ANS Light Water Reactors Fuel Assembly Mechanical Design & Evaluation

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 mechanical 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Standards Committee (NFSC) with the recognition that it may reference other standards and documents that may have been superceded or withdrawn.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 by 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 referenced in this standard should be evaluated and documented.

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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 ANS-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.



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American Nat ional Standard

Designation of this document as an American National Standard attests that the principles 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 procedures 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 American 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, processes, 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. T h e content of this standard reflects acceptable practice at the time of its approval 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 are 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 interpretation 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.

Published 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 publisher, 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



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Foreword (This Foreword is not a part of American National Standard for Light Water Reactors Fuel Assembly 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 requirements for design and specific rules for demonstrating compliance are also 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 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 Working Group ANS-57.5, a t the time it submitted this revision of this standard, was as follows:

R. H. Ripley, Chairman, Union Electric Company J. A. Nevshemal, Raytheon UE&C

The American Nuclear Society's Nuclear Power Plant Standards Committee (NUPPSCO) had the following membership at the time of its approval of the standard:

W.H. D'Ardenne, Chairman M.D. Weber, Secretary

R.E.Ailen ............................................................. UE&CNuclear (for the Institute of Electrical and Electronics Engineers, Inc.)

P.L. Ballinger ............................................. Nebraska Public Power District F. Boorboor .............................................. Nuclear Placement Services, Inc. J.C.Bradford ................................................................. Bechtel R.H. Bryan, Jr. ............................................... Tennessee Valley Authority T.W.T. Burnett .......................................... Westinghouse Electric Corporation J.D. Cohen ........................................ Westinghouse Savannah River Company J.B. Cotton ................................................ Philadelphia Electric Company W.H. D'Ardenne ....................................................... DAE Enterprises

Commonwealth Edison Company M. Drouin ............................................ US. Nuclear Regulatory Commission P.H. Hepner ............................................ ABB/Combustion Engineering, Inc.

.......................................................... G.E. Nuclear Energy ................................................ Florida Power & Light Company

J.F. Mallay .................................................... Liberty Consulting Group C.H. Moseley, Jr. ..................................... Performance Development Corporation J.A. Nevshemal ........................................................ Raytheon UE&C

.............................................. Babcock & Wiiwx Company W.C. Ramsey, Jr. ......................................... Southern Company Services, Inc. W.B. Reuland ........................................... Mollerus Engineering Corporation R.F. Sacramo ............................................... Haliiiurton NUS Corporation J.C. Saldarini ........................................... Raytheon Engineers & Constructors J. Savy .......................................... Lawrence Livermore National Laboratory R.E. Scott ............................................................ Scott Enterprises D.J. Spellman ............................................. Oak Ridge National Laboratory S.L. Stamm ....................................... Stone & Webster Engineering Corporation J.D. Stevenson ................................................... Stevenson & Associates C.D. Thomas, Jr. ......................................... Yankee Atomic Electric Company G.P. Wagner ............................................. Commonwealth Edison Company N.Weber ................................................................. Consultant G J . Wrobel ........................................... Rochester Gas & Electric Corporation

(for the American Nuclear Society) L.E. Davis ...............................................

R A Hili J.T. Luke

W.N. Priilaman

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Page

1 . Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2 . Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

3 . Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

4 . Compliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

5 . Design and Evaluation .......................................... 2 5.1 Design Conditions .......................................... 2 5.2 Functional Requirements .................................... 2 5.3 Design Parameters ......................................... 3 5.4 Limits and Margins ........................................ 6 5.5 Specific Requirements for Design .............................. 7

6 . Documentation Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 6.1 Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 6.2 Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

7 . References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

Appendices Appendix A Design Condition Events ............................. 13 Appendix B Illustration of the Use of the Standard . . . . . . . . . . . . . . . . . . 17

Tables Table 1 Matrix ........................................... 18



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Light Water Reactors Fuel Assembly Mechanical Design and Evaluation 1. Scope

This standard sets forth a series of design conditions and functional requirements for the design of fuel assemblies for light water cooled commercial power reactors. It includes specific requirements for design, as well as design criteria to ensure adequate fuel assembly performance. The standard establishes a procedure for per- forming an evaluation of the mechanical design of fuel assemblies. It does not address the various aspects of neutronic or thermal-hydraulic performance except where these factors impose loads or constraints on the mechanical design of the fuel assemblies.

2. Purpose

The purpose of this standard is to establish a set of design requirements for the mechanical design of initial core or reload fuel assemblies and for the evaluation of that design.

These design requirements include:

(i) A comprehensive set of functional requirements for fuel assemblies

(2) A method for selecting the specific events in each of the design conditions

(3) A comprehensive list of parameters, including material properties, chemical reactions, irradiation effects, and failure modes, which could affect the capability of fuel assemblies to satisfy one or more functional design requirements

(4) A method to achieve two goals: define which parameters and assumptions affect the capability of the fuel assembly to fulfill each functional requirement under each postulated event, and establish an appropriate limit for the defined parameters and assumptions which ensures that some aspect of a functional requirement for that event is met

(5) A procedure to document that the fuel assembly design has been evaluated in accordance with the design limits discussed in 5.4

(or similar) and has been shown to fulfill each functional requirement for each event.

3. Definitions

designer. The organization that has the responsibility for preparing the fuel assembly design.

design parameters. Material properties, dimensional charàcterizations, or physical response phenomena necessary to describe or evaluate fuel assembly behavior.

event., A describable situation that must be accounted for in design.

fuel assembly. The smallest modular unit com- prised of individual fuel rods and associated integral component parts for handling, control, support, and maintenance of the unit's geometry. For boiling water reactors, the channel that encloses the fuel bundle and the channel fastener is included as part of the fuel assembly for design purposes.

functional requirement. One of several required capabilities of a fuel assembly that is necessary to meet its design function.

limit. A bounding value of a variable or parameter, which is established to ensure that one or more aspects of a functional requirement are satisfied.

margin. A quantitative relationship between a design evaluation result for a given event and a limit associated with a functional requirement.

shall, should, and may. The word "shall" denotes a requirement; the word "should' denotes a recommendation; and the word "may" denotes permission, neither a requirement nor a recommendation.

4. Compliance

Design documentation shall be prepared to show how the criteria and requirements of this standard are satisfied. Provisions for dissemination of

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American National Standard ANSI/ANS57.5-1996

and access to such design documentation are beyond the scope of this standard.

5. Design and 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 design conditions? In addition, the design conditions should be applied as appropriate 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 are those that are expected frequently or regularly in the course of power operation, refueling, maintenance, 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 result 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 that such an event could result in some damage to a fuel assembly that might necessitate repair or replacement of the assembly before normal 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 their consequences would include the potential for release of significant amounts of radioáctive 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 American 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 requirements throughout its anticipated operating lifetime. It is not intended that all of these requirements 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 functional requirements could be different for different 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 geometry and position axially and radially, so that the fuel rods are located correctly within the fuel assembly and the fuel assembly 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 expansion 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 resistance to distortion by lateral and axial loads.

Withstand action by fluid forces, i.e., accommodate the effects of vibration, wear, lift, 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 burnable poison rods or chemical shims; accommodate the effects of flux, temperature and pressure gradients; endure wear and impact associated with control element motion.

Provide for in-core instruments and neutron sources.

Accommodate chemical, thermal, mechanical, and irradiation effects on materials, e.g., corrosion, hydriding, irradiation embrittlement, expected interactions, fuel densification, creep and relaxation during reactor service, and post-irradiation storage



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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 necessary 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 assemblies. Compatibility includes fitup and cross-flow in open 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 shall be identified and justified. These parameters are usually in the form of material properties, dimensional characteristics, or physical response phenomena that are necessary to describe or evaluate fuel assembly behavior. These parameters shall be justified by generally accepted engineering methods, such as reference to test and experimental data, experience, analysis, use of reference 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 necessarily 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 the 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


Coolant chemistry

Neutron flux

Flow, temperature, and pressure variations

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.

5.3.2.2 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) Young's modulus

(b) Poisson's ratio

(c) Tensile strength

(d) Compressive strength.

(3) Metallurgical parameters

(a) Grain size and distribution



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(b) Pore size (11) Surface condition, including crud

(c) Pore size distribution (12) Closures

(d) Pore type (openíclosed). (13) Identification symbols.

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 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) Young's modulus

5.3.2.5 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 5.3.2) addressed.

(d) Ductility

(e) Density

(0 Poisson's ratio. 5.3.3.1 Physical Features

(1) Length

(2) Diameter

(3) Cladding wall thickness and thickness variations

(4) ûvality

(5) Fuel stack heights

(6) Surface roughness, including scratches

(7) Void and plenum volumes

(8) Initial internal pressure

(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 embrittlement

(c) Surface preparation prior to irradiation.

5.3.3.4 Models and Correlations

(1) Void volume used for gas accommodation



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Creep, creep collapse, creep rupture

Thermal performance

Water to clad heat transfer coefficient

Thermal expansion (radial, 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. Performance and mechanical limits of fuel rods are as specified in 5.4.

5.3.4 Fuel Assembly

6.3.4.1 Physical Features

(1) Dimensional characteristics required for interfaces among fuel assemblies, reac-

tor internals, storage racks, and other components

(2) Means of radial support

(3) Means of axial support

(4) Grid and channel dimensions

(5) Means of positioning

(6 ) Means of handling

(7) Integral control rods and instruments.

5.3.4.2 Chemical Compositions. Alloys for all assembly structural components.

5.3.4.3 Material Properties

(1) Thermal Parameters

(a) Thermal conductivity coefficients

(b) Thermal expansion coefficients

(c) Microstructural transformations

(d) Specific heat.

(2) Mechanical parameters

(a) Ultimate strength

(b) Poisson's ratio

(c) Young's modulus

(d) Yield strength

(e) Ductility

(0 Fatigue strength

(g) Density

(h) Impact strength.

(3) Metallurgical parameters . .

(a) Grain size ~

(b) Anistropy factors

(c) Texture coefficients.



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(4) Chemical parameters

(a) Corrosion rates

(b) Hydrogen pickup and embrittlement.

5.3.4.4 Models and Correlations

Wear

Vibration (frequency and amplitude)

Stress relaxation

Deformation (permanent)

Irradiation and temperature induced materials growth and property changes

Cladding to support compressive forces

Assembly hold-down forces

Creep

Crud buildup

Crack propagation

Bowing.

5.3.4.5 Performance and Mechanical Limits. Performance and mechanical limits of fuel assemblies are as specified in 5.4.

5.4 Limits and Margins

5.4.1 Limits. Limits shall be established by the designer for the purpose of demonstrating that the functional requirements (5.2) pertinent to a given design condition (5.1) are satisfied. These limits are established for the purpose of ensuring that there is a sufficiently high probability of meeting the functional requirements.

The following limits for structural components shall be specified as appropriate for each design condition.

6.4.1.1 For time-independent effects, the primary and secondary stress (S) shall be demonstrated to be less than its ultimate value, Su.

5.4.1.2 For components subjected to cyclic loads, the sum of the terms defined by the number of cycles in each stress range or strain range (N) divided by the corresponding number of cycles to failure (N,) shall be less than 1.0.

5.4.1.3 For time-dependent effects, the summation of actual time (t) at a given stress level divided by the time of failure (t,) at that stress level shall be less than 1.0. Also, the summation of creep strain incurred (E') divided by the creep strain to failure

(ecf) shall be less than 1.0.

5.4.1.4 Any load (P) with potential for Caus- ing structural instability shall be demonstrated to be less than the critical value of that load (PJ, which would result in crippling of the structure.

P < P,

5.4.1.5 The integral of cyclic growth rate

in a component shall be demonstrated to result in a final crack size that is less than the critical value of crack size (a,) for sudden failure.

where

a o = initial length of maximum acceptable defect for growth

s < su No = number of load cycles necessary to

initiate growth of defect



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American National Standard ANSIL4NS57.5-1996

= crack growth rate, extension per load cycle

N d = design number of load cycles

8, = critical crack length for sudden fracture.

5.4.1.6 The calculated maximum stress intensity (KI ) from a single tensile loading of a component shall be less than the critical stress intensity (Klc) calculated for the same conditions.

5.4.2 Margin. Sufficient margin shall be demonstrated so that inherent uncertainties in experimental or analytical predictions do not result in failure to meet the particular functional requirement. The particular method used to establish that an adequate margin exists is optional, but it shall be selected from one or more of the following:

Probability analyses in which the vari- ances of independent parameters are statistically combined.

Sensitivity analyses in which the variance of the dependent parameter is predicted as a function of the tolerance ranges of the independent variables.

Worst-case analyses in which each independent variable is deliberately biased to produce the most adverse predicted dependent parameters.

Combined analyses in which certain independent variables are worst-case and others are statistically determined or nominally chosen and weighted for sensitivity.

Reference to experimental data or operational performance which clearly verifies the adequacy of the design for fulfilling a specific functional requirement for a given design condition.

6.5 Specific Requirements for Design. A comprehensive set of fuel assembly functional requirements and design considerations is pre-

sented in 5.2 through 5.4. These are represented without specific instructions as to how the various considerations should be applied so as not to place unnecessary constraints upon the design. However, certain considerations for designing fuel assemblies are necessary for all designs and are set forth beIow as mandatory requirements.

5.5.1 Material Properties. The following general criteria shall be met in the selection of material properties for specific evaluations:

(1) Component performance shall be evaluated for both the non-irradiated and irradiated material properties.

(2) Irradiation-induced densification shall be determined through evaluation of fuel redistribution, power peaking, and stored energy.

(3) The effect of temperature on individual material properties shall be taken into account by use of properties appropriate to the expected component temperature.

(4) For material property data that are corre- lated with neutron flux or fluence, the neutron flux energy spectrum shall be addressed.

5.5.2 Corrosion. The following criteria shall be applied in the evaluation of the effect that corrosion of fuel components has upon performance.

(1) Corrosion behavior characteristics of fuel assembly materials shall be obtained under conditions representative of the reactor environment.

(2) The effect of corrosion and crud film buildup on heat transfer surfaces shall be addressed in the calculation of pellet and cladding temperatures.

(3) "he effects of fabrication processes such as cold work, heat treatment, stress relief, and welding shall be taken into account on corrosion behavior.

5.5.3 Control of Hydriding of Zirconium Alloy. The maximum acceptable hydrogen content in the fuel and burnable poison rod shall be determined in order to minimize clad perforations from the mechanism of primary hydriding.



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"he effects of a "getter" shall be evaluated if it is added to the fuel for reduction of hydriding, including the potential for mechanical and chemical interactions with other fuel rod components and with reactor coolant. (A "getter" is material added to a fuel rod that effectively competes with the cladding for free hydrogen that may be present in the rod.)

5.5.4 Fretting Corrosion. The spacer grids shall be used to control relative motion between the rods and the support surfaces so that wear of the cladding at these surfaces does not penetrate the cladding o r violate the capability of the cladding to withstand operating loads.

The adequacy of the spacer grid design and its position within the fuel assembly shall be established by test or analysis under conditions consistent with the intended reactor's operating coolant temperature, pressure, flow rate, and chemistry.

For the establishment of an initial design, the following criteria shall apply:

Known or predicted excitation frequencies shall be taken into account when estab- lishing fuel rod grid spacing.

The test or analysis used to demonstrate the adequacy of the grid design shall include grid cells whose interference with the cladding outside diameter is adjusted to account for sources of the predicted reduction of restraint, i.e., grid tab stress relaxation, cladding diametrical creep, differential thermal expansion, and lateral grid irradiation-induced growth. "he range of initial preset forces specified for new assemblies shall be considered.

The redistribution of flow within and between fuel assemblies shall be considered. Examples of regions of redistribution are end fittings and grids. Cross-flow can be caused by jetting of fluid from core baffle joints.

T h e designer shall analyze or test for cladding wear and its effect on related analyses.

5.5.5 Fuel Assembly Hold-down Force. For designs that utilize hold-down mechanisms (e.g., springs) to accommodate hydraulic loads, the

designer shall show that an adequate hold-down force exists based on the following:

Stress relaxation for hold-down springs.

Maximum expected Condition I flow rate.

Fuel assembly pressure drop, including possible increases due to crud deposition within the assembly.

Combination of dimensional tolerances of the fuel assembly and supporting struc- tures.

Differential thermal expansion between the fuel assembly and reactor internals.

Fuel assembly irradiation-induced growth.

5.5.6 Fuel Rod Axial Growth Allowance. Sufficient axial clearance between the fuel rods and the fuel assembly structure shall be provided to accommodate expected dimensional changes of the components during the design life of the fuel assembly.

Demonstration of compliance with this requirement shall take into account the following:

(1) Differential thermal expansion between the fuel rods and the fuel assembly structure. '

(2) The effect of tolerances.

(3) Differential irradiation-induced growth between fuel rods and between fuel rods and fuel assembly structure. This allowance should include evaluation of axial extension of cladding induced by interaction between fuel and cladding.

(4) The effect of fuel assembly structure axial compression and creep.

5.5.7 Fuel Rod Internal Pressure. The variations in pressure that occur over the life of the fuel rods shall be taken into account. The fuel rod performance is significantly affected by internal pressure from creep, ballooning, or rod collapse. Calculations of fuel rod internal pressure shall account for the following effects:



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American National Standard A"ANS57.5-1996

Differential thermal expansion (axial and (4) Combinations of component dimensional radial) between pellets and cladding. tolerance and fill gas pressure tolerances.

Irradiation-induced swelling of the fuel pellets.

The accumulation of nonvolatile fission products.

Solubility of the fill gas in the fuel material.

The higher temperature of gas contained in pellet end dishes, pellet cracks, and pellet open porosity, than the temperature of the gas in the annulus between pellets and cladding.

6.5.9 Rod Bow. The amount of acceptable bow of the fuel rod shall be determined. It has been observed that some fuel rods (and burnable poison rods) bow during operation. As a result, the lateral spacing between fuel rods can vary. The following effects shall be considered:

(1) The local variation of the fuel-moderator volume ratio on the peak local fuel rod power level.

(2) Variations in coolant subchannel diver- sions on the departure from nucleate boiling (DNB) margin or critical heat flux.

The release of gaseous fission products (3) Control element operation for designs in from the fuel material. which the expected magnitude of fuel rod

bow would be sufficient to cause the fuel Irradiation-induced densification in the rod to intrude into a control element fuel material. path.

Irradiation-induced growth and creep of the cladding, as well as thermal creep of the cladding.

Expected variations in initial fill gas pressure and component dimensions.

5.5.10 Fuel Assembly Bow and Twist. The amount of acceptable fuel assembly bow and twist shall be accounted for in the design in accordance with the following:

(1) It shall be shown that the maximum expected bow and twist can be accommo-

(10) Release of absorbed gases from the fuel dated by the handling equipment and material. fuel storage facilities.

5.5.8 Cladding Collapse. The fuel rod shall be designed so that the cladding is not suscepti- ble to collapse from the long-term effects of cladding creep. The designer shall specify the criterion used for collapse. Cladding collapse refers to the dimpling of cladding into short, unsupported gaps in the fuel column. Demon- stration of compliance with this requirement shall take into account the following:

The effect of fuel rod burnup and power level on internal pressure, including the effect of a conservatively low assessment of fission gas release.

Irradiation-induced densification of fuel pellets and the solubility of the fill gas in fuel.

The range of power histories to which the rod is likely to be subjected.

(2) The effect of the fuel assembly bow and twist on control rod motion (e.g., through friction drag) shall be assessed.

(3) The effect of fuel assembly bow and twist on local power and coolant flow distribution shall be assessed.

5.5.1 1 Component Cooling Flow. Where other components (such as control rods, poison rods, neutron sources, or instrumentation) are included within the fuel assembly, the design shall provide for sufficient cooling of these components. The following shall be considered in the establishment of cooling adequacy:

(1) Minimum expected pressure head.

(2) Contributors to maximum flow resistance.

(a) Component dimensional tolerances



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American National Standard ANSVANS67.6-1996

(b) Differential thermal expansion and irradiation induced dimensional changes

(c) Component heat generation

(d) Surface roughness, component corrosion, and crud deposition.

5.5.12 Fuel Handling. All expected fuel handling activities shall be provided for in the fuel assembly design in accordance with the following:

Each fuel assembly shall be marked in such a way that its identity and orientation in the core can be verified throughout the life of the assembly. American National Standard for Fuel Assembly Identification, ANSUANS-57.8-1994 [il3 defines a system of serialization which may be used to satisfy this requirement.

The fuel assembly shall be capable of sus- taining the loads induced by normal handling operations, including shipment before and between usage.

The fuel assembly shall be designed to facilitate core loading and unloading without damage.

The fuel assembly shall be capable of sus- taining the expected loads induced by possible reconstitution operations (i.e., the partial disassembly and subsequent reassembly of irradiated fuel assemblies for purposes of inspection, repair, o r other reason).

5.5.13 Fuel-Cladding Interactions and Stress Corrosion Cracking of the Cladding. Experience has shown that the cladding of fuel rods might be breached during normal modes of reactor operation prior to the designed end-of-life condition. In some cases these failures have been attributed to stress corrosion cracking initiated on the inner surface of the clad due to a combination of local clad stress, produced by mechanical interaction between the fuel pellet and the clad, and the presence of certain fission products. It is beyond the scope of this standard to prescribe the means by which the fuel designer assesses the

%umbers in brackets refer to corresponding numbers in Section 7, References.

10

failure mechanisms, or to require that fuel be designed to limit failures to a particular number of rods. However, the methods used in the assessment of these failure mechanisms, the design features, reactor loading and power maneuvering limitations, and fuel duty, that lead to an accept- ably low probability of failure shall be stated.

5.5.14 Analytical Evaluations of Stresses and Strain. For those components in a fuel assembly for which mechanical integrity is to be demonstrated by means of stress analysis, the following requirements shall apply:

For components subject to multi-axial stress conditions, the analysis shall use one of the recognized methods for com- bining such stresses, such as the maximum strain energy or maximum resolved shear stress, and the criteria used for determining acceptable results shall be identified. (See 5.4.)

For components that are subjected to cyclic loading, the cumulative effect shall be determined. The method used shall be identified. (See 5.4.)

For structural components that are subject to significant creep strains as a result of operational loadings, the magnitude of the resultant creep strain shall not be sufficient to produce rupture of the component. (See 5.4.)

For components that are subjected to both cyclic loadings and creep strains, an acceptance criterion that takes both con- stant and cyclic loads into account shall be established. (See 5.4.)

5.5.15 Debris in Coolant. The design of fuel assemblies shall include tolerance for metallic debris in the reactor coolant, Experience has shown that small metallic debris in the coolant can be a significant source of fuel failures.

5.5.16 Crud Induced Localize ' d Corrosion. The potential effects of crud deposits which can lead to increased corrosion rates on cladding surfaces shall be addressed. Such deposits can interact with the clad chemically or can cause altered local heat transfer characteristics. Ai- though crud deposits can be minimized in reactors by appropriate selection of reactor coolant system materials and plant chemistry, the fuel



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American National Standard ANSI/ANS-57.5-1996

assembly design should be as tolerant as possible to deleterious crud deposits.

6.5.17 Interim Storage. The effects of post- irradiation fuel assembly handling-i.e., between the reactor core and storage in the at-reactor spent fuel storage pool, or insertion into an independent spent fuel storage installation (wet or dry type)-should be considered.

5.5.18 Seismic and LOCA Loads. Evalua- tions shall be performed to ensure that the mechanical loads resulting from a combination of a seismic event and a loss of coolant accident do not cause fuel assembly damage to such an extent that control rod insertion is prevented or that coolability cannot be mair~tained.~

6. Documentation Requirements

6.1 Objectives. This section sets forth requirements for design documentation to satisfy Section 4, Compliance. The documentation shall demonstrate that the fuel assembly design meets this standard with respect to:

(i) Defining the function and desired performance under stated conditions for both the fuel assembly as a unit and for individual components as appropriate.

(2) Defining and documenting the set of criteria that provide assurance of achiev- ing the stated function and performance requirements.

(3) Demonstrating that the design does, in fact, meet the criteria.

6.2 Content

6.2.1 Design Conditions. The specific conditions used as a basis for fuel assembly design shall be identified according t o the Design Condi- tions presented in 5.1. Specific events, or combinations of events for each condition, shall be described in terms of cause, such as plant process conditions or a postulated accident.

6.2.2 Functional Requirements. The fuel assembly functional requirements appropriate to

4See Appendix A of Section 4.2 of Shndard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition “REG-0800). Available from the US. Nuclear Regulatory Commission, Washington, DC 20555.

each design condition shall be established and documented. The functional requirements shall include the requirements in 5.2 where appropriate for the specific fuel assembly design.

6.2.3 Design Description. A description of the fuel assembly design shall be presented. The important components, dimensions, physical features, and chemical and material properties of the fuel assembly relevant to function and performance shall be presented. The physical features, chemical composition, materials properties, and model subsections of 5.3 shall be consulted as a guide for important items to be presented.

6.2.4 Fuel Assembly Loading. The quantitative loading component or response for specific design conditions shall be established. Pertinent environmental parameters, presented in 5.3.1, shall be explicitly stated for each loading combination to be evaluated. The effects of the chosen design conditions shall be identified and included in the design package.

6.2.5 Design Limits. Specific design limit5, together with the associated justification, which assure meeting the fuel assembly functional requirements of each design condition, shall be established and appropriately recorded in the design .documentation. A discussion of limits is presented in 5.4.1.

6.2.6 Design Evaluation. The documentation shall present the detailed evaluation and the resulting margin to the limits established. The margin shall be established, as per 5.4.2, to ensure that the functional requirements are satisfied. The methods for design evaluation shall account for appropriate parameters, analysis techniques, experimental testing, or operational results. A listing of all reference material utilized in the evaluation shall be provided.

7. References

Cl1 American National Standard for Fuel As- sembly Identification, ANSI/ANS-57.8-1994. Available from American Nuclear Society, 555 N. Kensington Avenue, La Grange Park, IL 60526.

Only the standards explicitly referenced in this document qualify as references. Subsequent revisions of these standards shall not be substi- tuted.



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Appendix A (Tliis Appendix is not a part of American National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluation, ANSI/ANS57.5-1996. but is included for information purposes only.)

Design Condition Events

Condition I-Normal Operation and Operational Transients

Condition I events are those that are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, Condition I events are accommodated with margin between any plant parameter and a value of that parameter that would require either automatic or manual protective action. Inasmuch as Condition I events occur frequently or regularly, they must be considered from the point of view of affecting the consequences of Condition II, III, and IV events. In this regard, analysis of each event is based on a set of initial conditions corresponding to the expected mode of plant operation up t o the time of the subject event.

A typical list of Condition I events is given below:

1. Steady state and shutdown operations (a) Power operation (b) Startup (c) Hot shutdown (subcritical, residual heat removal system isolated) (d) Cold shutdown (subcritical, residual heat removal system in operation) (e) Refueling, including fuel handling (0 Standby (less than 10% full power).

2. Operational maneuvers (a) Plant heatup and cooldown (b) Load changes.

3. Operation with permissible deviations. Various deviations, which may occur during continued operation as permitted by the plant technical specifications, must be considered in conjunction with other operational modes. These include: (a) Operation with components or systems out of service (such as power operation with reactor

coolant pump out of service) (b) Leakage from &el with cladding defects (c) Activity in the reactor coolant

(i) fission products (2) corrosion products (3) tritium

allowed by the technical specifications. (d) Operation with steam generator leaks (PWR) or condenser leaks (BWR) up to the maximum

4. Preoperational and operational testing of systems with fuel in place.

Condition II-Events of Moderate Frequency

Condition II events are those that could occur in a calendar year €or a particular plant and that could result in a reactor shutdown. The reactor is expected to be capable of a return to power without special fuel inspection or repair procedures being required. By itself, these events cannot generate a more serious incident-i.e., Condition III or IV category-without other incidents occurring independently. The fuel assembly should be designed in anticipation of the frequency of occurrences for expected Condition II events during the full residence time. Margins to fuel design limits (stress, strain, temperature, creep, etc.) should include provisions for the expected Condition II events.



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American National Standard ANSI/ANSS7.6-1996

PWR Events

For purposes of illustration, the following events have been grouped into this category for pressurized water reactors:

Uncontrolled rod cluster or blade control assembly bank withdrawal from a subcritical ,condition Uncontrolled rod cluster or blade control assembly bank withdrawal at power Rod cluster or blade control assembly misalignment Uncontrolled boron dilution Partial loss of forced reactor coolant flow Startup of an inactive reactor coolant loop Loss of external load, or turbine trip, or both Loss of normal feedwater Loss of offsite power to the station auxiliaries (station blackout) Excessive heat removal due to feedwater system malfunctions Excessive load increase Accidental depressurization of the Reactor Coolant System Accidental depressurization of the Main Steam System Design load rejection transient Operating basis earthquake.

BWR Events

For purposes of illustration, the following events have been grouped into this category for boiling water reactors:

Turbine trip or load rejection Isolation of any or all main steamlines Loss of condenser cooling Loss of feedwater heating Inadvertent moderator cooldown Loss of feedwater flow Total loss of offsite ac power Inadvertent pump start in a hot recirculation loop Inadvertent opening of a relief valve or safety valve. Single failure of a control component or an active component such as:

Turbine pressure regulator failure Feedwater controller failure Recirculation flow control failure Single failure in the electrical system.

Operating basis earthquake.

Condition III - Infrequent Events

Condition III events are those that could occur during the lifetime of a particular plant. They will be accommodated with the failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude resumption of the operation for a considerable outage time. The release of radioactivity will not be sufficient to interrupt or restrict public use of those areas beyond the exclusion area. A Condition III event will not, by itself, generate a Condition IV fault or result in a consequential loss of function of the reactor coolant system or containment barriers.

PWR Events

For purposes of illustration, the following events have been grouped into this category for pressurized water reactors:



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Loss of reactor coolant, from small ruptured pipes or from cracks in large pipes, which actuates

Minor secondary system pipe break Inadvertent loading of fuel assembly into an improper position Complete loss of forced reactor coolant flow Fuel handling accident (minor).

emergency core cooling

BWR Events

For purposes of illustration, the following events have been grouped into this category for boiling water reactors:

Blowdown of reactor coolant through multiple safety or relief valves Loss of reactor coolant from a break or crack which does not depressurize the reactor system, but

which requires the safety-related functions of isolation of containment, emergency core cooling, and reactor shutdown

Improper assembly of core during refueling Seizure of one recirculation pump Startup of an idle recirculation pump in a cold loop Reactor overpressure with delayed scram Turbine trip without bypass.

Condition IV - Limiting Faults

Condition IV events are not expected to occur but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material. They are the most drastic that must be designed against and thus represent limiting design cases. Condition IV events are not to cause a fission product release t o the environment resulting in an undue risk to public health and safety in excess of guideline values of Title 10, "Energy," Code of Federal Regulations, Part 100, "Reactor Site Criteria." A single Condition IV event shall not cause a consequential loss of required functions of systems needed to cope with the fault including those of the emergency core cooling system and the containment.

PWR Events

For purposes of illustration, the following faults have been grouped into this category for pressurized water reactors:

Major rupture of a pipe containing reactor coolant up to and including double-ended rupture of the

Major secondary system pipe rupture up t o and including double-ended rupture (rupture of a

Steam generator tube rupture Single reactor coolant pump locked rotor, or failed coolant pump shaft Fuel handling accident resulting in major clad damage of an irradiated fuel assembly Rupture of a rod drive mechanism housing (rod cluster assembly ejection) Safe shutdown earthquake.

largest pipe in the reactor coolant system (loss of coolant accident)

steam pipe)

BWR Events

For purposes of illustration, the following faults have been grouped into this category for boiling water reactors:

Control rod drop accident Fuel handling accident resulting in major cladding damage of an irradiated fuel assembly



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Major rupture of that portion of a steam line that is not a part of the reactor coolant pressure

Major rupture of pipe in the reactor coolant pressure boundary up to and including a double-ended

Safe shutdown earthquake.

boundary up to and including a double-ended rupture of the steam line

rupture of the largest pipe



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Appendix B (This Appendix is not a part of American National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluation, ANSIhNS57.5-1996, but is included for information purposes only.)

Illustration of the Use of the Standard

Implementation of the standard may be accomplished in four steps:

1. Prepare a matrix of functional requirements versus the Design condition events. (See 5.1,5.2, and Appendix A. See Table 1 as an example of an appropriate format.)

2. .For each block in the matrix, identify the following: a. Whether the functional requirement for this event must be met by regulation, established

practice, or design decision b. The design methodology that has been established to deal with this event (see 5.3.2.4, 5.3.3.4,

5.3.4.4, and 5.5) c. Design input parameters and other information needed to solve the problem characterized by

the methodology (see 5.3.1, 5.3.2.1, 5.3.2.2, 5.3.2.3, 5.3.3.1, 5.3.3.2, 5.3.3.3, 5.3.4.1, and 5.3.4.3) d. The limiting values of the output parameters (limits) from methodology calculations (see 5.4.1) e. Compare the limits and output parameters (margin) (see 5.4.2).

3. Complete calculations and document in a convenient manner (ie., in a manner consistent with company policy and procedure).

4. Prepare a summary report of the design (see Section 6, Documentation Requirements).



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American National Standard ANSI/ANS-57.5-1996

Table 1 - Matrix

Functional

I. Condition I - Normal Operation and Operational Transients

II. Condition II - Events of Moderate Frequencyb

m. Condition III - Infrequent Eventso

IV. Condition IV - Limiting Faultsd

t.: l m l a ;

a Steady state and shutdown operations, operational maneuvers, operation with permissible deviations, preoperational and operational testing, etc.

For PWRs: Uncontrolled rod cluster or blade control assembly bank withdrawal from a subcritical condition; uncontrolled rod cluster or blade control assembly bank withdrawal at power; rod cluster or blade control assembly misalignment; uncontrolled boron dilution; partial loss of forced reactor coolant flow; startup of an inactive reador coolant loop; loss of external load, or turbine trip, o r both; loss of normal feedwater; loss of offsite power to the station auxiliaries (station blackout); excessive heat removal due to feedwater system malfunctions; excessive load increase; accidental depressurization of the Reactor Coolant System; accidental depressurization of the Main Steam System; design load rejection transient; operating basis earthquake. For BWRs: Turbine trip or load rejection; isolation of any or all main steamlines; loss of condenser cooling; loss of feedwater heating, inadvertent moderator cooldown; loss of feedwater flow; total loss of offsite ac power; inadvertent pump start in a hot recirculation loop; inadvertent opening of a relief valve or safety valve; single failure of a control component or an active component (such as turbine pressure regulator failure, feedwater controller failure, recirculation flow control failure, or single failure in the electrical system); operating basis earthquake.

e For PWRs: loss of reactor coolant, from small ruptured pipes o r from cracks in large pipes, which actuates emergency core cooling, minor secondary system pipe break; inadvertent loading of fuel assembly into an improper position; complete loss of forced reactor coolant flow; fuel handling accident (minor). For BWRs: blowdown of reactor coolant through multiple safety or relief valves; loss of reactor coolant from a break or crack which does not depressurize the reactor system, but which requires the safety-related functions of isolation of containment, emergency core cooling, and reactor shutdown; improper assembly of core during refueling; seizure of one recirculation pump; startup of an idle recirculation pump in a cold loop; reactor overpressure with delayed scram; turbine trip without bypass.

, -

For PWRs: Major rupture of a pipe containing reactor coolant up to and including double-ended rupture of the largest pipe in the reactor coolant system (loss of coolant accident); major secondary system pipe rupture up to and including double-ended rupture (rupture of a steam pipe); steam generator tube rupture; single reactor coolant pump locked rotor, or failed coolant pump shaft; fuel handling accident resulting in major clad damage of an irradiated fuel assembly; rupture of a rod drive mechanism housing (rod cluster assembly ejection); safe shutdown earthquake. For BWRs: Control rod drop accident; fuel, handling accident resulting in major cladding damage of a n irradiated fuel assembly; major rupture of that portion of a steam line that is not a part of the reactor coolant pressure boundary up to and including a double-ended rupture of the steam line; major rupture of pipe in the reactor coolant pressure boundary up to and including a double-ended rupture o f the largest pipe; safe shutdown earthquake.



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ANS Light Water Reactors Fuel Assembly Mechanical Design & Evaluation

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