1、ANSIIANS-57.5-1996 This standard does not necessarily reflect recent industry initiatives for risk informed decision-making or a graded approach to quality assurance. Users jhOUkl consider the use of these industry initiatives in the application of this standard. light water reactors fuel assembly m
2、echanical design and evaluation This standard has been reviewed and reaffirmed by the ANS Nuclear Facilities Standards Committee (NFSC) with the recognition that it may reference other standards and documents that may have been superceded or withdrawn. The requirements of this document will be met b
3、y using the version of the standards and documents referenced herein. It is the responsibility of the user to review each of the references 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
4、 referenced in this standard should be evaluated and documented. ANSIIANS-57.5-1996 American National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluation Seme t ariat American Nuclear Society Prepared by the American Nuclear Society St andards Commit tee Working Group AN
5、S-57.5 Published by the American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA Approved February 8, 1996 by the American National Standards Institute, Inc. American Nat ional Standard Designation of this document as an American National Standard attests that the prin
6、ciples of openness and due process have been followed in the 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 procedu
7、res are accredited by the American National Standards Institute, 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 America
8、n National Standard is intended to aid industry, consumers, governmental 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, proc
9、esses, or procedures not conforming to the standard. By publication 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 app
10、roval and publication. Changes, if any, occurring through 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 accordance with established procedures. Users of this standard a
11、re cautioned to determine the validity of copies in their 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 interpretati
12、on should be sent to the Standards Department at Society Headquarters. 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 Headquar- ters. Pu
13、blished by American Nuclear Society 555 North Kensington Avenue, La Grange Park, Illinois 60526 USA Copyright O 1996 by American Nuclear Society. Any part of this standard may be quoted. Credit lines should read “Extracted from American National Standard ANSU4NS57.5-1996 with permission of the publi
14、sher, the American Nuclear Society.“ Reproduction prohibited under copyright convention unless written permission is granted by the American Nuclear Society. Printed in the United States of America Foreword (This Foreword is not a part of American National Standard for Light Water Reactors Fuel Asse
15、mbly Mechanical Design and Evaluation, ANSUANS57.5-1996.) This American National Standard provides a procedure for determining the mechanical adequacy of fuel assembly designs for light water nuclear reactors. Specific require- ments for design and specific rules for demonstrating compliance are als
16、o included. It is not the intent of this standard to endorse any design feature, material, material property information, analysis method, or other procedure, or in any way to inhibit development or innovation in any of these areas. However, this standard does include. certain requirements intended
17、to ensure that the methods or material properties which are used are appropriate and adequately documented. Suggestions for improvement of this standard are welcome. They should be sent to the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60526. The membership of Wo
18、rking Group ANS-57.5, at the time it submitted this revision of this standard, was as follows: R. H. Ripley, Chairman, Union Electric Company J. A. Nevshemal, Raytheon UE the word “should denotes a recommendation; and the word “may“ denotes permission, neither a requirement nor a recom- mendation. 4
19、. Compliance Design documentation shall be prepared to show how the criteria and requirements of this stand- ard are satisfied. Provisions for dissemination of 1 American National Standard ANSI/ANS57.5-1996 and access to such design documentation are beyond the scope of this standard. 5. Design and
20、Evaluation 5.1 Design Conditions. Design condition events in this standard generally correspond to Plant Conditions as presented in a proposed American National Standard under development. The designer shall specify the number of cycles or frequency of occurrence used for the various events in the d
21、esign conditions? In addition, the design conditions should be applied as appropri- ate during long term post irradiation storage in the reactor site spent fuel storage pool, or other spent fuel storage facilities. 5.1.1 Condition I - Normal Operation and Operational Transients. Condition I events a
22、re those that are expected frequently or regularly in the course of power operation, refueling, mainte- nance, or maneuvering of the plant. 5.1.2 Condition II - Events of Moderate Frequency. Condition II events are those that could occur in a calendar year for a particular plant and that could resul
23、t in reactor shutdown. The reactor is expected to be capable of a return to power without special fuel inspection or repair procedures being required. 5.1.3 Condition III - Infrequent Events. Condition III events are those that could occur during the lifetime of a particular plant. It is expected th
24、at such an event could result in some damage to a fuel assembly that might necessitate repair or replacement of the assembly before nor- mal operation can be resumed. 5.1.4 Condition IV - Limiting Faults. Condi- tion IV events are those that are not expected to occur, but are postulated because thei
25、r conse- quences would include the potential for release of significant amounts of radioctive material. 5.2 Functional Requirements. This section provides the functional requirements that shall be addressed for the design condition events specified in 5.1. The fuel assembly shall be de- Proposed Ame
26、rican National StandardNuclearSafety Design Criteria for Light Water Reactors. ANSI/ANS50.1; assigned correspondent Ralph C. Suman, Westinghouse Electric Corpora tion. 2!3ee Appendix A for typical lists of design condition events. signed to fulfill the specific functional require- ments throughout i
27、ts anticipated operating life- time. It is not intended that all of these require- ments be met for all design condition events. The designer shall specify the applicable functional requirements for each design condition. The specific features of a design that fulfill the func- tional requirements c
28、ould be different for differ- ent designs. Likewise, the specific methods by which the fuel assembly is shown to fulfill its functional requirements could vary for different designs. The fuel assembly design shall satisfy the following requirements: Provide and maintain acceptable fuel geom- etry an
29、d position axially and radially, so that the fuel rods are located correctly within the fuel assembly and the fuel as- sembly within the core. Provide for acceptable coolant flow and heat transfer. Provide a barrier to separate the fuel and contain fission products. Allow for axial and radial expans
30、ion of the fuel rods, the fuel assembly, and contiguous reactor internals. Provide self-support, i.e., be free standing when required and offer well-defined resist- ance to distortion by lateral and axial loads. Withstand action by fluid forces, i.e., accom- modate the effects of vibration, wear, li
31、ft, cavitation, pressure pulses, and flow insta- bili tie s. Provide for the presence of control elements, i.e., provide physical guidance to control rods or blades; accept the presence of burn- able poison rods or chemical shims; accom- modate the effects of flux, temperature and pressure gradients
32、; endure wear and impact associated with control element motion. Provide for in-core instruments and neutron sources. Accommodate chemical, thermal, mechani- cal, and irradiation effects on materials, e.g., corrosion, hydriding, irradiation em- brittlement, expected interactions, fuel densification,
33、 creep and relaxation during reactor service, and post-irradiation storage 2 in the reactor site spent fuel storage pool, or other spent fuel storage facilities. (10) Provide for handling, shipping, and core loading, i.e., provide gripping and contact locations, hold-down springs, or other nec- essa
34、ry hardware, including provisions for loads and compatibility with interfacing equipment in the reactor. (11) Provide mutual compatibility for all fuel assemblies within the core, including reload, reconstituted, and partially spent fuel as- semblies. Compatibility includes fitup and cross-flow in o
35、pen lattice designs. Nuclear compatibility is beyond the scope of this standard. (12) Provide features to identify proper rotation- al positioning in the core, and for placement of readable assembly identification number. 5.3 Design Parameters. Design parameters used to demonstrate design adequacy s
36、hall be identified and justified. These parameters are usually in the form of material properties, dimen- sional characteristics, or physical response phe- nomena that are necessary to describe or evalu- ate fuel assembly behavior. These parameters shall be justified by generally accepted engineer-
37、ing methods, such as reference to test and experi- mental data, experience, analysis, use of refer- ence material, and correlations. The designer shall identify parameters and justify their application as employed in the evaluation. It is recognized that not all the parameters and models are necessa
38、rily treated explicitly in design. For example, interaction between fuel and cladding might not involve a calculation of pellet cracking. Likewise, some portion of fuel swelling could be implicit in the densification model. Wherever parameters are implicitly handled in design, it is sufficient for t
39、he designer to point this out. The parameters addressed by the designer shall include the items listed below. 5.3.1. General Environmental Conditions (i) Coolant temperature (2) Coolant pressure (3) Coolant flow rate American National Standard ANSI/ANS57.5-1996 Coolant chemistry Neutron flux Flow, t
40、emperature, and pressure varia- tions Core internals motion Spent fuel storage conditions Accelerations due to shipping, handling, seismic, transient, and accident conditions. 5.3.2 Fuel and Control Material 5.3.2.1 Physical Features (1) Dimensions (2) Geometry (3) Density (4) Surface roughness. 5.3
41、.2.2 Chemical Composition 5.3.2.3 Material Properties (1) Thermal parameters (a) Thermal conductivity coefficients (b) Thermal expansion coefficients . (c) Specific heats (d) Phase-structure transformations (e) Melting temperatures. (2) Mechanical parameters (a) Youngs modulus (b) Poissons ratio (c)
42、 Tensile strength (d) Compressive strength. (3) Metallurgical parameters (a) Grain size and distribution 3 American National Standard ANSI/ANS67.5-1996 (b) Pore size (11) Surface condition, including crud (c) Pore size distribution (12) Closures (d) Pore type (openclosed). (13) Identification symbol
43、s. 5.3.2.4 Models and Correlations 5.3.3.2 Chemical Composition. Material (1) Pellet cracking (2) Fission and sorbed gas release designation of the fuel rod subcomponents. 5.3.3.3 Material Properties for Cladding and Other Subcomponents as Appropriate (3) Creep (1) Thermal parameters (4) Irradiation
44、 induced swelling (a) Thermal conductivity coefficients (5) Densification (b) Thermal expansion coefficients (6) Thermal conductivity, including porosity (c) Specific heats (d) Phase-structural transformations. factors (7) Thermal expansion (2) Mechanical parameters (8) Melting. (a) Youngs modulus 5
45、.3.2.5 Performance and Mechanical (b) Yield strength Limits. Performance and mechanical limits of fuel and control material are as specified in 5.4. (c) Ultimate strength 5.3.3 Fuel Rod. The fuel rod shall be treated as a system, with all subcomponents except fuel and control materials (covered in 5
46、.3.2) addressed. (d) Ductility (e) Density (0 Poissons ratio. 5.3.3.1 Physical Features (1) Length (2) Diameter (3) Cladding wall thickness and thickness variations (4) vality (5) Fuel stack heights (6) Surface roughness, including scratches (7) Void and plenum volumes (8) Initial internal pressure
47、(9) Fill gas composition , (10) Inclusion of other nonfuel components (i.e., spacer pellets, getters, springs) (3) Metallurgical parameters (a) Grain size (b) Anistropy factors (cl Texture coefficients (d) Hydride orientation (4) Chemical parameters (a) Corrosion rates (b) Hydrogen pickup and embrit
48、tlement (c) Surface preparation prior to irradia- tion. 5.3.3.4 Models and Correlations (1) Void volume used for gas accommodation 4 American National Standard ANSI/ANS67.6-1996 Creep, creep collapse, creep rupture Thermal performance Water to clad heat transfer coefficient Thermal expansion (radial
49、, circumfer- ence, and axial) Fuel to clad gap conductance, including gas compositions and gas thermal con- ductivities, and the contributions from non-volatile fission products Bowing Irradiation growth, including anisotropic correlations Stress relaxation (10) Fatigue W) Waterlogging (12) Fuekladding interaction (13) Stress corrosion cracking (14) Corrosion (15) Hydriding (16) Axial gap formation in the fuel stack (17) Plastic deformation (18) Stored energy (19) Fretting (20) Stress rupture (21) Crud-induced localized corrosion 5.3.3.5 Performance and Mechanical Limits. Performanc
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