An apparatus and method for system safety analysis evaluation is provided, the apparatus including processing circuitry configured for generating a calculation matrix for a system, generating a plurality of models based on the calculation matrix, performing a benchmarking or convolution analysis of the plurality of models, identifying a design envelope based on the benchmarking or convolution analysis, deriving uncertainty models from the benchmarking or convolution analysis, deriving an assessment judgment based on the uncertainty models and acceptance criteria, defining one or more limiting scenarios based on the design envelope, and determining a safety margin in at least one figure-of-merit for the system based on the design envelope and the acceptance criteria.

A nuclear power plant includes: a nuclear reactor; a steam turbine that is driven by main steam that has been generated by the nuclear reactor; and a gland steam supply device that supplies gland steam to a gland of the steam turbine. The gland steam supply device includes a gland steam generator that heats external water, and generates the gland steam. The external water is independent of the main steam and condensate of the main steam.

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.

Arrangement for supporting U-bend tube sections in the high heat environment of steam generators using flat bars. The invention uses a combination of thicker and thinner flat bars to impart a serpentine path to the arc of the normally curvilinear U-tubes. The support system accommodates the dilation and contraction of coolant tubes and other elements caused by the extreme and varying conditions inside a steam generator, and which can cause gaps between coolant tubes and prior art tube support bars. Bars of alternating thickness provide alternating offsets to tensionally push and support each tube on multiple sides and in multiple locations, and this tension keeps the tubes in contact with at least some flat bars on multiple sides regardless of size and shape changes. Support arrangement includes a set of fan bars, each fan bar including thick and thin flat bars projecting up and out from a collector bar.

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 reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU.

Nuclear Aircraft Transportation System “KARAVAN” with its components is represented by a group of inventions in the technical and organizational relations. The main and basic invention is Nuclear Aircraft Transportation System “KARAVAN” (NATS). This invention includes two other ones: Aircraft Thrust Nuclear Power Plant, (ATNPP), which in turn includes—Thermal Power Cycle of ATNPP, (TPC ATNPP). In addition, the represented group of inventions is made up of two more inventions: Maintenance System of ATNPP, (MS ATNPP) and Emergency Response System of NATSK, (ERS NATSK).

The concept of practical implementation of the presented group of inventions involves the fact that ATNPP, which is a large unmanned drone aircraft “Tiagach”, supplies the aero-train composed of a number of passenger liners and cargo transport planes using electric motors with traction electric energy in the air.

The power supply of such an aero-train is based on the onboard Nuclear Power Plant of the aircraft “Tiagach”. In this case, the transmission of electric power to the towed electric aircraft of the aero-train is carried out by means of electric split feeders and cables, connecting and disconnecting of which between airplanes of the aero-train is carried out in the air, by analogy with refueling of airplanes in the air with JP fuel.

During the flight of the aero-train on a logistically optimized route, electric airplanes can detach from and attach to the aero-train, taking off and landing along the flight route of the aero-train using their own electric accumulators. In addition, extra ATNPP may be included in the aero-train during its flight, if it is necessary to increase the thrust. At the same time, due to the use of nuclear power, such ATNPP can remain in the air for a conditionally indefinite period of time.

The invention is aimed at creating cost-effective air freight and passenger traffic.

A method of disassembling a steam generator including a body portion, a water chamber, a tube plate and a plurality of heat transfer tubes, the method includes: a step of obtaining a disassembly target including the tube plate and a part of the heat transfer tubes; a step of specifying the heat transfer tube fixed to the tube plate; a step of releasing fixation between the part of the heat transfer tube and the tube plate; and a step of pulling out the part of the heat transfer tube from the through-hole, in the step of releasing the fixation, the TIG heating head is inserted from the primary region side, and in the step of releasing the fixation, the TIG heating head is moved to a plurality of streaks only in a direction from the primary region side to the secondary region side.

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.