A sensor assembly for determining an operating characteristic of a nuclear reactor. The sensor assembly includes a solid-state lasing media doped with a fissile species and disposable within a core of the nuclear reactor and an optical fiber operably coupled to the solid-state lasing media and configured to extend out of the core of the nuclear reactor and to control system of reactor. The fissile species include one or more of uranium, plutonium, americium, or californium. A method of determining an operating characteristic of a nuclear reactor includes during operation of the nuclear reactor; receiving from the optical fiber a laser light, analyzing the laser light, and based on the analysis of the laser light, determining the operating characteristic of the nuclear reactor.

A method is for controlling a nuclear power plant comprising pressurized water nuclear reactor (3) having a reactor core producing power, a primary circuit (5) connecting the reactor core to a steam generator (9), one or more of control rods (16), which can be moved into the reactor core for controlling the power of the reactor core, an injecting device (22, 23, 24, 26, 28, 30) for injecting boric acid and/or deionized water into the primary circuit (5) for controlling the reactivity of the reactor core.

A nuclear fuel decay heat utilization system usable for space heating in one embodiment comprises a nuclear generation plant building housing a spent fuel pool containing submerged fuel assemblies which emit decay heat that heats the pool. Plural fluidly isolated but thermally coupled heat removal systems comprising pumped flow loops operate in tandem to absorb thermal energy from the heated pool water, and transfer the thermal energy through the systems in a cascading manner form one to the next to a final external heat sink outside the plant building from which the heat is rejected to the ambient environment. A programmable controller operably regulates the intake and flowrate of water from the heat sink into the heat removal systems and monitors ambient air temperature inside to building. The flowrate is regulated to maintain a preprogrammed building setpoint air temperature by increasing fuel pool water temperature to a maximum permissible limit.

A method of determining variable trip setpoints at the time of performing a safety analysis on a plant protection system includes: selecting a fixed analysis setpoint including a first analysis setpoint at which safety functions are initiated according to process variables of a power plant, and a first reaching time representing a time required to reach the first analysis setpoint; deriving a variable analysis setpoint satisfying conditions of the first fixed analysis setpoint; and determining a variable trip setpoint by reflecting uncertainty of an instrumentation and control system in relation to the variable analysis setpoint.

A self-powered in-core detector arrangement for measuring flux in a nuclear reactor core includes a first in-core detector and a second in-core detector. The first in-core detector includes a first flux detecting material, a first lead wire extending longitudinally from a first axial end of the first flux detecting material, a first insulating material surrounding outer diameters of the first flux detecting material and the first lead wire and a first sheath surrounding the first insulating material. The first sheath includes a first section surrounding the first flux detecting material and a second section surrounding the first lead wire. The first section of the first sheath has a greater outer diameter than the second section of the first sheath. The second in-core detector includes a second flux detecting material, a second lead wire extending longitudinally from a first axial end of the second flux detecting material, a second insulating material surrounding outer diameters of the second flux detecting material and the second lead wire, and a second sheath surrounding the second insulating material. The second sheath includes a first section surrounding the second flux detecting material and a second section surrounding the second lead wire. The first section of the second sheath has a greater outer diameter than the second section of the second sheath. The first section of the first sheath is axially offset from the first section of the second sheath and radially aligned with the second section of second sheath.

Disclosed herein is a method pertaining to a power distribution of a reactor core of a nuclear installation, the method being executed on a general purpose computer. The method comprises: measuring current values from a plurality of vanadium neutron detector assemblies which are disposed in the reactor core of the nuclear installation; determining a measured relative core power distribution based upon the measured current values; producing a measured core power distribution based upon the measured relative core power distribution; and verifying that the reactor is operating within the licensed core operating limits based at least in part upon the measured core power distribution. Also disclosed herein is a vanadium neutron detector assembly.

A high-precision high-fidelity real-time simulation and behavior prediction method and device for a nuclear power station is provided. The method comprises the following steps: (1) constructing a nuclear power station simulator and a physical nuclear power station based on the same design parameters; (2) operating the nuclear power station simulator and the physical nuclear power station in parallel, and obtaining predicted parameters output by the nuclear power station simulator and operation parameters of the physical nuclear power station in real time; (3) comparing the predicted parameters and the operation parameters representing the same physical quantity one by one, and correcting prediction models in the nuclear power station simulator and input parameters of the prediction models by adopting a large-scale concurrent-parallel parameter search and correction algorithm and an artificial intelligence-based mode recognition and correction algorithm until the predicted parameters reach specified precision; and (4) operating the nuclear power station simulator according to a set operation condition to obtain the predicted parameters, thereby completing a behavior prediction of a physical nuclear power station system.

A dynamic characteristic analysis method of DET and RELAP5 coupling based on a universal instrumental variable method includes steps of: constructing a DET simulation model of a discrete dynamic event tree and modifying TRIP cards of an input file by adding universal instrumental TRIP variables according to state transition types of DET simulation objects, the universal instrumental TRIP variable being variable type or logical type; setting a simulation time of the RELAP5, controlling a simulation step, and analyzing an output result file of each simulation step of the RELAP5; backtracking the RELAP5 according to state transition types of DET simulation objects. The dynamic characteristic analysis method has advantages of simplifying TRIP setting process and method of DET state transition objects in an input file of the RELAP5 required for the coupling of DET and RELAP5, reducing a modeling complexity and improving a modeling efficiency.

Provided are containment apparatus monitoring system and method. The containment apparatus monitoring system includes at least one containment apparatus configured to contain nuclear material and generate containment information including integrity information and radiation information, a converter configured to collect the containment information from the containment apparatus, and a control server configured to analyze integrity of the nuclear material and a radiation state in the containment apparatus by using the containment information collected from the converter and monitor whether the containment apparatus is abnormal by using the analyzed result.

A full-digital control rod position measurement device and a method thereof. The full-digital rod position measurement device transforms the core process of rod position measurement that is normally processed by an analog circuit or analog-to-digital mixed circuit into a digital processing. The full-digital rod position measurement device comprises an excitation power supply, an integrated interface board, and a universal signal processor. The universal signal processor processes output signals of detectors according to a preset numerical processing algorithm and outputs Gray code rod position signals. The full-digital rod position measurement device and method disclosed by the present disclosure may effectively reduce the complexity of the primary excitation circuit and the secondary measurement signal processing circuit of the detectors, and improve the operation reliability and measurement accuracy of the rod position processing equipment.