Abstract: A nuclear power plant comprises a pressurized water reactor (PWR) and a steam generator driving a turbine driving an electric generator. A condenser condenses steam after flowing through the turbine. Responsive to a station blackout, the nuclear power plant is electrically isolated and a bypass valve is opened to convey bypass steam flow from the steam generator to the condenser without flowing through the turbine. The thermal power output of the PWR is gradually reduced over the transition time interval. After opening, the bypass valve is gradually closed over the transition time interval. A supplemental bypass valve may also be opened responsive to the station blackout to convey supplemental bypass steam flow from the steam generator to a feedwater system supplying secondary coolant feedwater to the steam generator. The supplemental bypass steam flow does not flow through the turbine and does not flow through the condenser.

Abstract: A blocking device for preventing the actuation of an automatic depressurization system in a pressurized nuclear reactor system due to spurious signals resulting from a software failure. The blocking signal is removed when the coolant level within the core makeup tanks drop below a predetermined level.

Abstract: An environmentally-friendly airport with renewable energy sources. The environmentally-friendly airport includes a security zone encompassing the land area of the airport, and a plurality of runways across the land area. There are one or more terminals within the security zone encompassing the land area of the airport. Further, a plurality of power generator systems are disposed within the land area for generating renewable, clean energy, and an industrial zone is placed within a boundary formed by the runways.

Abstract: Example embodiments are directed to upper tie plates for debris mitigation and fuel bundles that use the upper tie plates. Example embodiment tie plates may include a plurality of debris capture elements that overlap each other so as to create debris traps for particulate debris that would fall onto the fuel bundle. Example embodiment fuel bundles may use the upper tie plates for debris mitigation.

Abstract: A drain sump of a reactor containment vessel having a containment vessel floor down below a reactor pressure vessel, and includes a heat-proof sump cover and two or more drain flow paths. The drain sump is arranged inside the containment vessel floor. The heat-proof sump cover has a thickness, and covers an upper part of the drain sump. The thickness allows a top surface of the sump cover to lie in the same surface as a top surface of the containment vessel floor. The drain flow paths pass through the sump cover in a thickness direction to flow water therethrough and solidify a molten corium therein. The molten corium is produced in the unlikely event of a severe accident.

Abstract: A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

Abstract: The method for filling water into and changing the air of a main circuit of a water-cooled nuclear reactor includes a step of placing a connection and fluid isolation device which is connected to a hot leg of each cooling loop of the main circuit so as to substantially insulate, from inside the vessel, the assembly of hot legs. The method also includes a step of injecting water through an injection circuit on at least one hot leg until each cooling loop is filled with water having changed the air from a steam generator and until the water level in the vessel reaches above the side openings of the vessel that correspond to the loops, after which the connecting device is taken out of the vessel. The connecting device is capable of using telescopic connection elements.

Abstract: A nuclear engineering plant has a containment, whose interior chamber is subdivided by a wall into a systems chamber and an operating chamber which is accessible during normal operation. The containment ensures a particularly high operational reliability, in particular also in incident situations, in which hydrogen is released in the systems chamber. For this purpose, a number of overflow openings are provided in the partition wall, the respective overflow opening is closed by a closure element of a closure apparatus which opens automatically when a trigger condition associated with the respective overflow opening is reached. Closure apparatuses are provided which open both as a function of pressure and independently of pressure. The closure apparatus furthermore has a closure element containing a bursting film or a bursting diaphragm. The closure apparatus is configured such that it frees the overflow opening automatically when a predetermined environment-side trigger temperature is reached.

Abstract: A hold-down spring unit for a top nozzle of a nuclear fuel assembly. The hold-down spring unit is coupled to the upper end of the top nozzle of the nuclear fuel assembly. The hold-down spring unit includes a first spring which provides a hold-down force upon the nuclear fuel assembly under start-up conditions or hot full power conditions of a nuclear reactor, and a second spring which provides an additional hold-down force upon the nuclear fuel assembly under start-up conditions of the nuclear reactor. The hold-down margin under start-up conditions or hot full power conditions is reduced, thus enhancing the mechanical and structural stability of the nuclear fuel assembly.

Abstract: A hold-down spring unit for a top nozzle of a nuclear fuel assembly. The hold-down spring unit is coupled to the upper end of the top nozzle of the nuclear fuel assembly. The hold-down spring unit includes a first spring which provides a hold-down force upon the nuclear fuel assembly under start-up conditions and hot full power conditions of a nuclear reactor, and a second spring which provides an additional hold-down force upon the nuclear fuel assembly under start-up conditions of the nuclear reactor. The hold-down margin under start-up conditions and hot full power conditions is reduced, thus enhancing the mechanical and structural stability of the nuclear fuel assembly.

Abstract: A method of operating a nuclear reactor is disclosed. The reactor (1) encloses a core having a plurality of fuel rods (9). Each fuel rod (9) includes a cladding and fuel pellets of a nuclear fuel. The fuel pellets are arranged in an inner space of the cladding leaving a free volume comprising an upper plenum, a lower plenum and a pellet-cladding gap. The reactor is operated at a normal power and a normal inlet sub-cooling during a normal state. The reactor is monitored for detecting a defect on the cladding of any of the fuel rods. The operation of the reactor is changed to a particular state after detecting such a defect. The particular state permits an increase of the free volume in the defect fuel rod. The reactor is operated at the particular state during a limited time period, after which the reactor is operated at the normal state.

Abstract: A neutron reactor includes a neutron shield which is disposed outside a nuclear reactor core and adapted to absorb neutrons leaking from the core. The neutron shield includes a plurality of containers each of which contains a powdered neutron absorbing material and which are stacked with one another in a vertical direction, and a cladding tube which houses the containers. The neutron absorbing material is composed of B4C powder.

Abstract: A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger.

Abstract: In one embodiment, the method includes determining the safety limit minimum critical power ratio using the operating limit minimum critical power ratio, a change-in-critical-power-ratio distribution bias and a change-in-critical-power-ratio distribution standard deviation.

Abstract: Crystal habit modifiers (CHM) are provided to ameliorate deposit-related concerns in nuclear plant systems. The principal targets for utilization of crystal habit modifiers are Pressure Water Reactor (PWR) primary-side fuel rod crud and secondary-side steam generator deposits and Boiling Water Reactor (BWR) coolant system deposits.

Abstract: A pressurized water reactor nuclear containment radiation shield which surrounds the upper portion of a pressure vessel in an ice condenser containment. The vertical walls of the neutron shield are formed in vertical sections with the lower and upper sections operable during outages, to open to promote air flow cooling along the walls in the vicinity of the vessel head.

Abstract: The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Abstract: A method to capture, retain and remove debris falling into a nuclear reactor fuel bundle assembly including a bundle of fuel rods mounted below an upper tie plate and housed in a hollow metal channel, the method includes: inserting a debris shield in the upper tie plate; maintaining the shield in the upper tie plate and above the fuel rods, and water rods, while the fuel bundle assembly is in an operating nuclear reactor core; capturing debris falling in the fuel assembly on the debris shield; after capturing the debris, removing the fuel bundle assembly with the inserted debris shield from the nuclear reactor core to a maintenance/fuel inspection pool and thereafter removing the debris shield from the upper tie plate, cleaning and then reinserting the cleaned debris shield back into the upper tie plate, and moving the fuel bundle assembly from the maintenance/fuel inspection pool back into the nuclear reactor core.

Abstract: Core debris generated during a molten reactor core in a reactor containment vessel penetrating the reactor containment vessel is configured to be caught by a core catcher located beneath the reactor containment vessel which has a main body having first stage cooling water channels and second stage surrounded by cooling fins extending radially. The number of the second stage cooling channels is larger than that of the first stage cooling channels. Cooling water is supplied from a cooling water injection opening and distributed to the first cooling water channels at a distributor. An intermediate header is formed between the first and the second cooling water channels, and the cooling water is distributed to the second cooling water channels uniformly.

Abstract: An embodiment of a nuclear power plant has: a reactor vessel containing a core; a reactor containment vessel containing the reactor vessel; and a radiation heat reflecting member disposed at a portion below the reactor vessel inside the reactor containment vessel. The radiation heat reflecting member may block radiation heat, which is emitted toward a side wall surface of the reactor containment vessel from the core that has been put in a molten state by an accident and flowed downward from the reactor vessel to be accumulated at a lower portion of the reactor containment vessel. The radiation heat reflecting member may block radiation heat, which is emitted toward a supporting structure supporting devices disposed inside the reactor containment vessel.

Abstract: A jet pump diffuser weld repair device includes a lower ring section and an upper ring section respectively sized to fit around a circumference of the diffuser on opposite sides of the weld to be repaired. The lower and upper ring sections are provided with a plurality of aligned gripper slots. A corresponding plurality of grippers are fit into the gripper slots, where at least one of the gripper slots and the grippers defines cam surfaces shaped to drive the grippers radially inward as lower and upper ring sections are drawn toward each other. A plurality of connector bolts are secured between the lower ring section and the upper ring section. Tightening of the connector bolts draws the lower and upper ring sections toward each other.

Abstract: A device and method for stabilizing a dryer assembly in a reactor pressure vessel of a nuclear reactor including a spring assembly dimensioned and positioned within the reactor pressure vessel for applying a stabilizing force to the dryer assembly relative to the reactor pressure vessel.

Abstract: A pH adjusting apparatus arranged above a water pit for refueling disposed in a reactor container that stores a nuclear reactor includes a pH adjuster, a basket 50 that contains the pH adjuster, and a basket housing container 51 that houses the basket 50 therein and has formed therein an inlet through which a solvent for dissolving or diluting the pH adjuster flows in and an overflow pipe 52 through which a pH-adjusted solution generated by dissolving or mixing the pH adjuster in the solvent flows out to the water pit for refueling. Consequently, even if it is difficult to dispose the basket 50 on a floor of the water pit for refueling, it is possible to suitably perform pH adjustment in the reactor container.

Abstract: An emergency system of nuclear facility includes a gas turbine generator 35 serving as an emergency power supply that can supply power to the nuclear facility, a separation valve 38 that can separate an interior and an exterior of a containment vessel 10 from each other at a time of an accident in the nuclear facility, a battery 39 that can supply power to the separation valve 38 at a time of loss of the power from the external power supply, and a control device 32 that can control operations of the gas turbine generator 35, the separation valve 38, and the battery 39, wherein when loss of the power from the external power supply and an accident in the nuclear facility occur, the control device 32 starts the gas turbine generator 35 and closes the separation valve 38 to which power is supplied from the battery 39.

Abstract: An apparatus for forming stress corrosion cracks comprises a heating unit which includes a conductive member and a heating coil disposed adjacent to the conductive member to generate steam pressure in the tube specimen, an end holding unit, and a control unit for controlling the heating unit and the end holding unit. The stress corrosion cracks occurring in the equipment of nuclear power plants or apparatus industries during operation can be directly formed in a tube specimen using steam pressure under conditions similar to those of the actual environment of nuclear power plants, thus increasing accuracy for analysis of properties of stress corrosion cracks which are in actuality generated, thereby improving reliability of nuclear power plants or apparatus industries and effectively assuring nondestructive testing capability, resulting in very useful industrial applicability.

Abstract: A method for attaching a core spray sparger T-box clamp for a sparger T-box in a shroud of a nuclear reactor pressure vessel assembling the anchor plate, bearing plate and saddle bracket; positioning the assembly of the anchor plate, bearing plate and saddle bracket in front of the T-box such that the saddle bracket is below a sidewall of the T-box; elevating the assembly to seat the saddle bracket against a lower surface of the sidewall and sliding a locating pin on the saddle bracket into an aperture in the sidewall; lowering a carrier plate onto an upper surface of the sidewall and attaching the carrier plate to the anchor plate, wherein a locating pin on the carrier plate slides into an aperture on the upper surface, and advancing the bearing plate to the T-box to bias a bearing plate against a cover plate of the T-box.

Abstract: A nuclear island includes at least one nuclear reactor. A turbine island includes at least a turbine building housing at least one turbine driven by steam generated by the nuclear reactor. A protected area has a perimeter protected by at least one fence. An isolation zone surrounds the protected area and includes intrusion detection devices configured to detect unauthorized approach toward the protected area. The nuclear island is disposed inside the protected area, and the turbine island is disposed outside of and spaced apart from the protected area.

Abstract: The present invention relates generally to a process for a pressurized water reactor. The pressurized water reactor includes a primary circuit and a reactor core. The process includes adding a sufficient amount of an organic compound to coolant water passing through the primary circuit of the pressurized water reactor. The organic compound includes elements of carbon and hydrogen for producing elemental carbon.

Abstract: A method for treating or preparing a fuel rod cladding tube in such a way that an influence of iron oxide deposits on its surface can be studied and assessed precisely under virtually operational conditions with as little risk as possible, includes at least partially coating the fuel rod cladding tube with an iron oxide layer by immersing it in an aqueous electrolyte medium which contains iron oxide particles. The iron oxide particles are produced by anodic oxidation of an iron-containing working electrode. A test body and a device for pretreating a fuel rod cladding tube with an electrochemical three-electrode configuration, are also provided.

Abstract: An embodiment of the present invention may reduce the level of vibration experienced by a line, such as, but not limiting of, a pipe, a cable, tubing, or the like, that is connected to at least one separate structure. For example, but not limiting of, the structure includes: a reactor pressure vessel, a sparger pipe, steam generator, a pipe, a pressure vessel, a heat exchanger, a pump, a condenser, a tank, or the like. An embodiment of the present invention may provide support and a preload to the line at a new location or may replace an existing support, such as, but not limiting of, a weld; which may alter the natural frequencies to avoid resonance from occurring when the structure(s) is excited.

Abstract: A nuclear power plant with an improved cooling system using nanoparticles in solid or fluid form to improve heat transfer and reduce corrosion is provided. The nanoparticles are delivered to a closed cooling circuit such as a CCWS. Methods for providing the nanoparticles are also provided.

Abstract: A method of determining the R-factor for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind. The R-factor is a factor that accounts for the weighted local power influence on a fuel rod. A local R-factor (Ri(z)) is determined for each fuel rod (i) in said bundle and for each of a plurality of levels (z) in an axial direction. The individual axial heat generation profile for a certain fuel rod (i) is taken into account when determining the local R-factors (Ri(z)) for said fuel rod (i). A processor is configured for automatically determining the R-factor. A computer program product, a method of determining the critical power for a bundle of fuel rods, a nuclear energy plant, and a method of operating a nuclear energy plant are also described.

Abstract: A method is provided for evaluating pellet-cladding interaction (PCI) in a nuclear core having a reactor protection system and a plurality of elongated fuel rods each having fuel surrounded by cladding with a gap therebetween. The method includes: selecting a number of core parameters to be analyzed; evaluating the selected parameters at a plurality of statepoints; generating a model of an operating space of the core based, at least in part, upon the statepoints; selecting a subset or loci of statepoints from the model wherein each of the statepoints of the loci of statepoints, when subjected to a predetermined transient, falls within the operational limits of the reactor protection system; and evaluating the loci of statepoints for PCI in response to the transient. In this manner, the potential for PCI can be accurately determined without requiring every statepoint for every fuel rod in the core to be individually analyzed.

Abstract: A zirconium alloy suitable for forming reactor components that exhibit reduced irradiation growth and improved corrosion resistance during operation of a light water reactor (LWR), for example, a boiling water reactor (BWR). During operation of the reactor, the reactor components will be exposed to a strong, and frequently asymmetrical, radiation fields sufficient to induce or accelerate corrosion of the irradiated alloy surfaces within the reactor core. Reactor components fabricated from the disclosed zirconium alloy will also tend to exhibit an improved tolerance for cold-working during fabrication of the component, thereby simplifying the fabrication of such components by reducing or eliminating subsequent thermal processing, for example, anneals, without unduly degrading the performance of the finished component.

Abstract: A nuclear reactor includes a reactor pressure vessel 12 and a core barrel 14, contained within the reactor pressure vessel 12. The reactor 10 further includes a single vertical support 16, for transmitting vertical load from the core barrel to the reactor pressure vessel. Further, lateral support means is provided at an elevation which is above that of the support 16 to provide lateral support to the core barrel 14. This arrangement will result in reduced stress arising from the temperature fluctuations.

Abstract: A system for reducing an acoustic load of a fluid flow includes a first pipe to carry the fluid flow; a standpipe connected to the first pipe at an opening in the first pipe; and a standpipe flow tripper provided in the standpipe. The flow tripper includes an edge extending through the opening into the flow on a downstream side of the opening. A method of reducing an acoustic load of a standing wave in a standpipe connected to a first pipe configured to carry a flow includes disrupting the flow in the first pipe at a downstream side of an opening in the first pipe to which the standpipe is connected.

Abstract: A feeding system for an absorber liquid containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, has a storage container for the absorber liquid and is configured for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid or corrosion of the container wall of the storage container is to be excluded. For this purpose, the storage container is connected to a pressure container via an overflow line, wherein the pressure container is filled with a motive fluid.

Abstract: Disclosed herein are zirconium-based alloys that may be fabricated to form nuclear reactor components, particularly fuel cladding tubes, that exhibit sufficient corrosion resistance and hydrogen absorption characteristics, without requiring a late stage ?+? or ?-quenching processes. The zirconium-base alloys will include between about 1.30-1.60 wt % tin; 0.0975-0.15 wt % chromium; 0.16-0.24 wt % iron; and up to about 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising at least about 0.3175 wt % of the alloy. The resulting components will exhibit a surface region having a mean precipitate sizing of between about 50 and 100 nm and a Sigma A of less than about 2×10?19 hour with the workpiece processing generally being limited to temperatures below 680° C. for extrusion and below 625° C. for all other operations, thereby simplifying the fabrication of the nuclear reactor components while providing corrosion resistance comparable with conventional alloys.

Abstract: A pH adjusting system includes a water pit for refueling 35 that is disposed in a reactor container 10, which stores a reactor, and is capable of storing cooling water and a pH adjusting apparatus 37 that is disposed above the water pit for refueling 35 and houses a pH adjuster. The pH adjusting apparatus 37 causes a pH-adjusted solution generated by dissolving or mixing the pH adjuster to flow out to the water pit for refueling 35 below the pH adjusting apparatus 37. Consequently, even when it is difficult to dispose the pH adjusting apparatus on a floor of the water pit for refueling, it is possible to suitably perform pH adjustment in the reactor container 10.

Abstract: In a reactor vessel thermal load reducing system near the surface level of a coolant, the present invention is characterized in that a heat conductive member is installed not contacting the reactor vessel wall in an area above and below the coolant liquid surface, and the heat conducting member is attached to a guard vessel, the heat conducting member being made of material of good heat conductivity.

Abstract: An apparatus for passively cooling and retaining molten core material discharged from a damaged reactor vessel during a severe accident in the nuclear plant including: a molten core material retention tank to retain molten core material; a compressed gas tank storing high-pressure inert gas; a cooling water storage tank being installed higher than the molten core material retention tank; and a mixing means. The molten core material retention tank includes an outer retention vessel having at least one coolant hole, a porous protection vessel formed at an inside of the outer retention vessel, and a gravel layer formed between the outer retention vessel and the porous protection vessel. The apparatus can be installed in a reactor cavity without changing the compartment structure of a containment building, and makes it possible to prevent a steam explosion during the cooling process for the ultrahigh-temperature molten core material and to secure the reliability of the cooling process.

Abstract: A safety injection tank, used for quickly injecting emergency core cooling water (ECCW) to a reactor vessel in the case of a cold leg large break accident (CLLBA) in a pressurized water reactor (PWR), is disclosed. The safety injection tank has a gravity-driven fluidic device configured to efficiently change the ECCW injection mode from a high flow injection mode to a low flow injection mode. The gravity-driven fluidic device includes a spring placed in the upper end of the vertical pipe, and a vertically movable water tub placed on the spring so as to be movable in a vertical direction. When ECCW contained in the pressure vessel is discharged and the water level is reduced lower than the height of the tub, the tub is moved downwards such that the lower surface thereof comes into contact with the vertical pipe and closes the high flow inlet port.

Abstract: A vibration mitigation apparatus and method of reducing vibration in a reactor component are described, which may reduce the amplitude of vibration in the reactor component. The apparatus may be provided on the reactor component to reduce the amplitude of vibration of the component as the component and apparatus are subjected to a harmonic excitation that generally may occur during reactor operations.

Abstract: A feeding system for an absorber liquid containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, has a storage container for the absorber liquid and is configured for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid or corrosion of the container wall of the storage container is to be excluded. For this purpose, the storage container is connected to a pressure container via an overflow line, wherein the pressure container is filled with a motive fluid.

Abstract: A device is provided for mitigating vibration in a component of a nuclear reactor by removing vibration energy. To reduce vibration in the component, a device operatively connected to the component and including a magnet may be actuated within a conductive cylinder. This actuation may generate one or more eddy currents providing a damping function for removing vibration energy from the component, so as to alter vibration characteristics of the component.

Abstract: A drain sump of a reactor containment vessel having a containment vessel floor down below a reactor pressure vessel, and includes a heat-proof sump cover and two or more drain flow paths. The drain sump is arranged inside the containment vessel floor. The heat-proof sump cover has a thickness, and covers an upper part of the drain sump. The thickness allows a top surface of the sump cover to lie in the same surface as a top surface of the containment vessel floor. The drain flow paths pass through the sump cover in a thickness direction to flow water therethrough and solidify a molten corium therein. The molten corium is produced in the unlikely event of a severe accident.

Abstract: A method of suppressing deposition of radionuclides on components of a nuclear power plant comprises forming a ferrite film by contacting a first chemical including iron (II) ions, a second chemical for oxidizing the iron (II) ions to iron (III) ions, and a third chemical for adjusting the pH of a processing solution containing a mixture of the first and second chemicals to be 5.5 to 9.0 with the metal member surface in a time period from a finishing stage in decontamination step of removing contaminants formed on the surface of metal member composing the nuclear power plant, and suppressing deposition of radionuclides on the metal member by the ferrite film.

Abstract: A technical installation, especially a nuclear power plant, has a number of system components that are respectively supported by a number of beams, and a number of pressurized conduits. The technical installation is designed in such a way that secondary damage occurring in the surroundings of pressurized conduits are kept particularly low even if the pressurized conduits rupture. This is achieved in that at least one of the beams is embodied in a segmented manner in an area that is expected to be affected if a pressurized conduit ruptures, adjacent segments preferably being connected to each other via screw connections.

Abstract: An embodiment of the present invention provides an apparatus for reducing the magnitude or frequency of flow-induced vibrations. An embodiment of the present invention may be integrated with at least one structure of a reactor pressure vessel of a nuclear reactor. Other embodiments of the present invention may be integrated with other systems that require a dampening of and/or frequency change in the flow-induced vibrations. For example, but not limiting of, the other pressure vessels including: a steam generator, a heat exchanger, a condenser, a boiler, or the like.

Abstract: A method of calculating and using a constraint for fuel rods is provided. The method may utilize pin nodal exposures and pin nodal powers to obtain the constraint, calculate rod average exposures and rod average powers (kW/ft) in each fuel assembly, and develop maps from the calculated rod average exposures and powers (kW/ft) to operate design, optimization, licensing, and/or monitoring applications.