A resistive heater capable of delivering heat loads on the same order as those produced by in-pile nuclear fuel experiments. The heater rod provides the energy for high-temperature steady-state testing, as well as the power needed to simulate the transient pulse in the Transient Reactor Test Loop (TRTL) system. The resistive heater includes a removable housing, two or more thermal conductors in the housing; and one or more stabilizers coupled to the two or more thermal conductors to keep the two or more thermal conductors separated to avoid shorting, wherein the two or more thermal conductors are coupled to the housing via an inert gas (e.g., Helium). The two or more thermal conductors comprise a material with substantially zero infrared spectrum (e.g., sapphire, silica, or glass).

A probabilistic method for determining an operability interval for fasteners in a nuclear power plant assembly includes providing a geometric distribution of a given initial condition of fasteners in the nuclear power plant assembly at an initial time T0; generating a plurality of random future fastener failure patterns by applying a fastener failure probability model to the geometric distribution at a given time T1>T0; inputting the plurality of random future fastener failure patterns at time T1 into a machine learning model and outputting stresses of the fasteners and displacements of the components; iterating, by a processor of a computer, the applying and inputting steps for a given range of time values T2, T3, . . . , Tx>T0 and determining a maximum future time Tmax at which a predetermined acceptable probability of the fastener failure patterns having acceptable values of the stresses of the fasteners and displacements of the components, thereby justifying the acceptability of the fasteners for continued operation of the nuclear power plant assembly; and determining the operability interval as being the maximum future time Tmax minus the initial time T0.

A probabilistic method for determining an operability interval for fasteners in a nuclear power plant assembly includes providing a geometric distribution of a given initial condition of fasteners in the nuclear power plant assembly at an initial time T0; generating a plurality of random future fastener failure patterns by applying a fastener failure probability model to the geometric distribution at a given time T1>T0; inputting the plurality of random future fastener failure patterns at time T1 into a machine learning model and outputting stresses of the fasteners and displacements of the components; iterating, by a processor of a computer, the applying and inputting steps for a given range of time values T2, T3, . . . , Tx>T0 and determining a maximum future time Tmax at which a predetermined acceptable probability of the fastener failure patterns having acceptable values of the stresses of the fasteners and displacements of the components, thereby justifying the acceptability of the fasteners for continued operation of the nuclear power plant assembly; and determining the operability interval as being the maximum future time Tmax minus the initial time T0.

An environmentally sequestered nuclear spent fuel pool in one embodiment includes sidewalls and a base slab that confine a water impoundment. The pool includes fuel racks containing spent fuel assemblies which heat the water via radioactive decay. A dual liner system enclosing the pool forms an impervious barrier providing redundant provisions for preventing leakage of contaminated pool water into the environment. An interstitial space is formed between the liners which may be maintained at sub-atmospheric pressures by a vacuum pump system that evacuates the space. By maintaining the pressure in the space at a negative pressure with corresponding boiling point less than the temperature of the pool water, any leakage through the inner-most liner into the interstitial space will vaporize and be extracted via the pump for signaling a potential leak in the liner system.

A gravity-based, non-invasive method of measuring a level of fluid in a container comprises use of at least one gravity meter located as proximate a center of mass of the fluid as possible. In a nuclear reactor system a method for monitoring the level of fluid in a nuclear reactor module, a report of a loss or gain of fluid within a cylindrical module may be generated from capturing a time series of gravity data from a first gravity meter mounted as an upper gravity meter and a second gravity meter mounted as a lower gravity meter, for example, proximate a cylindrical nuclear reactor module so as not to require any invasive conduit through, for example, a containment pressure vessel (CPV) or a reactor pressure vessel (RPV). In one embodiment, the upper and lower gravity meters are mounted on stable mounts as close to the fluid in the module as possible within a coolant pool or a structure containing cooled air. If a coolant pool of water surrounds a nuclear reactor module, the meters may be housed within a dry housing in the coolant pool such that the meters may be accessed from above the coolant pool and are located as close as possible to the reactor module and its contained mass of fluid.

An examination and test system for nuclear-grade control valve is provided and includes a hermetic first chamber, a base, a guide unit, a winder, a steel cable, and a length measurement device. The hermetic first chamber includes a second chamber for accommodating a control valve. The base is disposed outside the hermetic first chamber. The guide unit is disposed on at least one of the control valve and the base. The winder and the length measurement device are disposed on the base. The steel cable connects with a valve rod of the control valve and extends out of the hermetic first chamber to connect with the winder. The steel cable is wound on the guide unit, wound up by the winder, and thus rendered taut at any time. The length measurement device has a measurement element coupled to the steel cable and displays the displacement of the measurement element.

An examination and test system for nuclear-grade control valve is provided and includes a hermetic first chamber, a base, a guide unit, a winder, a steel cable, and a length measurement device. The hermetic first chamber includes a second chamber for accommodating a control valve. The base is disposed outside the hermetic first chamber. The guide unit is disposed on at least one of the control valve and the base. The winder and the length measurement device are disposed on the base. The steel cable connects with a valve rod of the control valve and extends out of the hermetic first chamber to connect with the winder. The steel cable is wound on the guide unit, wound up by the winder, and thus rendered taut at any time. The length measurement device has a measurement element coupled to the steel cable and displays the displacement of the measurement element.

Non-intrusive error detection techniques for control and shutdown rod position in nuclear reactors, including methods of monitoring digital rod position indication (DRPI) signals of a DRPI system of a nuclear power plant. The methods include acquiring digital rod position signals at a point between a DRPI display cabinet and a DRPI data cabinet of the DRPI system, and processing the digital rod position signals to identify variations in a signal level and a signal timing of the digital rod position signals to determine rod position errors of the DRPI system.

Non-intrusive error detection techniques for control and shutdown rod position in nuclear reactors, including methods of monitoring digital rod position indication (DRPI) signals of a DRPI system of a nuclear power plant. The methods include acquiring digital rod position signals at a point between a DRPI display cabinet and a DRPI data cabinet of the DRPI system, and processing the digital rod position signals to identify variations in a signal level and a signal timing of the digital rod position signals to determine rod position errors of the DRPI system.

While typically-collected diameter data contains information for detecting some garter springs, many garter springs may not be detected without processing the diameter data. Responsively, a method for processing the diameter data to detect the garter springs has been developed. In particular, the processing involves fitting of the diameter data to a shape, determining residual errors and using the residual errors to locate garter springs.