Abstract: Thorium-based fuel bundles according to one or more embodiments of the present invention are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300/600/900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt %. Reprocessing of spent fuel bundles is also avoided.

Abstract: A method of generating power using a Thorium-containing molten salt fuel is disclosed. One example includes the steps of providing a vessel containing a molten salt fuel, generating a proton beam externally to the vessel, where the externally generated proton beam is of an energy level sufficient to interact with material within a fuel rod in the vessel to produce (p, n) reactions resulting in the generation of neutrons at a first energy level. Neutrons generated within the vessel through the (p, n) reactions are utilized to produce a fission reaction which increases the heat content of the molten salt within the vessel. In the example, a heat exchanger is used to extract heat from the molten salt within the vessel and power is generated from the extracted heat.

Abstract: A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises mixing a powdered fissile material selected from the group consisting of UN and U3Si2 with an additive selected from oxidation resistant materials having a melting or softening point lower than the sintering temperature of the fissile material, pressing the mixed fissile and additive materials into a pellet, sintering the pellet to a temperature greater than the melting point of the additive. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U3Si2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U3Si2.

Abstract: Systems, methods, and devices of the various embodiments enable a Nuclear Thermionic Avalanche Cell (NTAC) to capture gamma ray photons emitted during a fission process, such as a fission process of Uranium-235 (U-235), and to breed and use a new gamma ray source to increase an overall emission flux of gamma ray photons. Various embodiments combine a fission process with the production of Co-60, thereby boosting the output flux of gamma ray photons for use by a NTAC in generating power. Various embodiments combine a fission process with the production of Co-60, a NTAC generating avalanche cell power, and a thermoelectric generator generating thermoelectric power.

Abstract: The invention relates to a multi-component cladding for a nuclear fuel rod that includes a combination of ceramic and metal components. More particularly, the invention is directed to a cladding that includes a ceramic composite having a zirconium composition deposited thereon to form a zirconium coated ceramic composite. The ceramic composite includes a ceramic matrix and a plurality of ceramic fibers. The cladding is effective to protect the contents of the cladding structure from exposure to high temperature environments during various load conditions of a nuclear reactor.

Abstract: A nuclear reactor fuel rod is a fuel rod for a light-water reactor. The nuclear reactor fuel rod includes a fuel cladding tube and an end plug, both of which are formed of a silicon carbide material. A bonding portion between the fuel cladding tube and the end plug is formed by brazing with a predetermined metal bonding material interposed, and/or by diffusion bonding. The predetermined metal bonding material has a solidus temperature of 1200° C. or higher. An outer surface of the bonding portion, and a portion of an outer surface of the fuel cladding tube and the end plug, which is adjacent to the outer surface of the bonding portion are covered by bonding-portion coating formed of a predetermined coating metal. The predetermined metal bonding material and the predetermined coating metal have an average linear expansion coefficient which is less than 10 ppm/K.

Abstract: A method of preparing a surface includes applying rough material layer to a surface of a support material. The rough material layer has a thickness in a range of 50 nm to 5 micrometers and a roughness Ra in a range of 10 nm to 1 micrometer. The method includes depositing a ceramic nitride or oxynitride over the rough material layer. In an example, a vehicle includes a structural support and a component comprising a support material defining an exterior facing surface and a ceramic nitride or oxynitride layer disposed over the support material.

Abstract: A coupling device is provided for hydraulically and mechanically coupling a fuel injector to a fuel rail of a combustion engine, the coupling device including a fuel injector cup configured to be coupled to the fuel rail and to engage a fuel inlet portion of the fuel injector, the fuel injector cup including a first part including a recess and a second part configured to be received in the recess of the first part and configured to engage the fuel inlet portion of the fuel injector.

Abstract: Systems and methods for providing and using molten salt reactors are described. While the systems can include any suitable component, in some cases, they include a graphite reactor core defining an internal space that houses one or more fuel wedges, where each wedge defines one or more fuel channels that extend from a first end to a second end of the wedge. In some cases, one or more of the fuel wedges comprise multiple wedge sections that are coupled together end to end and/or in any other suitable manner. In some cases, one or more alignment pins also extend between two sections of a fuel wedge to align the sections. In some cases, one or more seals are also disposed between two sections of a fuel wedge. Thus, in some cases, the reactor core can be relatively long (e.g., to be a pipeline reactor). Other implementations are also described.

Abstract: While the described systems can include any suitable component, in some cases, they include a graphite reactor core defining an internal space that, in some cases, houses one or more fuel wedges, where each wedge defines one or more fuel channels that extend from a first end to a second end of the wedge. In some cases, one or more of the fuel wedges comprise multiple wedge sections that are coupled together end to end and/or in any other suitable manner. In some cases, one or more alignment pins also extend between two sections of a fuel wedge to align the sections. In some cases, one or more seals are also disposed between two sections of a fuel wedge. Thus, in some cases, the reactor core can be relatively long (e.g., to be a pipeline reactor). In some cases, the reactor core is also disposed within a graphite reflector. Other implementations are described.

Abstract: A traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. In a traveling wave nuclear fission reactor, a nuclear fission reactor fuel assembly comprises a plurality of nuclear fission fuel rods that are exposed to a deflagration wave burnfront that, in turn, travels through the fuel rods. The excess reactivity is controlled by a plurality of movable neutron absorber structures that are selectively inserted into and withdrawn from the fuel assembly in order to control the excess reactivity and thus the location, speed and shape of the burnfront. Controlling location, speed and shape of the burnfront manages neutron fluence seen by fuel assembly structural materials in order to reduce risk of temperature and irradiation damage to the structural materials.

Abstract: A computerized system for modeling reactor fuel element and fuel design to determine the thermo-mechanical performance thereof includes a processor coupled to memory, the memory configuring the processor to execute a fuel element analysis and an output configured to communicate data that describes the thermo-mechanical performance of the fuel element and fuel design based on the fuel element performance analysis. The processor is configured to estimate the mechanical behavior of a fuel by creating separate variables for the open and closed porosity components, conducting a routine for the open and closed porosity components that processes the current state of the fuel and updates the current state and forces of each of the open and closed porosity components, and combining the updates for the current state and forces according to a weighting; and estimate the creep and swelling behavior of a cladding.

Abstract: Systems and methods for providing and using molten salt reactors are described. While the systems can include any suitable component, in some cases, they include a graphite reactor core defining an internal space that houses one or more fuel wedges, where each wedge defines one or more fuel channels that extend from a first end to a second end of the wedge. In some cases, one or more of the fuel wedges comprise multiple wedge sections that are coupled together end to end and/or in any other suitable manner. In some cases, one or more alignment pins also extend between two sections of a fuel wedge to align the sections. In some cases, one or more seals are also disposed between two sections of a fuel wedge. Thus, in some cases, the reactor core can be relatively long (e.g., to be a pipeline reactor). Other implementations are also described.

Abstract: An electric power converter includes a semiconductor module, an electronic component, a plurality of cooling tubes, a case, a main pressure member for pressing a stacked semiconductor section in a stacking direction, and a sub-pressure member for pressing a stacked component section in the stacking direction. The stacked semiconductor section and the stacked component section are stacked in line. A pressing force of the main pressure member is greater than a pressure pressing force of the sub-pressure member. The main pressure member is disposed at an end portion of the stacked component section far from the stacked semiconductor section. A supporting portion that supports the stacked semiconductor section from the stacked component section side is disposed in the case so as to prevent the pressing force of the main pressure member from acting on the stacked component section.

Abstract: Illustrative embodiments provide a nuclear fission reactor, that includes a reactor vessel, a nuclear fission fuel element capable of generating a gaseous fission product, a valve body defining a plenum for receiving the gaseous fission product, and a valve in operative communication with the plenum for controllably venting the gaseous fission product from the plenum.

Abstract: Articles, such as tubing or strips, which have excellent corrosion resistance to water or steam at elevated temperatures, are produced from alloys having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weight percent iron, and optionally additional alloy elements selected from the group consisting of tin, chromium, copper, vanadium, and nickel with the balance at least 97 weight percent zirconium, including impurities, where a necessary final heat treatment includes one of i) a SRA or PRXA (15-20% RXA) final heat treatment, or ii) a PRXA (80-95% RXA) or RXA final heat treatment.

Abstract: A sintered fuel pellet for a water nuclear reactor fuel rod including a peripheral wall extending along a central axis and two end faces. At least one of the end faces includes at least a first chamfer extending from the peripheral wall towards the central axis with a first non-zero slope with respect to a plane perpendicular to the central axis and a second chamfer extending from the first chamfer towards the central axis with a second non-zero slope with respect to a plane perpendicular to the central axis, wherein the first slope is different from the second slope.

Abstract: A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

Abstract: An improved grid for a nuclear reactor fuel assembly that has an egg-crate base grid as the primary support structure with each support cell of the base grid that supports a fuel rod having a lock-support sleeve that is rotatable within the support cell between a first and second orientation. In the first orientation the lock-support sleeve fits loosely within the support cell of the base grid and respectively, loosely receives the fuel rods that are loaded therein. The lock-support sleeves are then rotated to a second orientation that locks the fuel rods axially within the support cells.

Abstract: A sheathed, annular metal fuel system is described. A metal fuel pin system is described that includes an annular metal nuclear fuel alloy. A sheath may surround the metal nuclear fuel alloy, and a cladding may surround the sheath. A gas plenum may also be present. Mold arrangements and methods of fabrication of the sheathed, annular metal fuel are also described.

Abstract: Disclosed is a zirconium alloy material having high corrosion resistance regardless of thermal history during its manufacturing process. The zirconium alloy material is obtained by providing a zirconium alloy containing on the mass basis: 0.001% to 1.9% of Sn, 0.01% to 0.3% of Fe, 0.01% to 0.3% of Cr, 0.001% to 0.3% of Ni, 0.001% to 3.0% of Nb, 0.027% or less of C, 0.025% or less of N, 4.5% or less of Hf and 0.16% or less of O with the remainder being inevitable impurities and zirconium, being formed of a bulk alloy and a surface layer, in which the surface layer has a plastic strain of 3 or more or a Vickers hardness of 260 HV or more and an arithmetic mean surface roughness Ra of 0.2 ?m or less.

Abstract: A method for joint configuration of nuclear power plant fuel includes the following steps: (S1) for at least one operating unit, based on an equilibrium cycle or transition cycle reactor core design, at least one new fuel element is added to at least one operating unit; (S2) after running a combustion cycle, and on basis of the new fuel elements in step (S1), more first spent fuel elements are obtained from the at least one operating unit than are obtained from the equilibrium cycle or transition cycle reactor core design, and said first spent fuel elements are kept in reserve; (S3) for at least one new starting unit, a scheduled number of new fuel elements, as well as the first spent fuel elements obtained for reserve in step (S2), are set in the first reactor cores of at least one new starting unit.

Abstract: In a method of producing isotopes in a light water power reactor, one or more targets within the reactor may be irradiated under a neutron flux to produce one or more isotopes. The targets may be assembled into one or more fuel bundles that are to be loaded in a core of the reactor at a given outage. Power operations in the reactor irradiate the fuel bundles so as to generate desired isotopes, such as one or more radioisotopes at a desired specific activity or stable isotopes at a desired concentration.

Abstract: Disclosed embodiments include methods of assembling a vented nuclear fission fuel module. Given by way of non-limiting example and not of limitation, an illustrative method of assembling a vented nuclear fission fuel module includes receiving a nuclear fission fuel element capable of generating a gaseous fission product. A valve body is coupled to the nuclear fission fuel element, and the valve body defines a plenum therein for receiving the gaseous fission product. A valve is disposed in communication with the plenum for controllably venting the gaseous fission product from the plenum. A flexible diaphragm is coupled to the valve for moving the valve. A cap is mounted on the valve, and a manipulator extendable to the cap for manipulating the cap is received.

Abstract: The container is comprised of a side wall of which the end edges are each provided with a blocking member, of which at least one is inserted in order to close an opening. The peripheral outline of the inserted blocking member and the interior outline of the associated end edge of the side wall have generally complementary shapes on at least one portion of their respective heights, in such a way that said inserted blocking member comes to be housed on at least one portion of its height in the space defined by said end edge of the side wall and comes to bear against said end edge by adjusting lugs in recesses of complementary shape. The inclined joining surfaces form an angle from 20° to 40° with the axis of the container.

Abstract: A duct for a nuclear fuel assembly includes a tubular body and an elongated member. The tubular body has a sidewall with an inner face and an outer face and is configured to contain nuclear fuel within a fuel region. The elongated member extends from the outer face along at least a portion of the fuel region and has a contact surface configured to stabilize the duct during operation of the nuclear fuel assembly.

Abstract: Nuclear fuel assemblies include non-symmetrical fuel elements with reduced lateral dimensions on their outer lateral sides that facilitate fitting the fuel assembly into the predefined envelope size and guide tube position and pattern of a conventional nuclear reactor. Nuclear fuel assemblies alternatively comprise a mixed grid pattern that positions generally similar fuel elements in a compact arrangement that facilitates fitting of the assembly into the conventional nuclear reactor.

Abstract: A lower end plug for an annular nuclear fuel rod includes a support with a great slenderness ratio, a coolant inflow part with an inner inflow space and a plurality of inflow holes in a wall thereof so as to interact with the inflow space, coupled to a lower end of the annular nuclear fuel rod at an upper end thereof, and extending from an upper end of the support in a reverse conical shape such that the inflow space interact with an inner channel of the annular nuclear fuel rod, and a coupler extending from a lower end of the support and coupled with a handing tool of the annular nuclear fuel rod. Thus, although a bottoming phenomenon of the annular nuclear fuel rod occurs, the coolant can be supplied to the inner channel of the annular nuclear fuel rod through the inflow holes of the lower end plug.

Abstract: A device for measuring and correcting a parallelism error of an upper plug end of a nuclear fuel rod comprising mechanisms for measuring a parallelism error and for correcting said error, and a mechanism for positioning said device on the fuel rod and cooperating with a rack on which the fuel rod is stored, said correction means being arranged opposite the measuring mechanism relative to the location of the fuel rod, in order to allow a measurement of the parallelism error during correction of said error.

Abstract: In a computer-implemented method of designing a nuclear reactor of a given reactor plant, an initial, test reactor core design is generated for the given plant based on a plurality of limits input by a user. The limits include a plurality of transient licensing constraints to be satisfied for operating the given plant. The method includes selecting, from a set of automated tools, one or more automated tools to evaluate the test core design, and operating one of more of selected automated tools to output data for display to the user. The displayed data is related to a core design that satisfies the limits inclusive of the transient licensing constraints.

Abstract: A grid (13) for supporting nuclear fuel pencils (3) for a nuclear fuel assembly (1) comprising a peripheral belt (17), the peripheral belt (17) comprising on at least one of its edges (35, 37) guide fins (33) is disclosed. The edge (35, 37) of the peripheral belt (17) has between the adjacent guide fins (33) recesses (39) towards the inside of the grid (13). The invention is applicable, for example, to pressurized water reactors.

Abstract: A spacer for holding fuel rods includes cells formed by a sleeve having an upper edge and a lower edge and a number of abutment surfaces. The lower edge has a wave shape with wave peaks aligned with a respective one of the abutment surfaces, and wave valleys located between two adjacent ones of the abutment surfaces. The upper edge has a wave shape with wave peaks, which are aligned with a respective one of the abutment surfaces, and wave valleys located between two adjacent ones of the abutment surfaces. Each of the abutment surfaces extend from a respective one of the wave peaks of the upper edge to a respective one of the wave peaks of the lower edge. The sleeves abut each other in the spacer along respective connection areas to make the abutment surfaces rotatable with respect to a center point of the connection area.

Abstract: A fuel rod for a nuclear reactor includes the fuel rod having a first axial zone positioned proximate to a bottom end, a second axial zone positioned adjacent to the first axial zone in the intermediate region, and a third axial zone positioned proximate to a top end. The first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone. Also includes fuel assemblies having a plurality of fuel rods and methods of designing and manufacturing of fuel rods and fuel assemblies.

Abstract: A rod assembly for a fuel bundle of a nuclear reactor may include an upper end piece, lower end piece and a plurality of rod segments attached between the upper and lower end pieces and to each other so as to form an axial length of the rod assembly. The rod assembly may include an adaptor subassembly provided at given connection points for connecting adjacent rod segments or a given rod segment with one of the upper and lower end pieces. The connection points along the axial length of the rod assembly may be located where the rod assembly contacts a spacer in the fuel bundle. One (or more) of the rod segments may include an irradiation target therein for producing a desired isotope when a fuel bundle containing one (or more) rod assemblies is irradiated in a core of the reactor.

Abstract: Example embodiments may include a nuclear fuel rod and/or segment design using fuel element spacers. Fuel element spacers may be placed at intervals within fuel rods and/or segments in order to manipulate operating characteristics of the fuel rods and/or segments and/or decrease consequences of fretting of the fuel rod and/or segment. Example methods may include using fuel rods and/or segments having fuel element spacing elements by adjusting intervals of the spacing elements so as to affect the mechanical, neutronic, and/or thermal properties of the fuel rod and/or segment.

Abstract: Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350° C. to 750° C., and a second additive having a solubility in zirconium over the temperature range extending from 350° C. to 750° C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350° C. to 750° C. and a solubility of the second additive in the first additive over the temperature range extending from 350° C. to 750° C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350° C. to 750° C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

Abstract: A filter including a plurality of flow ducts extending in a longitudinal direction and each delimited between one pair of walls is provided. The filter includes deflectors protruding into each duct alternately from the pair of walls with overlapping in the longitudinal direction to define a zigzag shaped flow channel and provided with at least one set of holes aligned in the longitudinal direction to define through the overlapping deflectors a passage for accommodating a lower end pin of a fuel rod.

Abstract: The present invention relates to a nuclear reactor, in particular a pool-type nuclear reactor cooled with liquid metal (for example, a heavy metal such as lead or lead-bismuth eutectic) or with sodium or molten salts, having a core formed by a bundle of fuel elements and immersed in a primary fluid circulating between the core and at least one heat exchanger; the fuel elements extend along respective parallel longitudinal axes and have respective bottom active parts immersed in the primary fluid to constitute the core, and respective service parts that extend at the top from the active parts and emerge from the primary fluid; the fuel elements are mechanically supported via respective top end heads anchored to supporting structures and can be operated via handling machines.

Abstract: An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

Abstract: A channel type heterogeneous reactor core for a heavy water reactor for burnup of thorium based fuel is provided. The heterogeneous reactor core comprises at least one seed fuel channel region comprising seed fuel channels for receiving seed fuel bundles of thorium based fuel; and at least one blanket fuel channel region comprising blanket fuel channels for receiving blanket fuel bundles of thorium based fuel; wherein the seed fuel bundles have a higher percentage content of fissile fuel than the blanket fuel bundles. The seed fuel channel region and the blanket fuel channel region may be set out in a checkerboard pattern or an annular pattern within the heterogeneous reactor core. Fuel bundles for the core are also provided.

Abstract: There is provided a fuel rod assembly comprising a first component of a zirconium-based material. The first component is in contact with or is located adjacent to a second component of a material different from the zirconium-based material, e.g. a nickel-based or iron-based alloy. A coating is disposed on an outer surface of the first component, which is effective to reduce an electrochemical corrosion potential difference between the first component and the second component relative to an electrochemical corrosion potential difference between the first component and the second component without the coating.

Abstract: The invention relates to a new and unique light water reactor (LWR) nuclear fuel pellet configuration formed using tri-isotropic (TRISO) fuel particles suspended in a metal, metal alloy, or ceramic matrix. The new TRISO LWR pellet would have the same dimensions as those of the standard uranium oxide pellet allowing its use without any change to the physical configuration of the reactor vessel, core internals or fuel assemblies. TRISO type fuels have a proven capability for retaining fission products within the confinement boundary created by the coating material. This robustness is expected to reduce or eliminate fuel failure risk and cost. Replacing standard pellets with TRISO LWR fuel pellets with the same, or higher, energy density can potentially extend the operating cycles of LWRs, reduce the number of fuel assemblies replaced in each refueling, reduce the quantity of spent fuel discharged from reactors, lower operating costs, and reduce radioactive waste.

Abstract: This bottom end-piece includes nozzles for directing the flow of water of the reactor along the lower ends of the fuel rods, the nozzles being arranged at nodes of the substantially regular network of the fuel rods, and an anti-debris device which delimits water flow channels. At least some of the water flow channels are arranged at nodes of the substantially regular network. Direction nozzles are arranged at least partially in the channels in order to delimit water passages therewith, and at least one water passage includes a first section and a second section which are mutually offset radially relative to the corresponding node of the network in order to form a baffle.

Abstract: The invention relates to a multilayer tubular part (1) comprising a metal layer forming a metal tubular body (3) and two layers in ceramic matrix composite material covering the metal tubular body, wherein one of the two layers in ceramic matrix composite material covers the inner surface of the metal tubular body to form an inner tubular body (4), whilst the other of the two layers in ceramic matrix composite material covers the outer surface of the metal tubular body to form an outer tubular body (2), the metal tubular body therefore being sandwiched between the inner and outer tubular bodies. The metal tubular body is in metal or metal alloy. Finally, the metal tubular body has a mean thickness smaller than the mean thicknesses of the inner and outer tubular bodies. A said part is useful in particular for producing nuclear fuel claddings.

Abstract: The invention relates to a fuel component and a method for manufacturing of a fuel component. The fuel component is adapted to be used in fission reactors. The fuel component comprises a core consisting of a first material, and a layer consisting of a second material. The layer encloses at least partly the core. The first material comprises a fissile substance. The fuel component comprises an intermediate layer between the core and the layer. The intermediate layer has a material gradient that comprises a decrease of the concentration of the first material from the core to the layer and an increase of the concentration of the second material from the core to the layer.

Abstract: This invention relates to a method of preparing nuclear fuel including the step of depositing at least two adjacent series of layers (16, 18) around a kernel (12) of fissile material, each series comprising a layer of pyrolytic carbon (16) contiguous with a layer of silicon carbide (18) and each layer (16, 18) having a thickness of at most (10) micrometers, such that alternate layers of (16, 18) of pyrolytic carbon and silicon carbide are deposited around the kernel (12). The invention extends to a nuclear fuel element (10).

Abstract: Illustrative embodiments provide a nuclear fission reactor, a vented nuclear fission fuel module, methods therefor and a vented nuclear fission fuel module system.

Abstract: A reactor core is immersed in a liquid metal coolant in a core barrel of a liquid metal cooled reactor. The reactor core includes a plurality of fuel assemblies contained in the core barrel, a neutron absorber that absorbs a neutron in the reactor core, and a neutron moderator that moderates a neutron therein so as to control a reactivity of the reactor core. The neutron absorber and the neutron moderator constitute a mixture contained in reactivity control assemblies of the reactor core in the liquid metal coolant prior to immersion of the reactor core. The neutron moderator is composed of zirconium hydride.

Abstract: The geometric dimensions and shape of a device for removing solid particles from the cooling medium that is circulated in the primary circuit of a nuclear reactor, in particular a boiling water nuclear reactor, are such that the device can be inserted in lieu of a fuel element or fuel assembly into an empty fuel element or assembly position of the reactor core of the nuclear reactor.

Abstract: The invention concerns a method comprising measurement on a fuel channel (14) of fuel assemblies (8) for nuclear boiling water reactors. The method comprises that: the measurement is done by the use of a non-destructive inductive eddy current measurement method, the measurement is done on a fuel channel (14) which has been used at least a certain time during operation in the core of a nuclear boiling water reactor, the measurement is done when the fuel channel (14) is located in water, the measurement is done on different places on the fuel channel (14), wherein through the method at least the hydride content of the fuel channel (14) at said places is determined. The method can be used in order to find out how shadow corrosion from a neighboring control rod influences the properties of the fuel channel (14).