod assemblies (CRAs) in the core so that the rotating plug and the IVTM can be rotated for fuel exchange. In some cases, the control rod core assembly is separated using a tool actuated through a control rod drive mechanism above the reactor head. Alternatively, a new CRDM design will allow for remote separation and lifting of the drive line so that the rotating plug and the IVTM can be moved.

[0055] The plug can be lifted from its reactor head ledge seal by a rotating plug jack so that the plug can be rotated. The power cables and control cables for the rotating plug and the IVTM may be reconnected to unlock the IVTM drive line. The IVTM may be calibrated by checking its position at up to six or more gauging locations around the outer periphery of the core.

[0056] By circulating the transfer station lift directly below the fuel exchange port, it may be confirmed that the core assembly transfer basket moves up and down between a level slightly above the sodium level of the reactor and the position of the length of the core assembly above the core. The lift may be driven by a rotating shaft coming through a fuel exchange port having a seal for the reactor head or cover gas. In some cases, the lift moves the core assembly from the side entry basket of the lift to a height where the IVTM can grip the core assembly.

[0057] In some embodiments, when it is confirmed that the IVS position of the core assembly is empty, one fuel exchange process is initiated in the process. The verified fuel exchange sequence for the movement of the core assembly may be confirmed to be within the integrated fuel exchange control system for the EVHM, rotating plug, IVTM, and EVST carousel.

[0058] 〔Transfer of Core Assembly between EVHM, Transfer Station Lift, IVTM, Reactor Core and IVS〕 Refer to FIG. 7. According to some embodiments, the EVHM 700 is a seismic - qualified cask support structure mounted on rails and handling all core assembly transfers during stops. The EVHM may be designed such that the movement of the centerline of the cask is precisely aligned with both the reactor fuel exchange port and the EVST access port. In some cases, the EVHM 700 has a grapple drive system 702, a transport device 704, and a control cab 706. The control cab 706 may include a control device for operating the transport device 704 and the grapple drive system 702.

[0059] The EVHM may further include a cask 710. In some cases, the cask 710 is shielded. The cask may have a movable bottom entry isolation valve 712 and may further have the ability to inactivate, heat, and cool new and spent core assemblies. The EVHM 700 may further include a blower 714 and / or an exhaust device 716 for circulating air or some other gas. The EVHM cask grapple may be of a common top entry design and may have three to four (or more) fingers that engage a structure on the core assembly head. The EVHM cask may have a fuel transfer control interface for interlocks between the EVHM cask, the transfer adapter floor valve, the transfer station lift, and the EVST carousel.

[0060] In use, for example during a shutdown, the EVHM 700 can mate with the EVST floor valve, individually grip up to three or more core assemblies, and raise them into its cask 710. The cask 710 can be inactivated, shielded, and / or heated. The EVST carousel may be rotated as needed so that a sequenced core assembly position is selected for fuel transfer. The EVST is isolated by closing the EVHM cask isolation valve and the EVST floor valve. The EVHM 700 may further include an indexing system 818 for receiving multiple core assemblies.

[0061] The EVHM 700 may move to and mate with the fuel transfer port transfer adapter floor valve. The transfer path into the reactor may be confirmed to be inactivated and heated to a fuel transfer state prior to any valve operation.

[0062] The gripped core assembly within the EVHM700 may have identification information that has been confirmed to be compatible with the fuel exchange sequence. If the inert fuel exchange status is equal across the transfer adapter floor valve and the cask isolation valve, both valves may be opened for the transfer path into the reactor. The EVHM700 hoist may lower the core assembly from the cask to the lift station basket in the cover gas region.

[0063] The lift station may be located within the sodium below the fuel exchange port and may lower the core assembly through the fuel exchange port thimble to a position of the core assembly length above the core. The transfer station lift may have a fuel exchange control interlock with the EVHM hoist / grapple and the IVTM. The lift station basket may have a side entry design to limit the required vertical movement of the IVTM. The lift station assembly may be vertically removable through the reactor's fuel exchange port for any necessary maintenance or replacement.

[0064] The IVTM may perform core and IVS transfers to complete the fuel exchange stop sequence (e.g., movement from the core to the IVS, from the IVS to the transfer station lift, and from the transfer station lift to the core). In some cases, depending on the transfer station design, the IVTM grapple head may be able to move horizontally (laterally) into a slotted thimble above the core assembly in the transfer station basket. In some cases, the IVTM grapple aligns with the center of the core assembly and then descends to grip the inner diameter. The IVTM grapple may rise vertically to retract the core assembly head from the basket. The IVTM grapple may then translate the gripped core assembly horizontally outside of the side entry basket. This process may be repeated by the IVTM such that the core assembly is retrieved and placed in the lift station basket.

[0065] The IVTM may perform core and IVS transfers to complete a refueling outage sequence (e.g., movement from the core to the IVS, from the IVS to the transfer station lift, and from the transfer station lift back to the core). The IVTM may be installed on the reactor's rotating plug and may have a horizontal moving pantograph that carries a grapple head with a telescoping configuration for raising and lowering the core assembly. The IVTM grapple may be positioned at the center of the core, IVS, or transfer station location by rotation of the rotating plug, rotation of the IVTM, and / or extension of the pantograph. In some cases, the IVTM grapple is a top entry design with three to four fingers that engage the inner diameter of the core assembly head. However, the IVTM grapple can be configured to have any suitable design. Also, the IVTM grapple head may have a rotational ability to orient the hexagonal outer shape of the core assembly for core insertion. The reactor's UIS may have slots that allow the IVTM to refuel in the centerline region of the core. The mapped positions of the core assemblies within the reactor may be stored in the computer control memory of the machine that enables refueling by an automated sequence. The automated sequence may be verified by an operator. The IVTM may have a refueling control interlock with the control rod drive, rotating plug, and / or transfer station lift. The in-vessel portion of the IVTM may be exchangeable by pulling the assembly into a cask on the refueling floor through a transfer adapter.

[0066] According to some embodiments, to initiate a fuel exchange sequence, the IVTM moves the spent core assembly from the core to an open IVS position and then picks up a new core assembly from the transfer station basket and places it within an available, numbered core position. The lift station may operate to raise the head of the core assembly to a gripping height in the cover gas region. The EVHM cask isolation valve and the transfer adapter floor valve may be opened and the cask hoist may lower to grip the core assembly from the transfer station lift basket. The core assembly coming from the IVS storage may have less than 1.2 kw (e.g., twice a 18-month cycle) of decay heat to be transferred in the EVHM.

[0067] The cask hoist may raise the spent core assembly into the shielded portion of the cask above the isolation valve. The sodium drip catcher may rotate under the assembly to capture dripping sodium. In some cases, a passive or active cooling configuration of the cask may be established such that the decay heat of the core assembly is removed. The identification information of the core assembly may be verified and / or recorded, for example, by a cask camera capturing the upper end of the core assembly. The cask hoist may release the spent core assembly within the cask holding position for transfer to the EVST. This process may be repeated according to the fuel exchange sequence for other core assemblies until the EVHM is filled with one or more spent fuel assemblies.

[0068] The EVHM cask isolation valve and the transfer adapter isolation valve may be closed, and the EVHM may be moved from the fuel exchange port to the EVST for unloading into the carousel of the EVST. This fuel exchange process is repeated according to the fuel exchange sequence for all core assemblies within a stop batch. In some cases, the stop batch may be 45 assemblies. However, of course, in other embodiments, any suitable number of core assemblies may be utilized. In some cases, fewer core assemblies than all of the core assemblies are exchanged at a given time. The IVTM can perform independent in-core fuel, reflector, or shield shuffling or rotation in the stop sequence in parallel with the movement of the EVHM so that core assemblies are offloaded and reloaded in the EVST.

[0069] 〔Transfer of EVHM Core Assembly to EVST〕 After the EVHM is loaded with spent core assemblies, the EVHM may separate from the fuel exchange port floor valve and move to the EVST. The EVHM cask isolation valve may mate with the EVST floor valve, and the valve may be opened. Each of the spent core assemblies within the EVHM cask may be gripped and placed in a numbered storage position in the EVST carousel. The fuel assemblies may be placed in the outer row of the carousel for maximum decay heat removal, and any non-fuel assemblies may be placed in the inner ring of the carousel. The storage position of each core assembly may be recorded and tracked by integrated fuel exchange system control. The EVHM cask grapple, hoist, and isolation valve may have a fuel exchange interlock with the EVST floor valve and carousel drive.

[0070] The EVST may be used as a short-term storage point during fuel exchange stops. After the stop, the BLTC may transfer the core assemblies to the spent fuel processing and waste disposal cycle.

[0071] 〔Example: Transfer of the BLTC Core Assembly to the FSF〕 After the reactor is shut down, the BLTC may move above the EVST and engage with the floor valve. The carousel may be rotated to the appropriate core assembly, and the cask isolation valve and the floor isolation valve may be opened. The cask hoist and grapple may be lowered into the EVST, the core assembly may be grasped, and lifted into the cask. The cask may be secured by closing the cask isolation valve and the floor valve. Next, the cask isolation valve may be raised to separate the BLTC for movement on its rails. In some cases, the BLTC may have a transfer capacity (volume) equal to the EVHM. In some cases, the transfer capacity is up to three core assemblies at a time. The BLTC may be a seismically qualified cask that operates in both the reactor building and / or the FSF and is mounted on rails. The BLTC may be used for multiple reactors and may support one, two, three, four, or more reactors in an extended layout plan.

[0072] 〔Example: Spent Fuel Storage and Wet Cask Loading Method〕 For conventional wet cask loading methods that require long times (e.g., times exceeding 2 hours, 4 hours, 6 hours, 8 hours, 10 hours, 15 hours, 18 hours, or 20 hours, etc.), a rapid method for directly transferring irradiated core assemblies into a spent fuel pool filled with water for long-term decay and wet cask loading is described. Figure 8 shows an approach for pool immersion cells and wet cask loading. There may be more efficient methods for removing sodium, but the use of this method is justified by at least two major factors. First, the core assembly components may not have materials that are susceptible to corrosive stress corrosion cracking. Second, the core assembly is going directly into a pool of water for interim storage that has high purity and tight chemical control. Further, the described method is very efficient and significantly faster than conventional methods. This form of long-term fuel storage is approved and demonstrated in light water reactors with high reliability.

[0073] The preparation for direct pool immersion depends on two major factors. The first is to blow an inert gas downward onto the core assembly when the core assembly is removed from the reactor core, or in the pool immersion cell, or both. Second, the assembly may have a wet inert gas reaction cycle in the pool immersion cell, and then complete flooding of the core assembly may be performed. After the flooding process, the core assembly may be transferred to the main part of the pool for long-term storage. This process significantly reduces the time of 18 to 24 hours that is usually taken in conventional processes. This is because the assembly does not become dry and inactivated, but instead goes directly into the pool and is not handled again by BLTC.

[0074] Refer to FIGS. 8A, 8B, and 9. According to some embodiments, the detailed pool immersion process step 900 may include some of the following.

[0075] In block 902, the core assembly may be handled in a dry state in an inert environment of EVHM, carousel, and BLTC. This will limit the residual sodium remaining after removing them from EVST.

[0076] The used core assembly may be moved, for example, by BLTC, to the initially dried and inactivated pool immersion cell 802.

[0077] BLTC may engage with the immersion cell floor valve 804 and lower the core assembly onto the core assembly elevator 806. The core assembly elevator 806 captures the core assembly 500, for example, just below the upper handling socket, by means of a clamping device 808.

[0078] After the load is transferred to the clamping device, BLTC may release its grapple and raise its hoist, and the cask valve and the floor valve may be closed.

[0079] Within the immersion cell 802, the retractable arm and nozzle 810 may move over and engage with the core assembly head, and a cell vent path to the pool water may be established through the immersion cell 802.

[0080] In block 904, an inert gas flow may be established through and / or over the assembly 500. Optionally, the flow of the inert gas may be measured to ensure that the gas flows freely through the assembly.

[0081] In block 906, the moisture content in the inert gas stream may be gradually increased to about 100% into the gas stream, thereby reacting the residual sodium.

[0082] After the moisture content reaches 100%, the flow of the gas and moisture may be stopped, and the retractable arm and nozzle assembly 810 may be retracted from the core assembly 500.

[0083] In block 908, the cell isolation valve 812 may be opened at the bottom of the cell. Thereby, flooding and covering of the assembly 500 can be achieved.

[0084] Next, the elevator 806 may lower the core assembly 500 into the movable trolley rack 814 on the bottom of the pool.

[0085] The elevator clamping device 808 may release the core assembly seated on the movable trolley rack 814 and raise it back into the cell 802.

[0086] In block 910, the movable trolley rack 814 may move the core assembly 500 into the main pool 820. In the main pool 820, a simple fuel handling machine 822 may grip the assembly 500 and transfer it into the designated top entry storage rack position.

[0087] The sodium reaction process may be repeated for the stopped batches of the fuel core assembly and non - fuel core assembly.

[0088] The core assembly may continue to decay in the fuel rack for an appropriate length of time (e.g., in some cases, 10 to 15 years) and be cooled by the pool until the decay heat requirement for the wet cask loading process is met.

[0089] As an optional step, the overhead crane may transfer the inner cask container 824 into the cask pit for wetted pool loading.

[0090] The simple fuel handling machine 822 may grip the collapsed core assemblies and transfer them to the submerged cask 824.

[0091] The process may be repeated until the inner cask 824 is full, after which the inner cask 824 may be capped and lifted to the fuel facility floor by the FSF crane.

[0092] The inner cask may be drained, dried, and / or completely inerted, for example using helium.

[0093] The FSF crane may assemble the final dry cask shielding package and place it on a transporter.

[0094] The spent fuel dry cask may be transported to the site storage pad. The waste cask may be filled with non-fuel assemblies and sent to an appropriate waste disposal site by a similar or the same method.

[0095] Figure 10 shows the flow of a sample process for pool storage and wet cask loading. In block 1002, the core assembly is transferred to the pool immersion cell. In some cases, the pool immersion cell has sufficient volume such that there is little or no risk of pressure rise due to hydrogen generation. In block 1004, sodium reacts with water, for example as described herein. In some cases, gas passes over and / or through the core assembly such that any removal of sodium within the core assembly is first facilitated. Also, the gas may be used for leak detection. For example, after the gas has passed through the core assembly, leakage may be detected by inspecting the gas for fission products. The water content in the gas may be gradually increased, for example by adding steam to the gas, such that any sodium remaining on or within the core assembly reacts and forms a passivation layer of sodium hydroxide, encapsulating any sodium remaining on or within the core assembly.

[0096] In block 1006, any damaged assemblies are immersed in the pool, canned, and fixed. In block 1008, the core assembly is immersed in a pool of water and transferred to a storage rack within the pool. This may be accomplished by using a pool handling machine.

[0097] In block 1010, the dry cask is loaded into the pool with the core assembly using a pool handling machine. In some cases, the core assembly is loaded after a residence time in the pool of about 10, 12, or 15 years.

[0098] In block 1012, the canned damaged fuel assembly is transferred for cask loading. This may be done prior to loading the core assembly into the cask.

[0099] In block 1014, the fuel core assemblies are identified for waste tracking and cask closure. In block 1016, the non-fuel core assemblies are identified for waste tracking and waste reduction and are transferred to waste casks in the pool.

[0100] In block 1018, the non-fuel assembly casks are closed in the pool. In block 1020, the casks (both dry casks or waste casks) are lifted, dried (e.g., using helium), and sealed. In block 1022, the casks are loaded onto a cask transporter.

[0101] In block 1024, the dry casks are transferred to a site pad for storage. In block 1026, the waste casks with non-fuel are sent for long-term waste disposal.

[0102] 〔Example: Maintenance Strategy for Infrequently Serviced Sodium Wetted Equipment〕 Transfer adapters may be individually removed or installed by a crane between the fuel change floor and the reactor head / rotating plug for each piece of equipment to be removed or installed. A large-opening maintenance floor valve may be placed on top of the associated transfer adapter. Multiple transfer adapters may be inerted, e.g., with argon, for equipment transfer. Large component transfer casks may be placed on the floor valve by an RB crane, a fleet crane assembly, or a temporary fleet cask carrier.

[0103] The equipment (machine) may be hoisted into the large component cask through a transfer adapter, and the cask may be handled by an RB crane, a fleet crane, or a fleet cask carrier for transportation to a maintenance / waste handling building outside the site for off-site transportation.

[0104] 〔Example: Handling of damaged fuel〕 If a damaged fuel assembly (damaged pins) is routed for disposal, it may be processed through a pool immersion cell, and then the assembly may be canned by a special pool tool and stored in the pool until the wet cask loading process is initiated. The damaged fuel assembly may have cracks in the cladding, which allow communication between the inside of the fuel pins and the external environment. In some cases, a detection system can determine whether there are fission products in the inert gas used to remove sodium from the fuel assembly. In some cases, gamma-ray spectroscopy may be used to detect neutrons emitted from the fuel assembly. This can be used to detect damaged fuel assemblies. In some cases, the inert gas is pressurized to assist in detecting leaks.

[0105] 〔Example: Additional utilization of FSF〕 The FSF may provide a working floor space for new fuel receipt and all spent fuel processing. It may also provide a crane and large cask floor space for all non-routine reactor maintenance, and components such as fuel exchange port lifts, IVTM, control rods, electromagnetic pumps, intermediate heat exchangers (IHX), and sodium traps may be removed.

[0106] In addition, additional floor space may be provided for casks, for repair and maintenance of damaged reactor components, or for radioactive decontamination operations, or for shutdown assistance. Further, the floor space may be used for mock-ups of critical reactor operation or training evolution.

[0107] Those skilled in the art will recognize that any process or method disclosed herein can be modified in many ways. The process parameters and arrangements of the steps described and / or illustrated herein are given by way of example only and can be varied as desired. For example, the steps illustrated and / or described herein may be shown or described in a particular order, but these steps need not necessarily be performed in the order shown or described.

[0108] The various exemplary methods described and / or illustrated herein can also omit one or more of the steps described or illustrated herein, or can include additional steps in addition to the disclosed steps. Further, the steps of any method disclosed herein can be combined with any one or more steps of any other method disclosed herein.

[0109] Unless otherwise specified, the terms "connected" and "coupled" (and their derivatives) as used in the specification and claims should be construed to allow for both direct and indirect (i.e., via other elements or components) connections. Further, the term "a" or "an" as used in the specification and claims should be construed to mean "at least one of". Finally, for ease of use, the terms "comprising" and "having" (and their derivatives) as used in the specification and claims are interchangeable with the word "including" and have the same meaning.

[0110] The processor may be configured using instructions for performing any one or more of the steps of any method disclosed herein.

[0111] As used herein, the term "or" is used inclusively to refer to items in the alternative and in combination.

[0112] As used herein, characters such as numbers refer to like elements.

[0113] Embodiments of the invention are shown and described herein as examples only. Those skilled in the art will recognize numerous adaptations, changes, modifications, and substitutions without departing from the scope of the disclosure. Some alternatives and combinations of the embodiments disclosed herein may be utilized without departing from the disclosure and the scope of the invention disclosed herein. Accordingly, the scope of the invention disclosed herein is to be defined only by the appended claims and their equivalents.

Brief Description of the Drawings

[0114]

Figure 1

Figure 2

Figure 3

Figure 4

Figure 5

Figure 6

Figure 7

Figure 8A

Figure 8B

Figure 9

Figure 10

Claims

1. A method for storing irradiated core components, comprising: transferring the irradiated core components in a dry state to a pool immersion cell by means of a bottom-loading transfer cask, said pool immersion cell being selectively in liquid communication with a storage pool; removing primary coolant from the outside of the irradiated core components by blowing an inert gas through the irradiated core components; applying a moist inert gas to the irradiated core components; opening a cell isolation valve in the floor of the pool immersion cell and flooding the pool immersion cell and the irradiated core components with water from the storage pool; lowering the irradiated core components through the cell isolation valve into the storage pool, said storage pool having a pool of water; A method comprising the above steps.

2. The method according to claim 1, wherein the step of applying a moist inert gas is achieved by gradually increasing the moisture content in the inert gas.

3. The method according to claim 2, wherein the step of applying a moist inert gas includes gradually increasing the moisture content in the inert gas to 100%.

4. The method according to claim 1, wherein the primary coolant is sodium.

5. Sodium is present on the irradiated core components, The method according to claim 4, further comprising reacting the sodium with water.

6. The method according to claim 1, further comprising loading the irradiated core components into a waste cask while the irradiated core components are immersed in the pool of water.

7. The method according to claim 1, wherein the method is carried out in a time of less than about 2 hours.

8. The method according to claim 1, further comprising washing the irradiated core components with water while the irradiated core components are immersed in the pool of water.

9. A method for handling spent nuclear fuel, comprising: removing a spent fuel assembly from an in-reactor storage system within a reactor vessel; transferring the spent fuel assembly in a dry state to a pool immersion cell by means of a bottom-loading transfer cask, said pool immersion cell being selectively in liquid communication with a storage pool; reacting residual sodium on the spent fuel assembly with water; Opening the cell isolation valve on the floor of the pool immersion cell and flooding the pool immersion cell with water from the storage pool; Lowering the spent fuel assembly through the cell isolation valve into the storage pool, the storage pool having a pool of water; A method comprising:

10. The method according to claim 9, further comprising storing the spent fuel assembly in the pool of water for long-term decay.

11. The method according to claim 9, further comprising loading the spent fuel assembly into a waste cask.

12. The method according to claim 11, wherein the step of loading into the waste cask is performed within the pool of water.

13. The method according to claim 9, wherein the step of reacting the residual sodium is performed by passing a gas having a positive moisture content over the spent fuel assembly.

14. The method according to claim 13, wherein the gas is an inert gas.

15. The method according to claim 14, wherein the inert gas is argon.

16. The method according to claim 13, further comprising increasing the moisture content in the gas.

17. The method according to claim 16, wherein the step of increasing the moisture content in the gas includes increasing the moisture to about 100%.

18. The method according to claim 9, further comprising forming a passivation layer over the residual sodium.

19. The method according to claim 18, wherein the passivation layer is formed by applying water to the residual sodium to form a layer of sodium hydroxide.

20. The method according to claim 9, further comprising blowing a gas through the spent fuel assembly and measuring the flow rate of the gas passing through the spent fuel assembly.

Images