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1、- 7- . ANS 53-3 83 E 0724838 0098067 4 , = - - x. ANSIIANS-51.1-1983 nuclear safety criteria for the design of stationary pressurized water reactor plants -,-,- ANSIIANS-51 .I-1 983 O i I American National Standard Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants
2、 O Secretariat American Nuclear Society Prepared by the American Nuclear Society Standards Committee Working Group ANS-51.1 Published by the American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60525 USA Approved April 29, 1983 by the American National Standards Institute, I
3、nc. “ “ -,-,- American National Standard An American National Standard implies a consensus of those substantially con- cerned with its scope and provisions. An American National Standard is intended as a guide to aid the manufacturer, the consumer, and the general public. The existence of an America
4、n National Standard does not in any respect preclude anyone, whether he has approved the standard or not, from manufacturing, marketing, purchasing, or using products, processes, or procedures not conforming to the standard. American National Standards are subject to periodic review and users are ca
5、utioned to obtain the latest editions. CAUTION NOTICE This American National Standard may be reviewed or with- drawn at any time. The procedures of the American National Standards Institute require that action be taken to reaffirm, revise, or withdraw this standard no later than five years from the
6、date of publication. Purchasers of this standard may receive current information, including interpretation, on all standards published by the American Nuclear Society by calling or writing to the Society. Published by American Nuclear Society 555 North Kensington Avenue, La Grange Park, Illinois 605
7、25 USA Price: $45.00 Copyright O 1983 by American Nuclear Society. 0 - Any part of this standard may be quoted. Credit lines should read “Extracted from American National Standard ANSIIANS-51.1-1983 with permission of the publisher, the American Nuclear Society.” Reproduction prohibited under copyri
8、ght convention unless written per- a- mission is granted by the American Nuclear Society. Printed in the United States of America - -,-,- ANS 51.1 83 W 0721.1818 0098070 4 W Foreword (This Foreword is not a part of American National Standard Nuclear Safety Criteria for the Design of Stationary Press
9、urized Water Reactor Plants, ANSIIANS-51.1-1983.) This standard is a complete revision and combination of N18.2-1973/ANS-51.1 and N18.2a-19751ANS-51.8. It has been prepared by Subcommittee ANS-51, Pressurized Water Reactor Criteria, to incorporate additional requirements for the design of pressurize
10、d water reactor (PWR) nuclear power plants and to address three major areas: 1. 2. 3. Safety Classes The results of the ANS Nuclear Power Plant Standards Committee (NUPPSCO) Ad Hoc Committee and NUPPSCO Coordinating Working Group 3 on Equip- ment Classification are incorporated. These results define
11、 Safety Classes and specify requirements for all equipment and structures in a stationary nuclear power plant having a nuclear safety function. A methodology is given to classify all equipment into one of three Safety Classes according to its importance to nuclear safety and its capability for maint
12、enance, surveillance testing, and inspection, or into a Non-Nuclear Safety Class. In addition, classification inter- face criteria are defined. Plant Conditions The results of the NUPPSCO Coordinating Working Group 2 have been incor- porated. The concept of Plant Conditions is developed that include
13、s individual process conditions, combinations of process conditions, and the combinations of process conditions and external hazards that could result in simultaneous effects on plant equipment. Probability of occurrence is the unifying basis for the categorization of Plant Conditions. Design Requir
14、ements This standard provides a set of design requirements for all Safety Classes and Non-Nuclear Safety Class in terms of industry codes and standards for each category of Plant Conditions. The design requirements reference specific stan- dards and ensure substantial interrelationship with other co
15、des and standards. The content of this standard reflects an attempt to achieve the following objectives: a. To establish a consistent set of requirements for light water reactor nuclear power plants; b. To establish a disciplined, systematic method for defining nuclear safety requirements for nuclea
16、r power plants; c. To establish and delineate the functional nuclear safety requirements for the design of nuclear power plants: d. To be responsive to both the regulatory requirements of the Nuclear Regulatory Commission and the design and technical requirements of industry codes and stan- dards; e
17、. To provide a framework for augmenting these criteria as additional standards are developed within the nuclear industry; and f. To provide a uniform basis for design safety requirements which may be reflected in regulatory documents. The existence of unique plant or site characteristics might requi
18、re the consideration of alternate design concepts. This standard has been developed along functional lines to permit this flexibility. The standard has, however, cited many standards, some of which were still in draft form at the time this document was published. Provisions -,-,- ANS 53.3 83 m 07248
19、38 0098073 b m contained in any draft standard should be considered and used with great discretion. It is strongly suggested that the prospective user fully understand the present status of the referenced standard and major factors on why it might be still in draft form; for example, controversial i
20、ssues should be recognized. A number of considerations under development concurrent with the preparation of this standard are not addressed in this standard. Examples of these considerations include: human factors engineering (HFE), probabilistic risk assessment (PRA), systems interaction, diversity
21、, plant security, emergency response facilities, degraded core, minimizing challenges to engineered safety features, safety goals and considera- tion of costhenefit analysis, and anticipated transients without scram. Subsequent revisions of this standard will address these considerations as appropri
22、ate when they become adequately defined. A designer is not restricted by this standard from proposing or using alternate criteria to ensure adequate nuclear safety. Frequently, a desirable overall result can be obtained by any of several design concepts. The designer may choose from several alternat
23、ives in satisfying the specifics of this standard by the proper consideration of the interrelationship of components and systems within the plant. For example, the PRA approach may be used as an alternative method to evaluate plant design; however, its usefulness is somewhat limited without safety g
24、oals that are currently under development. Portions of this standard were prepared separately under ANS-50 Nuclear Power Plant Systems Engineering and were reviewed individually by ANS-51, ANS-50, and NUPPSCO which replaced ANS-50 during this time. The separate documents that have been incorporated
25、into this standard include the Glossary (CWG-l), Conditions of Design (CWG-O), and Equipment Classification (Ad Hoc Committee on Equipment Classification). The structure of this standard is based on the standard format guide (CWG-4). This standard was approved by NUPPSCO in 1982. This standard and a
26、ll other ANS standards have been written for prospective use. Continuing efforts will be required to augment or modify the criteria in this standard to implement changing licensing requirements, to achieve standardization among the various industry criteria and standards currently being developed, a
27、nd to provide additional clarification or interpretation as appropriate. The ANS-51 PWR Criteria Committee meets periodically to consider revisions or modifications to this standard. Comments, suggestions, and requests for interpretations should be addressed to the Chairman, ANS-51 PWR Criteria Comm
28、ittee, American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Ill. 60525. Working Group ANS-51.1 of the Standards Committee of the American Nuclear Society, had the following membership at the time it developed this standard: G. A. Zimmerman, Chairman, Portland General R. C. Surman,
29、Westinghouse Electric Corporation Electric Company At the time of its approval of this standard, Subcommittee ANS-51 had the following membership: C. J. Gill, Chairman, Bechtel Power Corporation W. Moody, Southern California Edison R. Bone, Stone “Energy,” Part 50, “Licensing of Production and Utili
30、zation Facilities,” Appendix A, “General Design Criteria for Nuclear Power Plants” l. 1.2 Purpose. Incorporating the requirements of this standard provides a degree of assurance that, in their entirety, plants are designed and constructed so that they can be operated without undue risk to the health
31、 and safety of the public. It is intended that this standard lead to attain- ment of this objective by defining existing practices that are consistent with the licensing requirements of the U.S. Nuclear Regulatory Commission (NRC), appropriate industry codes, and good engineering practice. Reference
32、s to regulations, codes, and other standards are included where appropriate. 1Numbers in brackets refer to corresponding numbers in Section 6, References. A designer of a plant has a responsibility, even at the design stage, going beyond conformance to the criteria defined in this standard. In ad- d
33、ition to considering this standard, the NRC regulations, and other published guidance, the designer must ensure that the design bases and expected operational characteristics are sup- ported, to the extent practical, by design analyses, experimental verifications, and comparisons to accepted designs
34、 or experience gained from similar designs. Consideration of alternate or additional criteria and requirements may be necessary to accom- modate unique site characteristics. This standard is written specifically for a PWR nuclear power plant. A PWR plant is based on closed-cycle circuits, utilizing
35、two separate fluid systems that interface at two or more heat ex- changers called steam generators. These circuits are known as the reactor coolant (or primary) system and the power conversion (or secondary) system. The reactor coolant system contains the reactor core, a water-cooled and water-moder
36、ated nuclear assembly that utilizes fissionable fuel. Heat is transferred by the reactor coolant system from the reactor core to the power conversion system at the steam generators. The power conversion system converts thermal energy into electrical energy by means of a turbine generator. Both the r
37、eactor coolant system and the power conversion system are provided with a number of auxiliary systems that supply, service, and control circulated fluids, processes and environ- mental conditions, and remove undesirable by- products, distribute power, and ensure safe conditions, during normal or acc
38、ident conditions. A number of structures are provided to house, contain, protect, and shield both equipment and personnel. For the purpose of this standard, a PWR plant has the following characteristics: a. Solid ceramic fuel enclosed in metallic cladding, b. Fixed geometry for the fuel and coolant
39、(which acts as the moderator), 1 -,-,- ANS 53.3 83 American National Standard ANSUANS-51.1-1983 c. Core and core coolant enclosed in an en- velope of high integrity, and d. Core and core coolant envelope enclosed in a primary containment barrier of high integrity. 2. Definitions This section defines
40、 terms applicable to this standard. active component. A component in which me- chanical movement must occur to accomplish the nuclear safety function of the component. active failure. A malfunction, excluding passive failures, of a component that relies on mechanical movement to complete its intende
41、d nuclear safety function upon demand. Examples of active failures include the failure of a valve or check valve to move to its correct position, or the failure of a pump, fan, or diesel generator to start. Spurious action of a powered component origi- nating within its actuation or control system s
42、hall be regarded as an active failure unless specific design features or operating restrictions preclude such spurious action. An example is the unintended energization of a powered valve to open or close. administrative controls. Rules, orders, instruc- tions, procedures, policies, practices, or de
43、sig- nation of authority and responsibility. best-estimate value. The mean value of the prob- ability density function for the random variable of interest. Where the probability density function is not known, the mean value shall be estimated based on engineering judgment. cold shutdown. The conditi
44、on in which the reactor is subcritical and the reactor coolant system average temperature is below the required tem- perature to permit major maintenance, consistent with technical specification operational limits. common cause failure. Multiple failures of struc- tures, systems, or components as a
45、result of a single phenomenon. 2 L m 0724838 0098077 7 m controlled shutdown and cooldown. A shutdown and cooldown in which the fuel and reactor cool- ant pressure boundary conditions may exceed technical specification limits and implementation of plant emergency procedures may be required. departur
46、e from nucleate boiling (DNB). The onset of the transition from nucleate to film boiling. departure from nucleate boiling ratio (DNBR). The ratio of the heat flux required to cause d e parture from nucleate boiling (DNB) to the actual channel heat flux for given conditions. equipment. A constituent
47、of a component, a component, an assemblage of components, a system, or a structure having at least one func- tion. engineered safety feature. A nuclear safety- related structure, system, or component that serves to control and limit the consequences of releases of energy or radioactivity if an event
48、 were to occur to the extent that the public health and safety might be impaired if these energy or radioactivity releases were not ad- ditionally restrained. event. A condition that deviates from normal operation, i.e., an initiating occurrence or an initiating occurrence plus single failure or co-
49、 incident occurrence combination. fission product barriers. The fuel cladding, reactor coolant pressure boundary, and primary con- tainment. fuel cladding damage. Perforation or excessive distortion or rupture of fuel rod cladding which would permit the release of fission products to the reactor coolant. hot shutdown. In the PWR, the condition in which the reactor is subcr