Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21D—NUCLEAR POWER PLANT
  • G21D3/00—Control of nuclear power plant
  • G21D3/001—Computer implemented control
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
  • Y02E30/00—Energy generation of nuclear origin

Definitions

• This invention relates generally to nuclear reactors and more particularly to methods for increasing thermal power output of boiling water reactors.
• a typical boiling water reactor includes a pressure vessel containing a nuclear fuel core immersed in circulating coolant water which removes heat from the nuclear fuel.
• the water is boiled to generate steam for driving a steam turbine-generator for generating electric power.
• the steam is then condensed and the water is returned to the pressure vessel in a closed loop system.
• Piping circuits carry steam to the turbines and carry recirculated water or feed water back to the pressure vessel that contains the nuclear fuel.
• the BWR includes several conventional closed-loop control systems that control various individual operations of the BWR in response to demands.
• a control rod drive control system CRCS
• CRCS control rod drive control system
• RFCS recirculation flow control system
• TCS turbine control system
• monitoring parameters include core flow and flow rate affected by the RFCS, reactor system pressure, which is the pressure of the steam discharged from the pressure vessel to the turbine that can be measured at the reactor dome or at the inlet to the -NS-6034
• reactors were designed to operate at a thermal power output higher than the licensed rated thermal power level. To meet regulatory licensing guide lines, reactors are operated at a maximum thermal power output less than the maximum thermal power output the reactor is capable of achieving. These original design bases include large conservative margins factored into the design. After years of operation it has been found that nuclear reactors can be safely operated at thermal power output levels higher than originally licensed. It has also been determined that changes to operating parameters and/or equipment modifications will permit safe operation of a reactor at significantly higher maximum thermal power output (up to and above 120% of original licensed power).
• a computerized method for licensing increased power output of a boiling water nuclear reactor includes selecting generic safety evaluations from a database of generic evaluations, comparing reactor operating conditions at an increased power output with the reactor operating conditions of the selected generic evaluations, validating the applicability of the generic evaluations, and performing plant-specific evaluations at -NS-6034
• a system for licensing increased power output of a boiling water reactor includes a computer configured to simulate the operation and response of the nuclear reactor at an increased power output, select generic safety evaluations from a database of generic evaluations, compare reactor operating conditions at the increased power output with the reactor operating conditions of the selected generic evaluations, validate the applicability of the generic evaluations, and perform plant-specific safety evaluations at operating conditions outside the conditions of the selected generic evaluations and safety evaluations not included in the generic evaluations database.
• Figure 1 is a schematic diagram of the basic components of a power generating system that contains a turbine-generator and a boiling water nuclear reactor.
• Figure 2 is a graph of the percent of rated thermal power versus core flow illustrating an expanded operating domain and power uprate of the boiling water reactor shown in Figure 1.
• Figure 3 is a flow chart of a computer controlled safety analysis method to facilitate increasing the power output of the boiling water nuclear reactor shown in Figure 1, in accordance with an embodiment of the present invention.
• FIG 1 is a schematic diagram of the basic components of a power generating system 8.
• the system includes a boiling water nuclear reactor 10 which contains a reactor core 12.
• Water 14 is boiled using the thermal power of reactor core 12, passing through a water-steam phase 16 to become steam 18.
• Steam -NS-6034 is a schematic diagram of the basic components of a power generating system 8.
• An operating domain 40 of reactor 10 is characterized by a map of the reactor thermal power and core flow as illustrated in Figure 2.
• reactors are licensed to operate at or below a flow control/rod line 42 characterized by an operating point 44 defined by 100 percent of the original rated thermal power and 100 percent of rated core flow.
• reactors are licensed to operate with a larger domain, but are restricted to operation at or below a flow control/rod line 46 characterized by an operating point 48 defined by 100 percent of the original rated thermal power and 75 percent of rated core flow.
• Lines 50 represent the potential upper boundary of operating domain 40.
• An optimum power uprate level is defined based on the plant physical capabilities and financial goals of the owner/operator of the power plant.
• FIG. 3 is a flow chart of a computer controlled safety analysis method 60 to facilitate increasing the power output of boiling water nuclear reactor 10 in accordance with an embodiment of the present invention.
• a BWR utility owner needs to submit to the appropriate nuclear regulatory body a plant-specific power uprate safety evaluation report which details the various technical analyses performed in demonstration of the plant safe operation at the higher power output level.
• a safety report review period there can be several requests for additional information from the regulatory body that involve time and effort from the BWR utility owner and its contractor(s) to -NS-6034
• Method 60 includes selecting 62 generic computer-based safety evaluations from a database of generic safety evaluations already performed at the power uprate condition, comparing 64 plant design configuration with the range of plant characteristics assumed in the generic evaluations, and validating 66 the applicability of the generic safety evaluations to the specific plant application.
• Method 60 also includes performing 68 specific evaluations at reactor operating conditions outside the range of application of the selected generic evaluations, or are not included in the generic evaluations database. Some of these plant-specific evaluations are performed in a simplified manner based on the results obtained from the generic evaluations.
• Method 60 also includes inputting 70 data from the selected generic safety evaluations and the specific safety evaluations into licensing report templates stored in a report database and outputting licensing reports for submittal to a nuclear regulatory body.
• a licensing report electronic template is embedded with responses to questions from the regulatory body from similar power uprate submittals.
• Method 60 includes evaluating 72 the core and fuel performance at increased power output.
• the evaluations provide the predictions for the thermal and mechanical integrity of the fuel during normal steady-state operation, anticipated operational occurrences or accident events.
• the evaluations also account for the plant operating strategy, the length of the cycle of operation and contingency modes of operation such as with specific equipment declared out-of-service or equipment with degraded performance outputs.
• Evaluating 72 core and fuel performance impact at increased power output includes determining 74 limiting anticipated transient without scram (ATWS) events for increased core thermal power output
• Some (ATWS) events include Main Steam Isolation Valve Closure (MSIVC); Pressure Regulator Failure- Open (PRFO); Loss of Offsite Power (LOOP); and Inadvertent Opening of a Relief Valve (IORV).
• MSIVC Main Steam Isolation Valve Closure
• PRFO Pressure Regulator Failure- Open
• LOOP Loss of Offsite Power
• IORV Inadvertent Opening of a Relief Valve
• the analysis takes into account ATWS mitigating features, such as, the recirculation pump trip (RPT), alternate rod insertion (ART), and the Standby Liquid Control System (SLCS) performace. Plots of important parameters are created, and the peak values of neutron flux, average fuel heat flux and vessel pressure are calculated for each of the four events.
• the determined ATWS events for increased core thermal power output are compared to
• Method 60 also includes evaluating 76 the mechanical and structural integrity of system, structures and components (SSC) inside and outside the reactor pressure vessel (RPV) at the power uprate conditions, including effects from increase temperature, flow, pressure and radiation.
• SSC system, structures and components
• RSV reactor pressure vessel
• SSCs inside the RPV include, for example, the core shroud, the core support plate, the reactor core top guide, and the steam dryer.
• SSCs outside of the RPV include, for example, the biological shield wall, the piping/valves/pumps system, and the containment building.
• a plant-specific computer-based model of the RPV and the internal components is developed.
• the plant thermal-hydraulic initial conditions are also developed via computer simulation for steady-state as well as transients and accident conditions.
• the resulting loads on the SSCs are calculated and compared to specific design criteria to determine the SSCs mechanical integrity under steady-state or accident scenarios.
• Method 60 also includes evaluating 78 the capability of the safety equipment performance to maintain the plant in a continuously controlled state and to minimize any adverse impact to the public health and safety during anticipated operational occurrences or accident events.
• the evaluations are based on the original system design specifications, current system operational data and the contingency mode of operation with selected equipment either declared out-of-service or with degraded performance.
• Evaluating 78 safety equipment performance includes calculating 80 the range of core power over which the Reactor Core Isolation Cooling System (RCIC) prevents the core from uncovering during a loss of feedwater event.
• the primary purpose of the RCIC System is to maintain sufficient coolant in the reactor vessel such that the core is not uncovered in the event of reactor isolation accompanied by loss of coolant flow from the reactor feedwater system. This event is the limiting transient, which would challenge core cooling.
• the RCIC System should provide sufficient coolant makeup such that the water level in the reactor downcomer remains above the top of active fuel. If the downcomer water level falls below the top of active fuel, the emergency procedure guidelines direct the operator to depressurize the vessel and use the low pressure Emergency Core Cooling System (ECCS) to restore core cooling. This course of action is undesirable, because it results in exceeding the recommended vessel depressurization rate.
• ECCS Emergency Core Cooling System
• method 60 includes determining 82 stability interim corrective actions during increased core power output operation.
• Method 60 further includes evaluating 84 reactor control and instrumentation systems at increased power output operation.
• the instrument setpoints affected by the increase in thermal power, steam flow, operating pressure, and radiation are recalculated initially as analytical limits (ALs).
• the equipment specific characteristics, such as accuracy, drift and delay are factored in the ALs which are then converted into actual instrumentation setpoints.
• method 60 includes calculating 86 reactor set points at increased power output operating conditions to ensure safe plant operation at the power uprate condition.
• the analytical limit is the value of the sensed process variable prior to or at the point when a desired action is to be initiated.
• the AL is set so that appropriate licensing safety limits are not exceeded, as confirmed by plant performance analysis. This analysis considers instrument response time, transient overshoot and model accuracy.
• AV Allowable Value
• the Nominal Trip Set Point (NTSP) value is calculated from the AL by taking into account instrument drift in addition to the instrument accuracy, calibration and process measurement errors.
• the difference between the AL and the AV allows for channel instrument accuracy, calibration accuracy, process measurement accuracy, and primary element accuracy.
• the margin between the AV and the NTSP allows for instrument drift that might occur during the established surveillance period. If, during the surveillance period, an instrument setpoint drifts in a non-conservative direction but not beyond the AV, instrument performance is still within the requirements of the plant safety analysis.
• Method 60 also includes outputting 88 data to facilitate plant documentation updates in support of the power uprate operation.
• the output data serves to facilitate an update of the site operational procedures, engineering drawings and calculations, design bases documents, and training programs, including the plant simulator.
• method 60 includes calculating 90 the variables and limit curves which define when operator actions are required.
• the -NS-6034
• I Change rated reactor power only. ⁇ . Change lowest safety/relief valve lift pressure setpoint in addition to rated reactor power. in. Change containment operating temperatures in addition to rated reactor power.
• Method 60 further includes computing 92 a probabilistic risk assessment at an increased core thermal power output and comparing the assessment to a generic evaluation probabilistic risk assessment. Plants seeking a power uprate are expected to request an amendment to their license consistent with the considerations which govern their current license. That is, there is no change in the licensing basis for the plant. An amendment involves no significant hazard (NSH) -NS-6034
• a comprehensive assessment of the impact of power uprate on plant risk is obtained by reviewing the effect of uprate on the Individual Plant Examination (IPE). This includes the effect of the uprate on accidents and other events. Most nuclear plants have completed an IPE by performing a Probabilistic Safety Assessment (PSA).
• PSA Probabilistic Safety Assessment
• a Level 1 PSA models the events that lead to core damage and calculates the core damage frequency.
• a Level 2 PSA models the core melt progression and containment failure and calculates the frequency and magnitude of radioactive release.
• the above described method 60 provides a systematic, pre- approved approach for utility owners/operators of a boiling water reactor to license the thermal power uprate and thereby maximize revenues from the operation of the nuclear plant.
• Method 60 facilitates a BWR utility owner in developing the most reliable and proven approach to obtain a license amendment for power uprate in a timely manner and consistent with the current regulatory and licensing requirements.
• Standardized processes ensure consistency in all BWR power uprate projects and to bring increased efficiency to the overall approach.
• the amount of power increase can be very significant from the viewpoint of electrical power supply, for example, 20% above the original licensed thermal power.

Landscapes

• Engineering & Computer Science (AREA)
• General Engineering & Computer Science (AREA)
• Physics & Mathematics (AREA)
• Plasma & Fusion (AREA)
• High Energy & Nuclear Physics (AREA)
• Monitoring And Testing Of Nuclear Reactors (AREA)

Priority Applications (5)

Applications Claiming Priority (1)

Publications (1)

Family

ID=21742896

Family Applications (1)

Country Status (5)

Cited By (3)

Families Citing this family (1)

Citations (2)

• 2001
  • 2001-10-05 WO PCT/US2001/031404 patent/WO2003032327A1/en active Application Filing
  • 2001-10-05 JP JP2003535203A patent/JP2005505768A/ja not_active Withdrawn
  • 2001-10-05 EP EP01977599A patent/EP1436816A1/en not_active Ceased
  • 2001-10-05 MX MXPA04003150A patent/MXPA04003150A/es active IP Right Grant
  • 2001-11-22 TW TW090128954A patent/TW531758B/zh not_active IP Right Cessation

Patent Citations (2)

Also Published As

Similar Documents

Legal Events