Abstract: A modular nuclear reactor system includes a lift-out, replaceable nuclear reactor core configured for replacement as a singular unit during a single lift-out event, such as rather than lifting and replacing individual fuel assemblies and/or fuel elements. The system includes a reactor vessel and a power generation system configured to convert thermal energy in a high temperature working fluid received from the reactor vessel into electrical energy. The reactor vessel includes: a vessel inlet and an adjacent vessel outlet arranged near a bottom on the vessel; a vessel receptacle configured to receive a unified core assembly; locating datums in the base of the vessel receptacle and configured to constrain a core assembly in multiple degrees of freedom; and an interstitial zone surrounding the vessel receptacle and housing a set of control or moderating drums.

Abstract: Combined cleanup and heat sink systems work with nuclear reactor coolant loops. Combined systems may join hotter and colder sections of the coolant loops in parallel with any steam generator or other extractor and provide optional heat removal between the same. Combined systems also remove impurities or debris from a fluid coolant without significant heat loss from the coolant. A cooler in the combined system may increase in capacity or be augmented in number to move between purifying cooling and major heat removal from the coolant, potentially as an emergency cooler. The cooler may be joined to the hotter and colder sections through valved flow paths depending on desired functionality. Sections of the coolant loops may be fully above the cooler, which may be above the reactor, to drive flow by gravity and enhance isolation of sections of the coolant loop.

Abstract: The fuel cartridge may include a plurality of fuel channels, a first header disposed on a first side of a fuel matrix, a second header disposed on a second side of the fuel matrix opposite to the first side, and a plurality of cooling tubes through which a working fluid flows. Each of the plurality of cooling tubes may pass through each corresponding cooling channel of the plurality of cooling channels, where each of the plurality of cooling tubes has a first end connected to the first header and a second end connected to the second header. The fuel cartridge may include an interior space for sealingly containing the fuel matrix may include a pressure boundary independent from an interior of the plurality of cooling tubes, such that the interior space is not in fluid communication with the plurality of cooling tubes.

Abstract: A cable device includes a sheath member, a number of electrical cables provided within the sheath member, and an optical fiber sensing member provided within the sheath member. The optical fiber sensing member includes a functionalized optical fiber based sensor device structured to exhibit a change in one or more optical properties in response to changes in a parameter of interest. Also, a method of sensing radiation includes introducing a source light into an optical fiber sensing member provided within a structure, wherein the optical fiber sensing member comprises a functionalized optical fiber based sensor device structured to exhibit a change in one or more optical properties in response to radiation, detecting sensing light generated in response to the source light, and determining a radiation level at a plurality of locations within the structure using the detected sensing light and a distributed sensing scheme.

Abstract: A method of operating a nuclear reactor is provided. The method includes defining a layer increment of a deposit layer modeling a deposit on a heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of the deposit layer after each layer increment is added to the deposit layer; determining a temperature related variable of the heat transfer surface as a function of the recalculated properties of the deposit layer; and altering operation of the nuclear reactor when the temperature related variable of the heat transfer surface reaches a predetermined value. A method of modeling a deposit on a heat transfer surface of a nuclear reactor is also provided.

Abstract: An optical gamma thermometer includes a metal mass having a temperature proportional to a gamma flux within a core of a nuclear reactor, and an optical fiber cable for measuring the temperature of the heated metal mass. The temperature of the heated mass may be measured by using one or more fiber grating structures and/or by using scattering techniques, such as Raman, Brillouin, and the like. The optical gamma thermometer may be used in conjunction with a conventional reactor heat balance to calibrate the local power range monitors over their useful in-service life. The optical gamma thermometer occupies much less space within the in-core instrument tube and costs much less than the conventional gamma thermometer.

Abstract: A critical heat flux prediction device, a critical heat flux prediction method, a safety evaluation system, and a core monitoring system using the safety evaluation system can predict critical heat flux in a core of a reactor with a high degree of accuracy by obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on experimental data, approximating a correlation plot distribution through a logistic function that is a model function in which critical heat flux is expressed by a function of a thermal equilibrium quality, and obtaining a critical heat flux correlation of critical heat flux and a thermal equilibrium quality.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: Exemplary embodiments provide automated nuclear fission reactors and methods for their operation. Exemplary embodiments and aspects include, without limitation, re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven surface area/volume ratio neutron absorption, low coolant temperature cores, refueling, and the like.

Abstract: A method of controlling a temperature of nonthermal nuclear fusion fuel, wherein materials of low boiling point are added as a coolant to a nonthermal nuclear fusion fuel mixed with liquid lithium (02), or nuclear fusion materials fused into liquid lithium (02), and rapid elevation in temperature of a local nonthermal fusion reaction region is suppressed by a heat vaporization of the materials of low boiling point so as to enhance the nonthermal nuclear fusion reaction. The materials of low boiling point are sodium or potassium.

Abstract: A method of modeling a heat exchanger is disclosed and comprises assigning input temperatures, assumed output temperatures, and a set of flow rates, inputting the parameters into a set of equations arranged to calculate a heat transfer coefficient, inputting parameters into a second set of equations arranged to calculate output temperatures, substituting actual output temperatures for the assumed output temperatures, and again calculating the heat transfer cooefficient. The new heat transfer coefficient is then used to obtain revised actual output temperatures, and the initial actual output temperatures and the revised actual output temperatures are compared to determine whether they differ by less than a desired variance. If not, a new iteration is performed until the output temperatures converge.

Abstract: An enhanced protection system for protecting against transient overpower in a boiling water nuclear reactor which automatically adjusts the reactor over-power protection trip setpoints to be a controlled margin above the operating power level, so that enhanced fuel and reactor protection is provided at all power levels.

Abstract: A method and a system for fine control of the thermal power of a nuclear boiling water reactor (BWR) with natural circulation. The reactor vessel, in which the core of such a reactor is enclosed, embodies an in-vessel part of a water level gauge containing a water level sensor, such that together with the out-of-vessel part of the system, which is a signal conditioning and processing unit, the effective water level in the vessel can accurately be measured and adjusted either automatically or by hand. The in-vessel part of the system comprises a water level gauge, such as a vertical standpipe, in which the "steam-collapsed" or "effective" water level is accurately created. The effective water level is typically defined for a reactor vessel of a normally operating boiling water reactor (BWR), while partly containing liquid water, partly a mixture of steam and water, and partly just steam.

Abstract: A nuclear fuel assembly for a nuclear reactor has a plurality of vertically extending fuel rods arranged side by side in a square array and containing fissile material. The array has two adjacent first sides which are next to a control rod region of the core and two adjacent second sides which are next to a non-control rod region of the core.

Abstract: An autonomous, decentralized fast breeder reactor includes a single reactor main vessel which houses a plurality of small-size reactor subsystems each having a small-scale fast breeder reactor core, and a plurality steam generator subsystems. These subsystems function in an autonomous manner and are caused to undergo a heat transfer with one another by a coolant circulating naturally through the interior of the main vessel, thereby constructing a cooperatively operating system. Steam generated by the steam generators is introduced to a turbine system and utilized in generating electricity. The condensate from the turbines is cooled by a heat accumulating pool, and the heat is utilized in a separate system. The entire system is installed underground and use is made of the difference in elevation. Use also is made of solid bedrock to construct a housing facility for the reactor.

Abstract: Heat transfer between a surface and an environment, such as a steam/air mixture in a pressure vessel, is determined by directing heat transferred to the surface to a location away from the environment by use of a heat pipe. The heat content of the working fluid of the heat pipe is determined by causing a fluid to flow through a jacket encircling the condensation part of the heat pipe and measuring the mass flow of the fluid and the temperature change. The method is particularly useful in determining heat transfer coefficients for use in safety analysis for pressurized water reactor power stations.

Abstract: In the event of an incident in the secondary cooling circuit of a nuclear reactor preventing cooling the water in the primary circuit of the steam generators (2) which are common to both circuits, an emergency stop instruction is issued in order to cause control and stop rods to fall into the core of the reactor. If this instruction is not executed, and if the reactor remains at a high power level, an instruction is issued to stop the primary pumps. A rapid increase in the temperature of the water in the core then slows down the nuclear reaction soon enough to prevent subsequent excess pressure from damaging the primary circuit. The invention is particularly applicable to electricity-generating nuclear power stations.

Abstract: To provide intrinsic reactor safety, the fuel sub-assemblies of a nuclear reactor core are provided with reactivity control mechanisms operable, in response to temperature rises above desired limits, to interact with adjacent fuel sub-assemblies and effect radial dilation of the core in order to reduce reactivity. Various embodiments are disclosed in which differential thermal expansion of an operating device or devices (34; 40; 42, 44, 46; 50) is translated into radially outward displacement of a lever or levers (28) for contact with neighbouring sub-assemblies.

Abstract: A high temperature pebble bed nuclear reactor having a medium power capacity of 300 to 500 MW.sup.e is equipped with two different shut-down arrangements comprising reflector rods used exclusively for scram. The total shut-down reactivity of the reflector rods is proportioned in order to prevent the excessive cooling of the reactor core folowing scram. In this manner, the reactor core is rendered subcritical (in the event of accidents, for example, or at the beginning of any operating state) yet capable of returning to criticality at a reduced level of temperature and power output following the removal of heat. The use of all of the reflector rods for scram is effected only in the event of reactivity accidents. For all other scram incidents, only a portion of the reflector rods are used.

Abstract: A method of operating a pressurized water nuclear reactor comprising determining the present core power and reactivity levels and predicting the change in such levels due to displacer rod movements. Groups or single clusters of displacer rods can be inserted or withdrawn based on the predicted core power and reactivity levels to change the core power level and power distribution thereby providing load follow capability, without changing control rod positions or coolant boron concentrations.

Abstract: In the method of shutting down a high temperature nuclear reactor having a negative temperature coefficient of reactivity, such as a gas cooled pebble-bed nuclear reactor, while the core is operating in the critical state with the core having a predetermined critical average core temperature and with the heat generated in the critical state being removed by a coolant, effecting the shut down by reducing or discontinuing the removal of heat from the reactor core and increasing the average core temperature by an amount above the critical average core temperature for rendering the core hot sub-critical due to the negative temperature coefficient of reactivity. The core can be maintained in the hot sub-critical state by a controlled removal of after shut-down heat from the core.

Abstract: A nuclear reactor is supplied with feed water through a feed water pump system. A primary steam flow produced from the reactor is controlled by regulating a recirculated flow of feed water. The feed water pump system comprises two main pumps each of 55%-capacity and two auxiliary pumps each of 27.5%-capacity. Normally, the two main pumps are operated. Upon occurrence of abnormal condition of at least one main pump, the auxiliary pumps are started to supply feed water. At that time, the recirculated flow is controlled for a predetermined time to a reduced rate which is smaller as compared with that of the primary steam flow decreased rapidly due to the shutdown of the main pump. Subsequently, the recirculated flow is so controlled that the primary steam flow rate is slightly smaller as compared with the feed water flow which is determined by the available capacity of the pumps.

Abstract: A method of operating a pressurized water nuclear reactor comprising determining the present core power and reactivity levels and predicting the change in such levels due to displacer rod movements. Groups or single clusters of displacer rods can be inserted or withdrawn based on the predicted core power and reactivity levels to change the core power level and power distribution thereby providing load follow capability, without changing control rod positions or coolant boron concentrations.