METHOD FOR LICENSING INCREASED POWER OUTPUT OF A BOILING WATER NUCLEAR REACTOR

Abstract: A computerized method (60) for licensing increased power output of a boiling water nuclear reactor includes selecting (62) generic safety evaluations from a database of generic evaluations, comparing (64) reactor operating conditions at an increased power output with the reactor operating conditions of the selected generic evaluations, validating 66) the applicability of the generic analyses, performing (68) specific evaluations at reactor operating conditions outside the conditions of the selected generic evaluations and safety evaluations not included in the generic evaluations database, and outputting plant-specific licensing reports for increased power output.

Description: -NS-6034 METHOD FOR LICENSING INCREASED POWER OUTPUT 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 -NS-6034 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 DESCRIPTION OF THE INVENTION [0007] In an exemplary embodiment, a computerized method for licensing increased power output of a boiling water nuclear reactor is provided. The method 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 reactor operating conditions outside the conditions of the selected generic evaluations and safety evaluations not included in the generic evaluations database. [0008] In another exemplary embodiment, a system for licensing increased power output of a boiling water reactor is provided. The system 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. 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 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 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. 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 -NS-6034 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 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. To obtain a license amendment for power uprate, 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. During the 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 resolve. Upon a satisfactory review of the safety report, an amendment to the plant operating license is granted by the nuclear regulatory body to reflect the uprated core thermal power condition. The license amendment request should be 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 a power uprate. Consideration of potential significant hazards establish that operation of the facility in accordance with the proposed amendment do 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. [0016] 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. [0017] Typically, the generic safety evaluations have been reviewed and approved by the appropriate nuclear regulatory body. By showing that the operating conditions of reactor 10 is within the constraints of a pre-approved generic evaluation eliminates the need to re-evaluate reactor 10 at the increased core thermal power output for the conditions covered by the generic evaluation. Detailed specific -NS-6034 evaluations are conducted only for conditions that are outside the bounding conditions of the generic evaluations which simplifies the plant-specific evaluations as well as the overall licensing procedure. [0018] 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. [0019] 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). 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 generic evaluation ATWS events. [0020] 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. These SSCs must maintain their structural integrity under dynamic loadings or vibrational effects and perform their original intended functions, such pressure boundary components or core cooling geometry components. -NS-6034 [0021] 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. To determine their structural integrity, 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. [0022] 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. [0023] 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. [0024] The higher core power levels associated with power uprate will result in more boil-off and a lower water level in the reactor vessel, increasing the potential for core uncovery. The RCIC System should provide sufficient makeup, such that the reactor core remains covered with water until stable conditions are achieved. -NS-6034 [0025] In addition, 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. [0026] To confirm continued application of the stability corrective actions during uprated power operation and to describe the affect of uprated operation on specific long-term solutions, method 60 includes determining 82 stability interim corrective actions during increased core power output operation. [0027] 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. [0028] To show that to operation of reactor 10 is within the envelope of the pre-approved generic evaluations, 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 determination of setpoints for sensed parameters, which are directly associated with an abnormal plant transient or accident analyzed in the Safety Analysis Report (SAR), are based on the Analytical Limits (AL) which are established as part of the safety analysis. 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. -NS-6034 [0029] When a change is made to an AL due to power uprate, a new Allowable Value (AV) must be established. An AV is determined from the AL by providing allowances for the specified or expected calibration capability, accuracy of the instrumentation and process measurement errors. This value is then defined as the Technical Specification (Tech Spec) limit for the parameter and prescribed as a license condition for the plant. [0030] 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. [0031] Not all parameters have an associated AL based on safety analysis (e.g., main steamline radiation monitor). An AV, or design basis Tech Spec limit, may be defined directly based on plant licensing requirements, previous operating experience or other appropriate criteria. The NTSP is then calculated from the AV, allowing for instrument drift. Where appropriate, a NTSP may be determined directly based on operating experience or engineering judgment. [0032] 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. [0033] To evaluate the effect that increased power output has on plant emergency operating procedures, method 60 includes calculating 90 the variables and limit curves which define when operator actions are required. The -NS-6034 operator actions in the plant emergency operating procedures do not change as a result of increasing rated reactor power; only the conditions at which some of the actions are specified will change. The scope of recalculation is dependent upon the magnitude of the plant changes associated with the power uprate. The recalculations are included in the following categories: I. Change rated reactor power only. π. 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. IV. Change fuel type in addition to rated reactor power, but the new fuel has the same peak linear heat generation rate and the same fuel rod dimensions. V. Change fuel type in addition to rated reactor power, and the new fuel has a different peak linear heat generation rate and/or fuel rod dimensions. [0034] These categories encompass all the expected changes associated with extended power uprate that affect the plant emergency operating procedures variables and curves. For example, if the power uprate causes both the lowest safety/relief valve lift pressure setpoint to change and has a new fuel type loaded, then both Categories II and IV (or V) need to be examined. However, when a plant-specific uprate program is defined, the affected plant values will be verified against the plant data required for plant emergency operating procedures calculations to ensure that no other values are affected. [0035] 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 consideration if 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. [0036] Accident probability is not significantly increased by power uprate. The small increase in operating pressure, and the smaller increase in temperature, have no significant effect on LOCA probability. The occurrence frequency of accident precursors and transients is addressed when required by applying an appropriate setpoint methodology to insure that acceptable trip avoidance is provided after the uprate during operational transients. [0037] 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). 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. [0038] The assessment of the effect of power uprate on the plant IPE will consider the effect of power uprate on IPE inputs and assumptions such as: Initiating Event Frequency; Success Criteria; Component Failure Rates; and Time Available for Operator Action and Equipment Restoration [0039] As part of the IPE, utilities identify any plant vulnerabilities associated with core damage potential and containment performance. The scope of the study that assesses the impact of power uprate on plant IPE is sufficient to identify any new vulnerabilities that are introduced by the power uprate. If new vulnerabilities are identified, they will be reported in the Licensing Report. If no new vulnerabilities are identified, it can be concluded that power uprate had negligible impact on plant -NS-6034 risk. Changes in accident frequency which do not add vulnerabilities or significantly increase core damage frequency are themselves insignificant. [0040] 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. [0041] 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: -NS-6034 WHAT IS CLAIMED IS: 1. A computerized method (60) for licensing increased power output of a boiling water nuclear reactor power plant comprising: selecting (62) generic safety evaluations from a database of generic evaluations; comparing (64) reactor operating conditions at an increased power output with the reactor operating conditions of the selected generic evaluations; validating (66) applicability of the generic evaluations; and performing (68) 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. 2. A method (60) in accordance with Claim 1 further comprising: 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 a plant specific licensing report for submittal to a nuclear regulatory body. 3. A method (60) in accordance with Claim 1 further comprising evaluating (72) core and fuel performance at increased power output. 4. A method (60) in accordance with Claim 3 wherein evaluating core and fuel performance at increased power output comprises: determining (74) limiting anticipated transient without scram events for increased core thermal power output; and -NS-6034 comparing the limiting anticipated transient without scram events at increased power output to a generic evaluation anticipated transient without scram events. 5. A method (60) in accordance with Claim 1 further comprising evaluating (76) mechanical and structural integrity of systems, structures, and components inside and outside the nuclear reactor pressure vessel. 6. A method (60) in accordance with Claim 1 further comprising evaluating (78) the capability of reactor safety equipment to maintain the reactor in a continuously controlled state. 7. A method (60) in accordance with Claim 6 wherein evaluating (78) the capability of reactor safety equipment to maintain the reactor in a continuously controlled state comprises: calculating (80) the range of core power over which the Reactor Core Isolation Cooling System prevents the core from uncovering during a loss of feedwater event; and comparing the calculated core power to a generic evaluation core power range. 8. A method (60) in accordance with Claim 1 wherein comparing (64) operating conditions comprises: determining (82) stability corrective actions during increased core power output operation; comparing the determined stability corrective actions during increased core power output operation to a generic evaluation stability interim corrective actions. 9. A method (60) in accordance with Claim 1 further comprising evaluating (84) reactor control and instrumentation systems at increased power output. -NS-6034 10. A method (60) in accordance with Claim 9 wherein evaluating (84) reactor control and instrumentation systems at increased power output comprises calculating reactor set points at increased power output operating conditions. 11. A method (60) in accordance with Claim 1 further comprising outputting (88) data to facilitate plant documentation updates for increased power output operation. 12. A method (60) in accordance with Claim 1 wherein comparing (64) operating conditions comprises calculating (90) the variables and limit curves which define when operator actions are required for increased power output operation. 13. A method (60) in accordance with Claim 1 wherein comparing (64) operating conditions comprises: computing (92) a probabilistic risk assessment at an increased core thermal power output; comparing the results of the probabilistic risk assessment at an increased core thermal power output to a generic evaluation probabilistic risk assessment. 14. A system for licensing increased power output of a boiling water nuclear reactor power plant (8), said system comprising a computer configured to: simulate operation of the nuclear reactor (10) at an increased power output; select (62) generic safety evaluations from a database of generic evaluations; compare (64) reactor operating conditions at an increased power output with the reactor operating conditions of the selected generic evaluations; validate (66) applicability of the generic evaluations; and -NS-6034 perform (68) 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. 15. A system in accordance with Claim 14 wherein said computer is further configured to: input (70) data from the selected generic safety evaluations and the specific safety evaluations into licensing report templates stored in a report database; and output plant-specific licensing reports for submittal to a nuclear regulatory body. 16. A system in accordance with Claim 14 wherein said computer is further configured to evaluate (72) core and fuel performance at increased power output. 17. A system in accordance with Claim 16 wherein said computer is further configured to: determine (74) limiting anticipated transient without scram events for increased core thermal power output; and compare the limiting anticipated transient without scram events at increased power output to a generic evaluation anticipated transient without scram events. 18. A system in accordance with Claim 14 wherein said computer is further configured to evaluate (76) mechanical and structural integrity of systems, structures, and components inside and outside the nuclear reactor pressure vessel. 19. A system in accordance with Claim 14 wherein said computer is further configured to evaluate (78) the capability of reactor safety equipment to maintain the reactor in a continuously controlled state. -4-NS-6034 20. A system in accordance with Claim 19 wherein said computer is further configured to: calculate (80) the range of core power over which the Reactor Core Isolation Cooling System prevents the core from uncovering during a loss of feedwater event; and compare the calculated core power to a generic evaluation core power range. 21. A system in accordance with Claim 14 wherein said computer is further configured to: determine (82) stability corrective actions during increased core power output operation; compare the determined stability corrective actions during increased core power output operation to a generic evaluation stability interim corrective actions. 22. A system in accordance with Claim 14 wherein said computer is further configured to evaluate (84) a reactor control and instrumentation system at increased power output. 23. A system in accordance with Claim 22 wherein said computer is further configured to calculate (86) reactor set points at increased power output operating conditions. 24. A system in accordance with Claim 14 wherein said computer is further configured to output (88) data to facilitate plant documentation updates for increased power output operation. 25. A system in accordance with Claim 14 wherein said computer is further configured to calculate (90) the variables and limit curves which define when operator actions are required for increased power output. -NS-6034 26. A system in accordance with Claim 14 wherein said computer is further configured to: compute (92) a probabilistic risk assessment at an increased core thermal power output; compare the results of the probabilistic risk assessment at an increased core thermal power output to a generic evaluation probabilistic risk assessment.

Inventors: ECKERT EUGENE C, HOANG HOA X, MARQUINO WAYNE, ROBARE DAVID J, SEDNEY KATHY K, ECKERT, EUGENE, C., HOANG, HOA, X., MARQUINO, WAYNE, ROBARE, DAVID, J., SEDNEY, KATHY, K.
Applicants: GEN ELECTRIC, GENERAL ELECTRIC COMPANY
Application year: 2001
Main classification: G21D3
IPC codes: G21C17, G21D3