ANSI ANS 51.10-1991 Auxiliary feedwater system for pressurized water reactors《加压水冷反应堆的辅助供水系统》.pdf

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1、 I 0 . 10 i li z c ANSI/ANS-51.101991 auxiliary feedwater system for pressurized water reactors !leCeliRiily reftect recent iildustry initiative$. for risk informed daoomach to qUality assurance. U$eis sllouid CQI!Siderthe use appiIC8tion of this standard Secretariat American Nuclear Society Prepare

2、d by the American Nuclear Society Standards Committee Working Group ANS-51.10 Published by the American Nuclear Society 555 North Kensington A venue La Grange Park, Illinois 60525 USA Approved May 10, 1991 by the American National Standards Institute, Inc. ANSI/ANS-51.10-1991 Revision of ANSI/ANS-51

3、.10-1979 American National Standard Auxiliary Feedwater System for Pressurized Water Reactors 0 EAffiIE . ! OCT 14 211811 I ANSIANS American National Standard Designation of this document as an American National Standard attests that the principles of openness and due process have been followed in t

4、he approval procedure and that a consensus of those directly and materially affected by the standard has been achieved. This standard was developed under the procedures of the Standards Committee of the American Nuclear Society; these procedures are accredited by the American National Standards Inst

5、itute, Inc., as meeting the criteria for American National Standards. The consensus committee that approved the standard was balanced to ensure that competent, concerned, and varied interests have had an opportunity to participate. An American National Standard is intended to aid industry, consumers

6、, govern mental agencies, and general interest groups. Its use is entirely voluntary. The existence of an American National Standard, in and of itself, does not preclude anyone from manufacturing, marketing, purchasing, or using products, processes, or procedures not conforming to the standard. By p

7、ublication of this standard, the American Nuclear Society does not insure anyone utilizing the standard against liability allegedly arising from or after its use. The content of this standard reflects acceptable practice at the time of its ap proval and publication. Changes, if any, occurring throug

8、h developments in the state of the art, may be considered at the time that the standard is subjected to periodic review. It may be reaffirmed, revised, or withdrawn at any time in accord ance with established procedures. Users of this standard are cautioned to determine the validity of copies in the

9、ir possession and to establish that they are of the latest issue. The American Nuclear Society accepts no responsibility for interpretations of this standard made by any individual or by any ad hoc group of individuals. Requests for interpretation should be sent to the Standards Department at Societ

10、y Head quarters. Action will be taken to provide appropriate response in accordance with established procedures that ensure consensus on the interpretation. Comments on this standard are encouraged and should be sent to Society Headquarters. American Nuclear Society 555 North Kensington Avenue, La G

11、range Park, Illinois 60525 USA Copyright 1991 by American Nuclear Society. Any part of this standard may be quoted. Credit lines should read “Extracted from American National Standard ANSI!ANSc51.10-1991 with permission of the publisher, the American Nuclear Society.“ Reproduction prohibited under c

12、opyright convention unless written per mission is granted by the American Nuclear Society. Printed in the United States of America Addendum to Foreword ANSI/ ANS-51 . 1 0- 1 991 ; R2002 Auxiliary Feedwater System for Pressurized Water Reactors This standard has been reviewed and reaffirmed by the AN

13、S Nuclear Facilities Standards Committee (NFSC) with the recognition that it may reference other standards and documents that may have been superseded or withdrawn. The requirements of this document are met by using the version of the standards and documents referenced herein. It is the responsibili

14、ty of the user to review each of the references cited and to determine whether the use of the original references or more recent versions is appropriate for the facility. Variations from the standards and documents referenced in this standard should be evaluated and documented. The standard does not

15、 necessarily reflect recent industry initiatives for risk informed decision-making or a graded approach to quality assurance. Users should consider the use of these industry initiatives in the application of this standard. Foreword (This Foreword is not a part of American National Standard for Auxil

16、iary Feedwater System for Pressurized Water Reactors, ANSI/ANS-51.10-1991.) This standard is applicable to pressurized light water reactor nuclear power plants. The standard was originally issued in 1979 and has been extensively updated to reflect current regulatory directives, industry practice and

17、 experience, and available design guidance. Among the major revisions incorporated in this issue of the standard is station blackout (i.e., the loss of all alternating current power sources). A requirement is included that the system be capable of operating for a plant specific duration with the los

18、s of all alternating current power sources. The rationale for this requirement and the method of determining the plant specific duration are identified in Title 10, Code of Federal Regulations, Part 50.63, “Loss of All Alternating Current Power.“ In the 1979 edition of this standard, a generic durat

19、ion of two hours was identified and no rationale was included. The membership of Working Group ANS-51.10 of the Standards Committee of the American Nuclear Society during the development of this document was: J. F. Garibaldi, Chairman, Ebasco Services, Inc. D. G. Keith, Bechtel Power Corporation S.

20、Ku, GPU Nuclear Corporation K. J. Vehstedt, New York Power Authority W. T. LeFave, U.S. Nuclear Regulatory Commission In addition, the following individuals assisted in the preparation of this document: M. A. Anzalone, Ebasco Services, Inc. W. J. McTigue, Ebasco Services, Inc. J. S. Pfabe, Ebasco Se

21、rvices, Inc. D. Santory, Ebasco Services, Inc. The membership ofMC-1, LWR Design Criteria Standards Management Committee, at the time of its concurrence with the technical approval was: W. H. DArdenne, Chairman, General Electric Company E. J. Borella, Ebasco Services, Inc. D. M. Crowe, Georgia Power

22、 Company R. Fortier, Stone refer to American National Standard Cooldown Criteria for Light Water Reactors, ANSI/ANS-58.11-1983 (R1989) 5. (C) The minimum delivered flow shall be based on the following: (1) The maximum allowed temperature of the auxiliary feedwater supply consistent with plant analys

23、es. (2) The initial steam generator second ary side water level at the start of the event equal to the steam generator lowest level of protection grade setpoint with an allowance for error, at which a reactor trip is designed to occur. (3) The maximum time at which flow might be initiated after the

24、start of the event. (4) The minimum mass of secondary water available in the operable steam genera tor(s) when auxiliary feedwater is initiated. (5) The maximum volume of hot water in the auxiliary feed water flow path to the steam generators. American National Standard ANSI/ANS-51.10-1991 (6) The r

25、esidual heat generation result ing from long term operation at 102 percent of the rated thermal power level. (The decay heat rate should be calculated based upon American National Standard for Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1-1979 (R1985) 6, with 20 percent uncertainty up to 10

26、3 seconds and 10 per cent from 103 to 107 seconds.)4 (7) The maximum heat capacity of the reactor coolant system and the maximum stored energy in the reactor core for the assumed plant operating condition. (8) The maximum number of reactor cool ant pumps in operation. (9) Capability to isolate the b

27、lowdown and sampling systems with assessment of inven tory loss, valve closure times, and effects of the single failure criterion application. (10) The maximum depletion of flow to the operable steam generator(s) due to flow out of any potential break for the assumed plant oper ating condition. (11)

28、 The specified maximum time at which break flow may be terminated (as a result of auto matic system action or operator intervention). (12) AFS pump highest volume minimum recirculation flow provisions consistent with flow controls and possible single failures. (13) The minimum voltage or frequency,

29、or both, at the motor terminals. (D) The minimum delivered flow shall be de signed to be delivered to the appropriate operable steam generator(s) against a steam generator pres sure equivalent to the maximum accumulation pressure of the steam generator safety valves that are required to open to remo

30、ve the decay heat load. 5.L1.2 Maximum Flow. The maximum flow shall be selected consistent with the ability to con trol steam generator water level and avoid over filling. In addition, maximum flow shall be based upon the following: (A) The contribution of the maximum deliv ered flow to any single s

31、team generator following a steam or feedwater line break inside contain ment to the total permissible mass and energy 4Current review guidance by the U.S. Nuclear Regulatory Com mission is contained in NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plan

32、ts,“ Section 9.2.5, “Ultimate Heat Sink,“ Branch Technical Position ASB 9-2, “Residual Decay Energy for Light-Water Reactors for Long-Term Cooling,“ Rev. 2, July 1981, which specifies an uncertainty factor of 20 percent to 103 seconds and 10 per cent from 103 to 107 seconds. (NUREGs are available fr

33、om the National Technical Information Service, Springfield, VA 22161.) 3 American National Standard ANSI/ANS-51.10-1991 release to the containment. (The maximum deliv ered flow may be restricted to minimize the mass and energy release to the containment.) (B) The reactivity effect due to the maximum

34、 delivered total flow to all steam generators following a steam or feedwater line break under any pos tulated combination of initial reactor conditions consistent with the reactivity transient analysis contained in the plant safety analysis report. (C) The maximum permissible delivered flow to any s

35、team generator consistent with water hammer effects. (D) The contribution to the maximum deliv ered flow of the inaximum expected AFS pump minimum recirculation flow if such flow is de liberately diverted to the steam generators. (If minimum recirculation flow could only be inad vertently diverted t

36、o the steam generator(s), then the additional delivered flow shall be determined consistent with the single failure criterion.) (E) The worst case AFS pump motor loads based upon the maximum power consumption associated with maximum delivered flow on one train with runout on the other train consiste

37、nt with system single failure assumptions. (F) Design provisions for motor-driven AFS pumps to preclude damage during operation, or alternatively designed to trip or otherwise protect the motor and pumps, using features that prevent the pumps from reaching runout flow conditions. (G) The maximum vol

38、tage or frequency, or both, at the motor terminals. 5.1.1.3 Pump Selection. The design charac teristics of each pump in conjunction with asso ciated ancillary system design shall be such that: (A) Delivery of the minimum required design flow specified in 5.LL1 is assured within the time delay assume

39、d in the analysis of the event. (B) Maximum flows shall not exceed those specified in 5.1.1.2. Pump design or system design shall ensure that pumps, when operating within the bounds of anticipated system performance in cluding applicable single failure assumptions, do not trip (e.g., due to motor ov

40、erload) or will not be damaged (e.g., by runout or cavitation) such that there is a loss of system function. Equipment protective trips may be used consistent with system single failure assumptions. Minimum net positive suction head requirements (e.g., water levels, water temperature, dynamic starti

41、ng effects including their affect on protective instrumentation, and source switch over) shall be met. (C) Allowance shall be provided for recir culation flow, pump Wiiter, and testing accuracy. 4 (D) Where turbine-driven pumps are utilized, a nuclear safety-related, seismic Category I source of ste

42、am of an acceptable quality shall be pro vided. The size of the turbine shall be based on the maximum steam generator backpressure, the minimum available steam inlet pressure, and maximum AFS delivered and recirculation flows consistent with plant single failure analysis assumptions. The turbine-dri

43、ven pump shall be capable of operation at reduced steam pressures while delivering sufficient flow to satisfy all AFS nuclear safety functions independent of the motor driven AFS pumps. The turbine control system shall be designed to provide sufficiently rapid response on initial steam admission to

44、the turbine to avoid overspeed trips. The steam supply lines shall provide features for the continuous removal of condensate, or its elim ination by heating, as required. The turbine con trol system shall be designed to preclude damage and adverse effects due to any condensate that might accumulate.

45、 Design features shall accom modate periodic verification of turbine operability which might include quick start testing from cold (i.e., normal) conditions to ensure the proper oper ation of turbine overspeed trip features. (E) Where diesel-driven pumps are utilized, a nuclear safety-related, seism

46、ic Category I supply of fuel oil of an acceptable quality shall be pro vided. When required to meet the single failure criterion, the diesel-driven pump shall provide sufficient flow to satisfy all AFS nuclear safety functions independent of the motor-driven AFS pumps. The diesel and ancillary equip

47、ment and support systems (i.e., fuel oil, lubrication, engine cooling, combustion and exhaust air, and starting system(s) shall be nuclear safety-related, seismic Category I as required to support diesel-driven pump operations consistent with system single failure assumptions. Design features shall

48、accom modate periodic verification of diesel-driven pump operability. (F) Secondary pressure reduction control shall be assessed to ensure that the system as a whole, including the pumps, their drivers, dis charge flow and pressure control devices, and steam generator pressure control valves are cap

49、 able of cooling the plant down from post-transient conditions to conditions required for residual heat removal (RHR) system operation. (G) Long-term makeup capability to main tain steam generator secondary side water levels within required limits for up to 30 days following a loss of coolant accident (LOCA) shall be provided if required by the plant safety analysis. 5 5.1.2 Auxililary Feedwater Sources. The AFS shall be provided with a primary water source and a long-term post-LOCA water source as des cribed in 5.1.2.1 and 5.1.2.3. These nuclear safety related sources s