METHOD AND SYSTEM FOR PERFORMING A SAFETY ANALYSIS OF A BOILING WATER NUCLEAR REACTOR

Abstract: A method (60) for performing a computerized safety analysis of a boiling water nuclear reactor (10) to facilitate a user in obtaining a licence amendment from a nuclear regulatory body to operate the boiling water nuclear reactor at an increased core thermal output is provided. The method has a minimum number of computer calculations and is demonstrated through computer simulation that safe operation of the nuclear reactor is not compromised. The method includes constraining (62) operation of the nuclear reactor to a safe operating domain (40) determined by a previous safety analysis, and demonstrating (64) that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output.

Description: METHOD AND SYSTEM FOR PERFORMING A SAFETY ANALYSIS OF A BOILING WATER NUCLEAR REACTOR BACKGROUND OF THE INVENTION [0001] This invention relates generally to nuclear reactors and more particularly to methods for increasing thermal power output of boiling water reactors. [0002] A typical boiling water reactor (BWR) 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. [0003] The BWR includes several conventional closed-loop control systems that control various individual operations of the BWR in response to demands. For example a control rod drive control system (CRDCS) controls the position of the control rods within the reactor core and thereby controls the rod density within the core which determines the reactivity therein, and which in turn determines the output power of the reactor core. A recirculation flow control system (RFCS) controls core flow rate, which changes the steam/water relationship in the core and can be used to change the output power of the reactor core. These two control systems work in conjunction with each other to control, at any given point in time, the output power of the reactor core. A turbine control system (TCS) controls steam flow from the BWR to the turbine based on pressure regulation or load demand. [0004] The operation of these systems, as well as other BWR control systems, is controlled utilizing various monitoring parameters of the BWR. Some 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 turbine, neutron flux or core power, feed water temperature and flow rate, steam flow rate provided to the turbine and various status indications of the BWR systems. Many monitoring parameters are measured directly, while others, such as core thermal power, are calculated using measured parameters. Outputs from the sensors and calculated parameters are input to an emergency protection system to assure safe shutdown of the plant, isolating the reactor from the outside environment if necessary, and preventing the reactor core from overheating during any emergency event. [0005] Historically, 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). [0006] To operate at a thermal power output higher than the rated thermal output licensed by the nuclear regulatory body, a license amendment approved by the nuclear regulatory body is needed. Typically, a safety analysis of the nuclear reactor at the proposed new operating parameters is required before approval can be obtained from the nuclear regulatory body. BRIEF SUMMARY OF THE INVENTION [0007] In an exemplary embodiment, a method for performing a computerized safety analysis of a boiling water nuclear reactor analysis to facilitate a user in obtaining a license amendment from a nuclear regulatory body to operate the boiling water nuclear reactor at an increased core thermal output is provided. The method has a minimum number of computer calculations and is demonstrated through computer simulation that safe operation of the nuclear reactor is not compromised. The method includes constraining operation of the nuclear reactor to a safe operating domain determined by a previous safety analysis, and demonstrating that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output. [0008] In another exemplary embodiment, a system for performing a safety analysis of a boiling water nuclear reactor to facilitate a user in obtaining a license amendment from a nuclear regulatory body to operate the boiling water nuclear reactor at an increased core thermal output is provided. The system includes a computer configured to simulate operation of the nuclear reactor, to constrain operation of the nuclear reactor to a safe operating domain determined by a previous safety analysis, and to demonstrate that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output. BRIEF DESCRIPTION OF THE DRAWINGS [0009] 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. [0010] Figure 2 is a graph of the percent of rated core thermal power versus core flow illustrating an expanded operating domain and power uprate of the boiling water reactor shown in Figure 1. [0011] Figure 3 is a flow chart of a computerized 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. DETAILED DESCRIPTION OF THE INVENTION [0012] Figure 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 18 flows through piping in a steam flow path 20 to a turbine flow control valve 22 which controls the amount of steam 18 entering steam turbine 24. Steam 18 is used to drive turbine 24 which in turn drives electric generator 26 creating electric power. Steam 18 flows to a condenser 28 where it is converted back to water 14. Water 14 is pumped by feedwater pump 30 through piping in a feedwater path 32 back to reactor 10. [0013] 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. Typically, 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. In some circumstances, 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. [0014] It is desirable to operate at a. thermal power greater than 100 percent of the original rated licensed thermal power, sometimes referred to as a power uprate. Lines 50 represent the potential upper boundary of operating domain 40. To operate in the uprate region of operating domain 40, operating conditions and/or equipment modifications are needed. An optimum power uprate level is defined based on the plant physical capabilities and financial goals of the owner/operator of the power plant. [0015] Figure 3 is a flow chart of a computerized 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. Method 60 has a minimum number of calculations and demonstrates through computer simulation that safe operation of the nuclear reactor is not compromised at increased power output of reactor 10. Method 60 includes constraining 62 operation of the nuclear reactor to a safe operating domain determined by a previous safety analysis, and demonstrating 64 that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output. [0016] Plants seeking a power uprate are required to request a license amendment consistent with the considerations which govern the current license. Particularly, there is no change in the licensing basis for the plant, and no significant increases in the amount of effluents or radiation emitted from the facility are anticipated because of power uprate. Consideration of potential significant hazards establish that operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. [0017] As part of an uprate review process, the applicable plant safety analyses are evaluated. Where necessary, analyses of all limiting accidents and transients are performed at the uprated conditions to show continued compliance with applicable regulatory requirements. [0018] A review of the existing plant Safety Analysis Report and reload transients is conducted at the uprated conditions. Where necessary, analyses are performed to demonstrate compliance with the fuel thermal margin requirements and other applicable transient criteria. Of primary importance is an analysis of the transient events which are most limiting from the viewpoint of fuel thermal margin. [0019] Analysis of these most limiting events for uprated power at the most limiting conditions on the operating power/flow map (see Figure 2) will assure that fuel operating limits are met. [0020] A safety analysis includes a broad set of transient events that include: (1) Decrease in Core Coolant Temperature. (2) Increase in Reactor Pressure. (3) Decrease in Reactor Core Coolant Flow Rate. (4) Increase in Core Flow Rate. (5) Increase in Reactor Coolant Inventory. (6) Decrease in Reactor Coolant Inventory. (7) Increase in Reactivity. (8) Increase in Core Coolant Temperature. [0021] Evaluations are performed to show continued compliance with the nuclear regulatory agency rule on anticipated transient without scram performance (ATWS). ATWS rule compliance primarily involves alternate shutdown equipment which has been previously installed at each unit. The equipment remains and its performance at any changed conditions (due to uprate) is evaluated, for example at higher operating pressure. Where applicable, a bounding case is reanalyzed at the uprated power to confirm that adequate overpressure protection and suppression pool cooling are maintained for limiting cases in each BWR product line. This analysis also includes evaluation of any changes in pressure setpoints of the safety/relief valves and/or high pressure recirculation pump trip. In some cases these setpoints, as well as the allowable number of relief valves out of service may be re- optimized to improve the. ATWS response. Power uprate operation does not significantly affect the long-term ATWS response because it does not involve a uniquely higher rod line, and, therefore, there is no increase in the power level following the ATWS recirculation pump trip. [0022] Radiological consequences are evaluated or analyzed for uprated power conditions. This evaluation/analysis is based on the methodology, assumptions, and analytical techniques described in previous Safety Evaluations (SEs). The evaluation of radiological consequences includes the effect of a higher power level. In general, the radiation sources inside the fuel rods, creation of activation products outside of the fuel rods, and concentration of coolant activation activities are directly proportional to the thermal power. Therefore, the original radiation inventories, expressed in terms of curies per megawatt of thermal power, will bound the uprated condition, provided that the core design, fuel loading, and mean exposure are not changed significantly. If significant changes to the fuel loading or design parameters are made to optimize for uprated conditions, the uprate license application will re-perform the radiological evaluation to account for changes to the isotopic concentrations in the fuel. Issues relating to burnup and enrichment also need to be addressed if the uprated burnup and enrichment are to exceed any regulated conditions. [0023] During normal operation, the radiation levels in the plant are the result of radiation streaming from the reactor vessel or from radioisotopes carried in the reactor water, steam, or radwaste process. In all cases, these quantities are approximately proportional to core thermal power. Increases in normal radiological releases from routine operation are considered in the power uprate amendment requests. [0024] The magnitude of the potential radiological consequences of a design basis accident (DBA) is proportional to the quantity of fission products released to the environment. This quantity is a product of the activity released from the core and the transport mechanisms between the core and the effluent release point. For a steam line break or instrument line break accident, the radiological consequences will be, at most, proportional to the increase in power, since (1) the quantity of activity in the primary coolant and in the offgas is unaffected by power uprate (it is limited by Technical Specifications), and (2) the increase in coolant mass discharged to the environment is dependent on reactor pressure, which increases less than the power increase. For the remaining DBAs, the radiological releases are expected to increase, at most, by the amount of power uprate, since the only parameter of importance is the actual inventory of radioisotopes in the fuel rod and the mechanism of fuel failure is not likely to be influenced by power uprate. In some cases, the magnitude of the uprate may be limited, to maintain the radiological consequences below regulatory guidelines. [0025] To facilitate the above described analyses, method 60 includes the step of computing 66 a stability and an anticipated transient without scram performance. The stability and the anticipated transient without scram performance are bounded by the previous safety analysis by simulating operation along a previously licensed core flow control line such that after a loss of forced recirculation, the power is reduced to a previously analyzed power/flow condition. [0026] Method 60 also includes computing 68 power and flow critical power ratio adjustment factors at core powers above previously licensed power. Specifically, the off rated power and flow critical power ratio adjustment factors (K- and K f ) are bounded by the previous safety analysis such that only off rated factors for core powers above the presently licensed power are calculated. [0027] Further, method 60 includes computing 70 a maximum reactor dome pressure/core inlet subcooling combination that results in a containment pressurization rate within containment pressurization rate parameters used in the previous safety analysis. Specifically, the environmental qualification and loss of coolant accident dynamic loads are bounded by the present license conditions. [0028] To avoid reevaluating the fuel bundle design limits, method 60 includes computing 72 a core design with a reduced core radial peaking factor based on the increased core thermal power. The reduced radial peaking factor can be achieved by increasing the fraction of new fuel bundles loaded for operation at the higher power output. The computer outputs data to facilitate a user in modifying the fraction of new fuel bundles loaded for operation at the increased core thermal power. [0029] To avoid reevaluating the fuel rod design limits, method 60 includes computing 74 a new bundle enrichment gadolina concentration to flatten a rod to rod power distribution based on the increased core thermal output. The flattened distribution can be achieved by increasing gadolina concentrations in the fuel rods. [0030] Method 60 further includes computing 76 a maximum suppression pool temperature using decay heat characteristics based on a specific predetermined isotopic mixture of reactor fuel rods. The increase in the maximum suppression pool temperature is minimized at the higher thermal power output by replacing generic decay heat characteristics with characteristics based on the specific isotope mixture that is used to achieve the higher power output. This prevents the maximum suppression pool temperature from being too high. [0031] Method 60 further includes using 78 a previous safety relief valve stress analysis as a bounding condition for the safety relief valve stress at the increased core thermal power. The safety relief valve stress analysis remains bounding by limiting the increase in the safety relief valve opening setpoints to the maximum pressure previously analyzed for safety relief valve discharge. The safety relief valve opening setpoints can be optimized by increasing the setpoints by the minimum margin between the reactor vessel pressure and the safety mode opening pressure necessary to prevent inadvertent opening of the safety valves. [0032] Method 60 further includes computing 80 an increase in turbine steam flow at the increased core thermal power. By increasing the flow capacity of turbine 24, for example by machining the diaphragms of the turbine first stage, the increased power output can be accomplished without increasing the reactor pressure. Data is outputted to facilitate modification of the turbine to increase steam flow through the turbine without increasing reactor pressure. [0033] Method 60 further includes computing 82 a subset of limiting anticipated operational occurrences which bound all operational occurrences, and computing 84 an analysis of the subset of limiting anticipated operational occurrences at the increased core thermal output. Events which were demonstrated to be far from limiting at the previous licensed power output and which are determined to be similarly non-limiting for a reference plant, need not be analyzed on a plant specific basis at the higher power condition. [0034] Further, method 60 includes computing 86 parametric curves that characterize the boiling water nuclear reactor performance at the increased core thermal output, and normalizing 88 the parametric curves using non-dimensional parameters. Some examples of non-dimensional parameters include, but are not limited to, a ratio of safety valve capacity to steam flow at licensed core power, residual heat removal heat exchanger capacity as a fraction of licensed core power, and suppression pool capacity in minutes of steam flow at licensed core thermal power. The parametric curves permit accurate estimation of amount of power increase possible without violating safety analysis acceptance criteria and assist the user in demonstrating safe operation of the reactor at the increased power output. [0035] Of course, performing the safety analysis at a plurality of core power outputs which are between the current licensed maximum core power output and a predetermined new maximum core thermal power output can also be performed by method 60. Performing safety analysis at intermediate power levels permits the implementation of any necessary plant modifications for increased power output to be made incrementally, for example during normal or slightly extended refueling shutdowns rather than an extended shut-down for all plant modifications. [0036] Computerized method 60, described above, in an exemplary embodiment is web enabled and is run on a business entity's intranet. In a further exemplary embodiment, computerized method 60 is fully accessed by individuals having authorized access outside the firewall of the business entity through the Internet. In another exemplary embodiment, computerized method 60 is run in a Windows NT environment or simply on a stand alone computer system having a CPU, memory, and user interfaces. In yet another exemplary embodiment, computerized method 60 is practiced by simply utilizing spreadsheet software. [0037] While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims.

Claims: WHAT IS CLAIMED IS: 1. A method (60) for performing a computerized safety analysis of a boiling water nuclear reactor (10) to facilitate a user in obtaining a license amendment from a nuclear regulatory body to operate the boiling water nuclear reactor at an increased core thermal output, said method having a minimum number of computer calculations by demonstration through computer simulation that safe operation of the nuclear reactor is not compromised, said method comprising: constraining (62) operation of the nuclear reactor to a safe operating domain (40) determined by a previous safety analysis; and demonstrating (64) that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output. 2. A method (60) in accordance with Claim 1 further comprising computing (66) a stability and an anticipated transient without scram performance wherein the stability and the anticipated transient without scram performance are bounded by the previous safety analysis by simulating operation along a previously licensed core flow control line. 3. A method (60) in accordance with Claim 1 further comprising computing (68) power and flow critical power ratio adjustment factors at core powers above previously licensed power. 4. A method (60) in accordance with Claim 1 further comprising determining an environmental qualification by computing (70) a maximum reactor dome pressure/core inlet subcooling combination that results in a containment pressurization rate within containment pressurization rate parameters used in the previous safety analysis. 5. A method (60) in accordance with Claim 1 further comprising: computing (72) a core design with a reduced core radial peaking factor based on the increased core thermal power; and outputting data to facilitate a user in modifying the fraction of new fuel bundles loaded for operation at the increased core thermal power to eliminate the need to reanalyze fuel bundle design limits at the increased core thermal power output. 6. A method (60) in accordance with Claim 1 further comprising: computing (74) a new bundle enrichment gadolina concentration to flatten a rod to rod power distribution based on the increased core thermal power output; and computing a new, increased gadolina concentration in the fuel rods to achieve the computed new rod to rod power distribution to eliminate the need to reanalyze fuel rod design limits at the increased core thermal power output. - 7. A method (60) in accordance with Claim 1 further comprising computing (76) a maximum suppression pool temperature using decay heat characteristics based on a specific predetermined isotopic mixture of reactor fuel rods. 8. A method (60) in accordance with Claim 1 further comprising using (78) a previous safety relief valve stress analysis as a bounding condition for the safety relief valve stress at the increased core thermal power. 9. A method (60) in accordance with Claim 1 further comprising: computing (80) an increase in turbine steam flow at the increased core thermal power; and outputting data to facilitate modification of the turbine (24) to increase steam flow through the turbine without increasing reactor pressure. 10. A method (60) in accordance with Claim 1 further comprising: computing (82) a subset of limiting anticipated operational occurrences which bound all operational occurrences; and computing (84) an analysis of the subset of limiting anticipated operational occurrences at the increased core thermal output. 11. A method (60) in accordance with Claim 1 further comprising: computing (86) parametric curves that characterize the boiling water nuclear reactor (10) performance at the increased core thermal output; normalizing (88) the parametric curves using non-dimensional parameters; and outputting the normalized parametric curves to facilitate the user in demonstrating safe operation of the reactor at the increased power output. 12. A method (60) in accordance with Claim 1 further comprising performing the safety analysis at a plurality of core power outputs which are between the current licensed maximum core power output and a predetermined new maximum core thermal power output. 13. A system for performing a safety analysis of a boiling water nuclear reactor (13) to facilitate a user in obtaining a license amendment from a nuclear regulatory body to operate the boiling water nuclear reactor at an increased core thermal output, said system comprising a computer configured to: simulate operation of the nuclear reactor; constrain the simulated operation of the nuclear reactor to a safe operating domain (40) determined by a previous safety analysis; and demonstrate that the safe operating domain determined by the previous safety analysis applies to the operation of the nuclear reactor at the increased core thermal output. 14. A system in accordance with Claim 13 wherein said computer is further configured to compute a stability and an anticipated transient without scram performance wherein the stability and the anticipated transient without scram performance are bounded by the previous safety analysis by simulating operation along a previously licensed core flow control line. 15. A system in accordance with Claim 14 wherein said computer is further configured to compute power and flow critical power ratio adjustment factors at core powers above previously licensed power. . 16. A system in accordance with Claim 14 wherein said computer is further configured to determine an environmental qualification by computing a maximum reactor dome pressure/core inlet subcooling combination that results in a containment pressurization rate within containment pressurization rate parameters used in the previous safety analysis. 17. A system in accordance with Claim 13 wherein said computer is further configured to: compute a core (12) design with a reduced core radial peaking factor based on the increased core thermal power; and output data to facilitate a user in modifying the fraction of new fuel bundles loaded for operation at the increased core thermal power to eliminate the need to reanalyze fuel bundle design limits at the increased core thermal power output. 18. A system in accordance with Claim 17 wherein said computer is further configured to: compute a new bundle enrichment gadolina concentration to flatten a rod to rod power distribution based on the increased core thermal power output; and compute a new, increased gadolina concentration in the fuel rods to achieve the computed new rod to rod power distribution to eliminate the need to reanalyze fuel rod design limits at the increased core thermal power output. 19. A system in accordance with Claim 13 wherein said computer is further configured to compute a maximum suppression pool temperature using decay heat characteristics based on a specific predetermined isotopic mixture of reactor fuel rods. 20. A system in accordance with Claim 13 wherein said computer is further configured to use a previous safety relief valve stress analysis as a bounding condition for the safety relief valve stress at the increased core thermal power. 21. A system in accordance with Claim 13 wherein said computer is further configured to: compute an increase in turbine steam flow at the increased core thermal power; and output data to facilitate modification of the turbine (24) to increase steam flow through the turbine without increasing reactor pressure. 22. A system in accordance with Claim 13 wherein said computer is further configured to: compute a subset of limiting anticipated operational occurrences which bound all operational occurrences; and compute an analysis of the subset of limiting anticipated operational occurrences at the increased core thermal output. 23. A system in accordance with Claim 13 wherein said computer is further configured to: compute parametric curves that characterize the boiling water nuclear reactor (10) performance at the increased core thermal output; and normalize the parametric curves using non-dimensional parameters. 24. A system in accordance with Claim 13 wherein said computer is further configured to perform the safety analysis at a plurality of core power outputs which are between the current licensed maximum core power output and a predetermined new maximum core thermal power output.

Inventors: ECKERT EUGENE C, MARQUINO WAYNE, PAPPONE DANIEL C, SEDNEY KATHY K, ECKERT, EUGENE, C., MARQUINO, WAYNE, PAPPONE, DANIEL, C., SEDNEY, KATHY, K.
Applicants: GEN ELECTRIC, GENERAL ELECTRIC COMPANY
Application year: 2001
Main classification: G21D3
IPC codes: G21C17, G21D3