Abstract An apparatus and process for the accurate measurement and calibration of core flow in a boiling water reactor is disclosed. The reactor includes a pressure vessel having internal forced circulation through the core. An annular downcomer region establishes reactor coolant flow downwardly in the periphery of the reactor vessel and finally upwardly and centrally into the core across a core plate. A pressure sensor detecting the pressure differential across the core plate is utilized. This pressure sensor has its measurement enhanced by input from local power range monitors in the core to utilize both the sensed pressure differential and the power to predict more accurately flow in the reactor. An algorithm is developed for utilizing the pressure differential and the real time readings from the local power range monitor to accurately gauge overall reactor coolant flow. To ensure accurate calibration, two calibration standards are itilized at steady states of reactor power output and coolant flow. A first calibration standard includes the installation of thermocouples adjacent the reactor core plate for measurement of fluid enthalpy of the coolant as it flows upwardly through the reactor core. The determined enthalpy is utilized in an energy flow balance wherein the core flow rate is solved for as an unknown. A second calibration standard utilizes the sensed pressure differential across the annular pump deck of the forced circulation pumps. The two standards are combined in output utilizing a least squares averaging, and the result combined to calibrate the disclosed algorithm.
In a boiling water nuclear reactor a process for estimating core flow rate comprising the steps of: providing a reactor core with a nuclear reaction therein generating neutrons having a core plate fluid flow barrier at the bottom portion of said reactor core; providing local power range monitors arrayed throughout said core for the measurement of moderated neutrons for the determination of the power of the nuclear reaction in said core; providing for coolant upflow through said core plate fluid flow barrier and said core for moderating neutrons from said nuclear reaction for the critical continuation of said nuclear reaction, said upflow being at said core plate fluid flow barrier at a single liquid phase portion of said core and continuing to a two phase steam/coolant region of said core; measuring the pressure differential of said upflowing coolant across said core plate fluid flow barrier; measuring the power of said reactor at said local z power range monitors; and, determining a quadratic relation for coolant flow between said local power range monitor readout and said pressure differential cross said core plate; and, computing flow from said quadratic equation.
Claim 2
Dependent Claim
188 words
In a boiling water nuclear reactor a process for estimating core flow rate comprising the steps of: providing a reactor core with a nuclear reaction therein generating neutrons having a core plate fluid flow barrier at the bottom portion of said reactor core; providing local power range monitors arrayed throughout said core for the measurement of moderated neutrons for the determination of the power of the nuclear reaction in said core; providing for coolant upflow through said core plate fluid flow barrier and said core for moderating neutrons from said nuclear reaction for the critical continuation of said nuclear reaction, said upflow being at said core plate fluid flow barrier at a single liquid phase portion of said core and continuing to a two phase steam/coolant region of said core; measuring the pressure differential of said upflowing coolant across said core plate fluid flow barrier; measuring the power of said reactor at said local z power range monitors; and, determining a quadratic relation for coolant flow between said local power range monitor readout and said pressure differential cross said core plate; and, computing flow from said quadratic equation.
Claim 3
Dependent Claim
135 words
The process of claim 1 and including the further steps of: providing means for measuring the absolute pressure of said upflowing coolant at said core plate fluid flow barrier; providing means for measuring the temperature of said upflowing coolant adjacent said core plate fluid flow barrier for determining the temperature of said upflowing coolant; operating said reactor at a steady state of power output and coolant flow; determining the enthalpy of said upflowing coolant to said core in the single phase region of said reactor coolant adjacent said core plate utilizing said absolute pressure and temperature; and, computing the flow rate of coolant in said reactor utilizing an energy balance with said determined enthalpy; and, utilizing said determined flow rate from said energy balance at said steady state flow rate to calibrate said quadratic relation.
Claim 4
Dependent Claim
99 words
The process of claim 2 and including the further steps of: providing for the measurement of pressure differential at a single phase portion of said reactor across a barrier of flow resistance other than said core plate fluid flow barrier; and, measuring flow rate utilizing said barrier of flow resistance and said pressure differential across said barrier of fluid flow resistance at said steady state of power and flow; and, utilizing said determined flow rate from said barrier of flow resistance with said flow rate from said energy balance in an average value to calibrate said determined quadratic relation.
Claim 5
Dependent Claim
34 words
The process of claim 3 and including the step of: utilizing the placement of internal recirculating pumps across a pump deck extending around said reactor as the site of said barrier of flow resistance.