EP1407458A1 - Procede et systeme permettant de realiser une analyse de surete sur un reacteur nucleaire a eau bouillante - Google Patents

Procede et systeme permettant de realiser une analyse de surete sur un reacteur nucleaire a eau bouillante

Info

Publication number
EP1407458A1
EP1407458A1 EP01950906A EP01950906A EP1407458A1 EP 1407458 A1 EP1407458 A1 EP 1407458A1 EP 01950906 A EP01950906 A EP 01950906A EP 01950906 A EP01950906 A EP 01950906A EP 1407458 A1 EP1407458 A1 EP 1407458A1
Authority
EP
European Patent Office
Prior art keywords
accordance
core
increased
power
nuclear reactor
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Withdrawn
Application number
EP01950906A
Other languages
German (de)
English (en)
Inventor
Wayne Marquino
Eugene C. Eckert
Daniel C. Pappone
Kathy K. Sedney
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
General Electric Co
Original Assignee
General Electric Co
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by General Electric Co filed Critical General Electric Co
Publication of EP1407458A1 publication Critical patent/EP1407458A1/fr
Withdrawn legal-status Critical Current

Links

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 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.
  • 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 neutron flux or core power
  • feed water temperature and flow rate feed water temperature and flow rate
  • steam flow rate provided to the turbine
  • various status indications of the BWR systems 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.
  • 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 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.
  • 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 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.
  • 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 core 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 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.
  • 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 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.
  • 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.
  • 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.
  • 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.
  • a safety analysis includes a broad set of transient events that include:
  • 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.
  • 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.
  • SEs Safety Evaluations
  • the evaluation of radiological consequences includes the effect of a higher power level.
  • 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.
  • 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.
  • 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.
  • the magnitude of the potential radiological consequences of a design basis accident 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.
  • DBA design basis accident
  • 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.
  • the magnitude of the uprate may be limited, to maintain the radiological consequences below regulatory guidelines.
  • 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.
  • 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.
  • 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.
  • 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.
  • 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.
  • 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.
  • 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.
  • Method 60 further includes computing 80 an increase in turbine steam flow at the increased core thermal power.
  • 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.
  • 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.
  • 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.
  • 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.
  • 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.
  • Computerized method 60 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.

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  • 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)

Abstract

L'invention concerne un procédé (60) permettant de réaliser une analyse de sûreté informatisée sur un réacteur nucléaire à eau bouillante (10) pour permettre à un utilisateur d'obtenir plus facilement une modification de permis délivrée par un organisme de réglementation nucléaire pour exploiter ledit réacteur nucléaire à eau bouillante à une puissance thermique de coeur accrue. Le procédé selon l'invention comprend un nombre minimum de calculs informatiques permettant de démontrer à travers une simulation informatique que la sûreté de l'exploitation du réacteur nucléaire n'est pas compromise. Ce procédé consiste à limiter (62) le fonctionnement du réacteur nucléaire à un domaine d'exploitation sûr (40), déterminé par une analyse de sûreté antérieure, et à démontrer (64) que le domaine d'exploitation sûr déterminé par l'analyse de sûreté antérieure s'applique au fonctionnement du réacteur nucléaire à une puissance thermique de coeur accrue.
EP01950906A 2001-07-05 2001-07-05 Procede et systeme permettant de realiser une analyse de surete sur un reacteur nucleaire a eau bouillante Withdrawn EP1407458A1 (fr)

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
PCT/US2001/021327 WO2003005376A1 (fr) 2001-07-05 2001-07-05 Procede et systeme permettant de realiser une analyse de surete sur un reacteur nucleaire a eau bouillante

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EP1407458A1 true EP1407458A1 (fr) 2004-04-14

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EP (1) EP1407458A1 (fr)
JP (1) JP2004534237A (fr)
MX (1) MXPA03011912A (fr)
WO (1) WO2003005376A1 (fr)

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Publication number Priority date Publication date Assignee Title
RU2005116169A (ru) 2005-05-20 2006-11-27 Вадим Игоревич Дунаев (RU) Способ и система анализа и оценки безопасности технологического процесса
US8700363B2 (en) 2010-09-08 2014-04-15 The Boeing Company ETOPS IFSD risk calculator
US9753894B2 (en) 2010-09-08 2017-09-05 The Boeing Company Establishing availability of a two-engine aircraft for an ETOPS flight or an ETOPS flight path for a two-engine aircraft
US10990714B2 (en) * 2015-12-22 2021-04-27 Bwxt Mpower, Inc. Apparatus and method for safety analysis evaluation with data-driven workflow
CN114125079B (zh) * 2021-09-07 2023-09-12 北京网藤科技有限公司 一种火电安全模拟平台协议解析系统及其解析方法

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JPS5552998A (en) * 1978-10-16 1980-04-17 Hitachi Ltd Reactor recirculation flow rate control device
US5293411A (en) * 1989-07-14 1994-03-08 Hitachi, Ltd. Nuclear reactor power control method and device
US5524128A (en) * 1993-11-17 1996-06-04 Entergy Operations, Inc. Boiling water reactor stability control
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US6697447B1 (en) * 1999-12-30 2004-02-24 General Electric Company Maximum extended load line limit analysis for a boiling water nuclear reactor

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MXPA03011912A (es) 2004-03-26
JP2004534237A (ja) 2004-11-11
WO2003005376A1 (fr) 2003-01-16

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