Abstract: A heat pipe configured to remove heat from a nuclear reactor core is disclosed herein. The heat pipe can include an inner housing defining an inner volume configured to accommodate a heat source and an outer housing configured about the inner housing and the heat source. A wick can be positioned between at least a portion of the inner housing and at least a portion of the outer housing, wherein the wick can include a capillary material, and wherein the wick can define an intermediate volume between the inner housing and the outer housing. A working fluid can be positioned within the intermediate volume, wherein the working fluid can evaporate at a first end of the heat pipe and condense at a second end of the heat pipe adjacent to a heat exchanger, and wherein the wick can return condensed working fluid to the first end of the heat pipe.

Abstract: An adjustable core assembly for a nuclear reactor is disclosed herein. The adjustable core can include a plurality of reactivity control cells configured to accommodate a reactivity control rod, and a plurality of unit cells. The plurality of unit cells defines a radial dimension corresponding to an initial power output of the core. Each unit cell of the plurality of unit cells is configured to accommodate fuel configured to generate energy and a heat pipe configured to transfer thermal energy away from the core. Each unit cell of the plurality unit cells can be coupled to an adjacent unit cell in a radial direction, thereby altering the radial dimension, wherein the altered radial dimension corresponds to an adjusted power output of the core, and wherein the adjusted power output of the core is different than the initial power output of the core.

Abstract: A nuclear battery is provided. The nuclear battery comprises a radiation source layer, a first electrical insulator layer, a casing layer, a first electrode, and a second electrode. The radiation source layer comprises a composition configurable to emit beta radiation. The first electrical insulator layer is disposed over the radiation source layer. The casing layer is disposed over the first electrical insulator layer. The casing layer comprises a composition configured to inhibit traversal of beta radiation. The first electrode is in electrical communication with the radiation source layer. The second electrode is in electrical communication with the casing layer. A voltage potential is present between the first electrode and the second electrode when the radiation source layer emits beta radiation.

Abstract: A nuclear reactor is disclosed including a reactor core and a reflector assembly surrounding the reactor core. The reflector assembly includes a stationary reflector component including a graphite support structure comprising a plurality of channels defined therein and a plurality of beryllium-oxide pins positioned in the channels.

Abstract: A block style heat exchanger for a heat pipe reactor having a plurality of heat pipes extending from a reactor core. The heat exchanger includes a plurality of primary channels, each for receiving heat transferred from the core via one of the heat pipes. The primary channels extending within a block of one or more materials. The heat exchanger also includes a plurality of secondary channels defined within the block for transmitting a flow of the secondary heat transfer medium through the heat exchanger from an inlet to an outlet. The block is formed from one or both of: a plurality of plates bonded together, with each plate defining at least a portion of one or more of the plurality of primary channels and/or the plurality of secondary channels, and/or a unitary piece of material formed from an additive manufacturing process.

Abstract: A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Abstract: A device configured to measure radioactivity emitted by a plurality of radionuclides is disclosed herein. The device includes a source cage with an outer ring that defines a cylindrical volume and includes an orientation notch and a plurality of holes configured to receive a radionuclide of the plurality of radionuclides. The device also includes a frame that includes an arm and a central rod, wherein the arm is configured to be coupled to the outer ring and includes an orientation pin. The central rod can be coupled to a gamma detector and positioned within the cylindrical volume when the arm is coupled to the outer ring of source cage. The orientation notch of the source cage is configured to engage the orientation pin of the arm and, when the orientation pin engages the orientation notch, the central rod is in a predetermined location of the cylindrical volume.

Abstract: A printed circuit heat exchanger for use in a reactor includes a core formed from a stack of plates diffusion bonded together. The core has: a top face, a bottom face disposed opposite the top face, a first side face extending between the top face and the bottom face, and a second side face disposed opposite the first side face. The printed circuit heat exchanger includes: a plurality of primary channels defined in the core, each of the primary channels extending from a primary inlet defined in the first side face to a primary outlet defined in the second side face; and a plurality of secondary channels defined in the core, each of the secondary channels extending among at least some of the primary channels from a secondary inlet defined in the top face to a secondary outlet defined in the top face.

Abstract: A system for use in shutting down a nuclear reactor includes a housing that defines a region therein sealed from an ambient environment and a gate member disposed within the region in a manner such that the gate member segregates the region into a first compartment and a second compartment isolated from the first compartment. The gate member is formed from a material having a predetermined melting point. The system further includes a neutron absorbing material disposed within the first compartment and a dispersion mechanism disposed within the region. The dispersion mechanism structured to encourage the neutron absorbing material from the first compartment into the second compartment.

Abstract: In a panel that uses the gamma radiation emitted by fission to produce electrical power, a source of an electrical current is connected to a layer of the panel made of a metal with a relatively high atomic number (Z) that forms an electron emitter. The emitter layer is surrounded by an insulation layer which in turn is surrounded by a relatively low Z value layer for collecting electrons from the emitter. Another layer of insulation and an outer sheath surround the collector. The improved panel may be used for reactor power level and power distribution measurements, and for initiating, maintaining or returning molten salt or metal coolants in the liquid state.

Abstract: A detection apparatus is a flow and temperature measuring device in the form of a tube having a channel positioned parallel to a coolant flow stream and contains a drogue that is firmly attached to a mount within the channel. The mount has a load detection apparatus firmly imbedded near the flow inlet to the tube. The drogue has at least neutral buoyancy in the fluid at the minimum fluid temperature of interest. A change in the buoyancy of the drogue as a function of temperature and a corresponding change in the strain detection apparatus output in the fluid is determined by a combination of simple physics and calibration measurements. The relationship between changes in strain detection apparatus signal output and flow-rate-induced forces applied to the drogue surface are also determined using a combination of simple physics and calibration measurements. A system and method are also described.

Abstract: Methods of installing an external dashpot tube around a control rod guide tube in a nuclear reactor fuel assembly are disclosed herein. The nuclear reactor fuel assembly may include a top nozzle, a bottom nozzle, and a plurality of grids. The various methods may comprise inserting a guide tube into a skeleton of the nuclear reactor fuel assembly to a lower middle grid, the lower middle grid being second closest grid to the bottom nozzle of the plurality of grids. The various methods may also include installing an external dashpot tube over the guide tube after it has been inserted to the lower middle grid; inserting the guide tube with the installed external dashpot tube to the bottom nozzle; attaching the guide tube to the skeleton; and bulging the guide tube onto the external dashpot tube.

Abstract: The invention relates to methods for mitigating the recontamination of carbon steel surfaces in a nuclear reactor or related water-containing systems and components, which have undergone a decontamination process. The methods include conducting a passivation process of the carbon steel surfaces directly following completion of the decontamination process, prior to the system or component being returned to service. In certain embodiments, a chelating agent is used in the decontamination process and is retained following completion of the process, for use in the subsequent passivation process. The passivation process forms a passivation film that is effective to reduce recontamination of the decontaminated carbon steel surfaces.

Abstract: A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Abstract: A system for determining coolant flow rate in a nuclear reactor primary cooling loop includes a processor and a memory. The memory stores physical measurements of the mechanical components comprising the primary cooling loop. The memory also stores instructions that cause the processor to: receive pressure data from a plurality of pressure sensors in the cooling loop; calculate a model of fluid flow through the primary cooling loop based on the mechanical component measurements; compare the data from the pressure sensors with estimated data from the fluid flow model; and calculate a statistical weighting of the pressure data from the pressure sensors based on the estimated pressure data from the fluid flow model. In another system, the flow rate is determined from a combination of the estimate from the modeled fluid flow with an estimate based on a calorimetric thermal exchange calculation.

Abstract: A modular dry spent fuel canister system in which several different types of inner spent nuclear fuel canisters can be loaded into the same outer cask family. This family typically includes a storage overpack, a transfer cask, a transportation cask and support or auxiliary hardware. The various canisters can be loaded interchangeably into the different types of outer casks. The inner canisters are differentiated not by physical fuel type or dimension, but by the engineering objective or criterion that applies to the spent fuel being stored. One such objective may be for a single canister to store a large number of assemblies economically and safely. A second is a canister designed to greatly reduce the cooling time (or radioactive decay time) that must pass in order to load spent nuclear fuel for off-site storage, so as to meet the decay heat requirements and capabilities of the off-site storage system.

Abstract: A method that provides a more direct indication of peak fuel rod centerline temperature and peak fuel rod clad temperature than conventionally inferred from the power distribution by directly and continuously measuring the fuel temperatures of the fuel pellets in one or more of the hottest fuel elements in the core. The peak fuel rod clad temperature is then obtained from the maximum measured peak fuel rod centerline temperature in combination with the maximum coolant core exit temperature and the minimum coolant flow rate.

Abstract: An improvement in a nuclear fuel rod is disclosed. The improved fuel rod includes a cladding tube, a plurality of fuel pellets stacked within the cladding tube, and a liquid material filling the gap between the fuel pellets and the cladding tube. The liquid material is selected from those having a thermal conductivity higher than that of helium, a melting point lower than about 400° C., a boiling point higher than 1600° C., and which are capable of wetting both the fuel pellets and the cladding sufficient to form a protective layer over the pellets and to wick into openings that may form in the cladding.

Abstract: A method for detecting a leak in a cladding tube in a nuclear reactor is described. The method is well-suited for use in a reactor having a plurality of cladding tubes housed in a plurality of linearly arranged channels for flowing coolant past the cladding tubes. The method includes monitoring the channels for the occurrence of an increase in radiation above a selected base line indicative of the presence of at least one fission product in the coolant in at least one of the plurality of channels, and monitoring the channels for the occurrence of time dependent changes in the strength of radiation in the coolant above the base line along the length of the at least one of the plurality of channels. The leak location is calculated by triangulating the radiation readings from a fixed linear array of detectors positioned adjacent to the channels to determine the location of the strongest radiation reading and the location along the length of the channel where the increase in radiation occurred.

Abstract: The single rod ultrasonic leak detection tool will be used to detect leaking fuel rods by evaluation if there is water inside the rod. The test will be performed by moving one fuel rod between two ultrasonic transducer oriented in a pitch/catch configuration. The detection tool uses ultrasonic transducers as the primary method to detect leaking fuel rods. Two probes are preferably positioned near the fuel rod in a pitch/catch or transmitting/receiving configuration. One probe 36 sends the signal into the rod and the other probe of a set of probes receives the signal. An evaluation of the received signal is performed in various embodiments, to determine leak tightness of the fuel rod. In various aspects, two sets of probes may be utilized so that measurements can be taken on both sides of the fuel rod.