ML20210U082

Requests Review of Eg&G Finding That GE Allowable Values Re Matl Strength Limits Nonconservative & Unacceptable

ML20210U082
Person / Time
Issue date:	03/21/1983
From:	Berlinger C Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
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ML20209B909	List: ML20209B904 ML20210T713 ML20210T743 ML20210T758 ML20210T767 ML20210T782 ML20210T866 ML20210T883 ML20210T925 ML20210T973 ML20210T989 ML20210U006 ML20210U066 ML20210U082
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Text

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, /p* ** cqh, UfJITED STATES

'" ?, NUCLEAR REGULATORY COMMISSION

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• h' WASHINGTON, D. C. 20555 s ~4v,,]c s,, -,

p MAR ^ ' 1583 MEi10RANDUM FOR:

Equipment Qualifications Branch, DE FROM: Carl H. Berlinger, Chief r 75 /-

Core Perfomance Branch, DSI

SUBJECT:

ADEQUACY OF GE MATERIAL STRENGTH LIMITS In a recent review of a GE report by our contractor, Richard Macek (of EG&G Idaho) concluded that GE's allowable values were nonconservative and unacceptable. As well as we can understand the situation, we agree with the contractor. However, we are not well qualified in this area and GE has challenged this finding.

In accordance with our discussion on March 16, I request that you review this finding including the GE strength limits and the SRP acceptance criteria to detemine what is appropriate. Details are in the Enclosure.

a Carl H. Berlinger, Chief Core Perfomance Branch, DSI

Enclosure:

As stated cc: L. S. Rubenstein

Contact:

R. O. Meyer, CPB:DSI X-29475 2 Y' //f[

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MAR .' ! 1983 ENCLOSURE ADEQUACY OF GE MATERIAL STRENGTH LIMITS

References:

"BWR Fuel Assembly Evaluation of Canbined SSE and LOCA Loadings," General Electric Report NEDE-21175-3-P, July 1982.

R. L. Grubb, " Review of LWR Fuel System Mechanical Response liith Recommendations for Component Acceptance Criteria," EG&G Idaho Report NUREG/CR-1018, September 1979.

" Evaluation of Fuel Assembly Structural Response to Externally Applied Forces,"

Appendix A to SRP-4.2, NURGE-0800, July 1981.

A general statement of GE's approach is found on p II.2-1 of the GE report and is given below.

Detailed evaluations are performed on each major fuel assembly component, including the upper tie plate, fuel rods, water rods, spacers, channel and lower tie plate. These evaluations were performed considering the fully irradiated material condition for the components, since this material condition has the lowest ductility and, hence, a higher propensity for material fracture when exposed to large strain anplitudes. The material properties used to demon-strate that fracture 'oll not occur (and, hence, functional adequacy will be maintained) are the ultimate strength of the material, the fatigue strength and the strain capability of the material. The material properties are described in conjunction with the individual component part evaluations.

The full value (i.e.,100%) of the irradiated ultimate tensile strength (UTS) is used on subsequent pages of this GE report as the allowable limit. GE says that this is justifiable on the basis of ANSI /ANS-57.5-19fk.

A general statement of NRC's acceptance criteria is found on p 4.2-20 of the SRP and is given below.

Strengths of fuel assembly components other than spacer grids may be deduced-fron fundamental material properties or experimentation. Supporting evidence for strength values should be supplied. Since structural failure of these - '

components (e.g., fracturing of guide tubes or fragmentation of fuel rods) could be more serious than grid deformation, al'lowable values should bound a O

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large percentage (about 95%) of the distribution of component strengths. There-fore, ASME Boiler and Pressure Vessel Code values and procedures may be used where appropriate for determining yield and ultimate strengths. Specification of allowable values may follow the ASME Code requirements and should include ,

consideration of buckling and fatigue effects.

This paragraph describes the degree of conservatism that we are locking for and implies that the ASME Boiler and Pressure Vessel Code procedures will produce those values.

The conservate margin that we sought when we wrote the SRP would (a) bound the variation due to irradiation effects (hence the use of unirradiated strengths),

(b) bound the normal materials variation (hence the 95% bound), and (c) include a modest safety margin (hence the 70% value). It appears to us that GE's limit values are significantly non-conservative because they were for irradiated rather than unirradiated material and because they use 100% rather than 70% of the UTS; we do not know whether they are using a value that bounds 95% of the normal variation in UTS.

Please advise us in writing of the appropriateness of (a) GE's limit values and (b) our SRP acceptance criteria. We will use this advice as a basis for a decision on the GE review and for modifying the SRP if indicated. It should be noted that the fact that GE's calculated stresses come very close to these questioned limits does not mean that GE fuel has no margin lef t. Those calculates stresses come from a bounding analysis that is essentially back calculated from the limits. GE hopes that we will approve that bounding analysis so they can avoid doing plant specific analyses to calculate more realistic stresses.

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l ARTICLE III-3000

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BASIS FOR ESTABLISHING ALLOWABLE

{ ,, STRESS VALUES 11I-3I00 HASIS FOR ESTABLISIIING (1) one-fourth of the specified minimum tensile STRESS VALUES FOR strength at room temperature; SIATERIALS FOR CLASS 2 (2) one-fourth of the tensile strength at tempera-ture; AND CLASS 3 CONIPONENTS (3) two-thirds of the specified minimum yield III 3110 GENERAL BASIS strength at room temperature; In the determination of allowable stress values for (4) 90% of the yield strength at temperature but pressure parts the Committee is guided by successful ". t to exceed two-thirds of the specified minimum experience in service, insofar as evidence of satisfac. yield strength at room temperature.

tory performance is available. Such evidence is consid-(c) For bolting materials, the basis for setting cred equivalent to test data where operating conditions stresses is the same as for all other materials with the

, are known with reasonable certainty. In the evaluation added requirement that the stresses for heat treated of new materials, it is necessary to be guided to a materials at temperatures below the creep range will i[' ,

certain extent by the comparison of test information n t exceed the lower of 20% of the room temperature 5Pecified minimum tensile strength or 25% of the with similar data on successful applications of similan materials- room temperature specified minimum yield strength.

(d) Stress values for high temperatures are based, whenever possible, on representative properties of the materials under laboratory test conditions. The stress III 3200 FERROUS SIATERIALS values are based on basic properties of the materials

, and no consideration is given to corrosive environ-

.! III 3210 PROCEDURE FOR ment, to abnormal temperature and stress conditions, FSTABLISIIING STRESS or to other design considerations.

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j The mechanical properties considered and factors applied to provide the maximum allowab!: stresses are given in (a) through (d) below.

(a) At any temperature listed below the creep III 3300 NONFERROUS SIATERIALS range, the maximum allowable stress value for ferntic III 3310 51ECllANICAL PROPERTIES steels is the lowest of the following:

. CONSIDERED

, (1) one fourth of the specified minimum tensile

, strength at room temperature; The mechanical properties considered and factors

'; (2) one-fourth of the tensile strength at tempera- applied to provide the maximum allowable stresses are

-l' ture; given in (a) and (b) below.

(3) two thirds of the specified minimum yield (a) At any temperature listed below the creep strength at room temperature; range, the maximum allowable stress value for nonfer-

' (4) two thirds of the yield strength at tempera- rous materials, except for bolting, is the lowest of the ture. following: ,

(b) The maximum a!!owable stress valoe for austen- (/) one fourth of the specified minimum tensile e itic steels is the lowest cf the following/ -

strength at room temperature; 219 ,

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III 3310

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SECTION !!! DIVISION I - APPENDICES l (2) one-fourth of the tensile strength at tempera-2 ture; ,

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' (4) two-thirds of the yield strength at tempera-ture; 6

(J) 90% of the yield strength at temperature but i

not to exceed two-thirds of the yield strength at room temperature for some alloys.

(b) For nonferrous bolting materials, the basis for setting stresses is the same as for all other nonferrous materials with the added requirement that the stresses for materials whose properties at temperatures below the creep range are enhanced by heat treatment or cold working shall not exceed the lower of 20% of the '

room temperature specified minimum tensile strength or 25% of the room temperature specified minimum yield strength.

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APPENDIX IV APPROVAL OF NEW MATERIALS UNDER

'a THE ASNIE BOILER AND PRESSURE VESSEL CODE FOR SECTION III APPLICATION f

i IV.1000 Procedure for Obtaining Approval IV-1100 Code Policy.. . . . . ......... . . . . . . . . . . . . . .... . .. ... . 223 IV.1110 Materials Which Will Be Considered ... . . . . . . . . . ... . . . . . . . . . . . . . . 223

• 1 IV 1200 Data Required to Be Submitted with Requests for Approvals.. . .. . . . . . . 223 IV.1210 Mechanical Properties . .... . ... . . . . . . . . . . . .. ..... ... . . .. . . . . 223 IV 1220 Weldability . .. . . . . . . . . . . .. . .. .. .... . .. .. . ... .. . ... .. 223 IV-1230 Physical Changes... . ........ .... .. . ..... .. ........ . . . . . . . . . . . . . . 223 IV 1300 Proprietary Materials .. . . .. .. ... . . .. ....... .. . . . . . 224

, (.I IV-1310 Patents and Licenses.. . . .................. ... . . . . . . . . . . . . . . . . . . . . . 224 IV 1400 Materials Not Yet Adopted by ASTM. . . .. .. . . . . . . . .. .. .. .. 224 IV.1410 Code Case.. . . .. . . . . . . .. . . . . . . .. 224 6

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4 ARTICLE IV-1000 PROCEDURE FOR OBTAINING APPROVAL 3

IV.1100 CODE POLICY the nil ductility transition temperature shall be sub-(, IV 1110 SIATERIALS WillCil WILL BE . ** e M CAPenence n e emPera.

ture range contemplated will be useful to the Code CONSIDERED ,

Committee.

(a) It is the policy of the Boiler and Pressure Vessel (b) If the material is to be used in components to Committee to adopt for inclusion in Section II only operate under external pressure, the stress-strain such specifications as have been adopted by the curves, either tension or compression, shall be fur.

American Society for Testing and Materials (ASTM). nished for the range of Design Temperature desired.

, (b) It is expected that requests for Code approval will normally be for materials for which there is an ASTM specification. For other materials, request IV.1220 WELDABILITY should be made to ASTM to develop a specification which can be presented to the Code Committee. The inquirer shall furnish complete data on the weldability of material intended for welding, including

,- data on procedure and performance qualification tests I made in accordance with the requirements of Section IV 1200 IX. Welding tests shall be made over the full range of DATA REQUIRED TO BE c ness in w c t matenal is t use . Pertsent SUBNII'ITED WITH REQUESTS ""*""'

FOR APPROVALS ** ****'"*** *"*9" * "*P bility to air hardemng, and the amount of expenence IV.1210 SIECilANICAL PROPERTIES in welding the material, shall be given.

(a) Together with the specification for the material, the inquirer shall furnish the Committee with ade.

quate data on which to base allowable stress values for IV 1230 PflYSICAL CIIANGES inclusion in the applicable stress table. The data shall It is important to know the structural stability include values of ultimate strength, yield strength, characteristics and the degree of retention of proper-reduction of area, elongation, strain fatigue, creep ties with exposure at temperature or neutron irradia-strength, and stress rupture strength of base metal and tion of new materials. The influence of fabrication welded joints over the range of temperatures at which practices, such as forming, welding, and thermal the material is to be used. Any heat treatment that is treatments, on the mechanical properties, ductility, required to produce the tensile properties shall be fully and microstructure of the material are important, described. Adequate data on the notch toughness in particularly where a degradation in properties may be the proposed senice temperatures shall be fumished. encountered. Where particular temperature ranges of The brittle fracture characteristics of the material and exposure or heat treatment, cooling rates, combina.

particularly of steels which will be subjected to tions of mechanical working and thermal treatments,

, neutron irradiation in service, are especially impor- and fabrication practices cause significant changes tant. This information is required for the base metal, (such as the mechanical properties, microstructure,

the weld metal, and the weld heat affected zone in the and resistance to brittle fracture), it is of prime heat treated condition in which they will be in senice. importance to call attention to those conditions uhich if in the intended service the material will be exposed shall be avoided in service or in the manufacture and to neutron irradiation, the effects of sudh exposure on fabrication of parts or components from the material.

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.l 1 % 1300-l % 1410 SECTION III. DIVISION I - APPENDICES IV 1300 PROPRIETARY MATERIALS (a) De inquirer provides evidence that a spec-

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IV 1310 PATENTS AND LICENSES (b) The material is commercially ava.lable i and can k*

ne inquirer shall state whether or not the material be purchased within the specified range of chemical is covered by patents and whether or not it is licensed and tensile requirements and other requirements and, if licensed, the limitations on its manufacture. described in IV 1210.

(c) The inquirer shows that there will be a reason.

able demand for the material by industry and that there exists an urgency for approval by means of a Code Case.

,,IV 1400 MATERIALS NOT YET (d) ne request for approval of the material shall ADOPTED BY ASTM clearly describe it in ASTM specification form, includ-ing such items as scope, process, manufacture, condi-IV 1410 CODE CASE tions for delivery, heat treatment, chemical and tensile in exceptional circumstances, the Code Committee requirements, bending properties, testing specifications l will consider the issuance of a Code Case, etiective for and requirements, workmanship, finish, marking, in-a period of three years, permitting the use of a spection, and rejection.

material providing that the conditions of(a) through (c) ne inquirer shall furnish the Code Committee (e) below are met. wi.h all the data specified in IV 1200 and IV 1300.

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i NCA.2000 - CLASSIFICATION OF COMPONENTS NCA 2142-NCA.2143

, Pressure and Design Mechanical Loads result in which the Level A Service Limits on primary stress stre ses of greater magnitude, relative to the allowable are applicable.

strer,s or stress intensity at the Design Temperature, than would the Service Loadings relative to the NCA 2142.2 Design and Service Limits allowables for the appropriate Service Level. When (a) Design Limiti The limits for Design Loadings this is not the case, and for piping and its supports, are designated as Design Limits.

Service Loadings shall be identified in the Design (b) Service Limits. The Design Specification may Specification. For the Class MC containment vessel, designate Service Limits as defined in (1) through (4) loadings associated with the containment fune. ion below, shall be identified as Design Loadings, excep as (1) LevellService Limits Level A Service Limits

'a provided in NE-3000. are those sets oflimits which must be satisfied for all (b) ne selection of limits for Service Loadings to loadings identified in the Design Specifications to assure operability is beyond the scope of this Section. which the component or support may be subjected in The rules of this Section do not assure operability of the performance ofits specified senice function.

(~ components in which mechanical motion is required. (1) LevelB Service Limitz Level B Service Limits When assurance of operability is required, it is the are those sets of limits which must be satisried for all responsibility of the Owner to define the appropriate loadings identified in the Design Specifications for limiting parameters by referring to documents which which these Service Limits are designated. The com-specify the requirements for operability. Such parame- ponent or support must withstand these loadings ters are outside the scope of this Section [NCA 1130, w:thout damage requiring repair.

NCA 2160, and NCA 5210(b)]. (3) Level C Scryice Limits Level C Service Limits are those sets of limits which must be satisfied for all loadings identified in the Design Specifications for which these Service Limits are designated. These sets

- NCA 2142.1 Design Loadings. Design Loadings for UIII*II* P"II large deformations in areas of strue.

Class I components and suppor's t shall be as defined in .

tural discontmuity. The occurrence of stress to Level

- NB-3112 and NF 3112. Design Loadings for Classes 2 C Limits may necessitate the removal of the compo-and 3 components and Classes 2,3, and MC supports nent from service for inspection er repair of damage to are those pressures, temperatures, and mechanical the component or support. Therefore the selection of loads selected as the basis for the design of the items, this limit shall be reviewed by the Owner for compati-in accordance with (a), (b), and (c) below and the ,

bility with established system safety entena (NCA-additional requirements of the applicable Subsections of this Section (NC-3tl2, ND-3112, NF 3112). De- 2141).

(4) Leiel D Service Limitz Level D Service L,mits i .

sign Loadings for Class'MC vessels are defined in NE.

are those sets oflimits which must be satisfied for all 3112. Design loadings for Class CS core support ,

I adings ident:6ed m the Des,gn i Specifications for structures are defined in NG 3112. which these Senice Limits are designated.m (a) Design Pressure De specified internal and external Design Pressure shall not be less than the maximum difference in pressure between the inside and outside of the item, or between any two chambers -

eref re the selvtion of of a combination unit, which exists under the most this limit shall be resiewed by the Owner far compati-severe loadings for which the Level A Service Limits bility with established system safety criteria (NCA-are applicable. The Design Pressure shall include 2141).

a!!owances for pressure surges.

(b) Design Temperature. Except as otheruise I') #""* " ' '#" U* "* '.*

• P* "'" * ' '" P' p rts may be alternatively designed using more defined in NB.3tl2 for Class I components, the restrictive Senice Limits than specified in the Design specified Design Temperature shall not be less than Specific ti n.1; r example, Level D Service Limits the expected maximum mean metal temperature may be used where Level C Senice Limits have been through the thickness of the part considered for which

• specified.

Level A Limits are specified.

(c) Design .ttechanical Loads The specified Design -

Mechanical Loads shall be selected so that when NC%2143 Consideration of Design Loadings combmed with the effects of Design Preuure, they represent the most severe coincident / oadings for L Components and supports shall comply with all 11

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TeMe F*I322.21 SECTION III, DIVISION I - APPENDICES 1

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) LIMITS 0F PRIMARY LOAD OR STRESS FOR SERVICE LOADINGS WITH LEVEL D SERVICE LIMITS 4

Method of Analysis Design Limets

} Load or 4 System Component Stress Components Compo.1 ent Supports i

F 1322.1 F.1322.2 tjecte (61) (Notes 01/6)) [ Note O)l 1

Elastac flastic Stress 2.4 !. **

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} NO-3221, N8 3230 0.73 l for materials Table 11-2 { gg,,, gg3g 1.23, Nt not >0.73, l F 1323.1 0.73, for materials Table 11.1 f

[ Note (1)l Afternative Limits:

f Valves (F 1350), in preparation Pipeng (F.1360), pressure s2 x Design Pressure 3.0$.

(Es.(9),NB.3652)

Coitapse load Lead P 0.9P, based on 3, = 2.33. or on P, 1.53 N B.3213.22 F.1323.2 [ Note (71) derived from F.1321.1(d) or 1.21,3 gbut

• not >0.73 F 1321.3(a) . [ Note (1))

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F.1324.1 $, + ($, - $,)/ 3 i I"* IIII $, + (3. - 3,)/ 3 J

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Collapse load Load P 0.97, based on $, = 2.3$. or on l F.1324.2 P derived from F.1321.1(d) 1.$3.[

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,g For loads P, [ Note (4)] Same as components f F.1324.3 6

) Plastic LoadP 0.7P, or loads psp., where 1

imtabihty F 132e.4 P. = $, + ($, - 3,)/ 3 Same as components F.1321.1(e) (Note (S)]

$ train limit Load P 0.7P, or loads P sP., where j

foad F 1321.1(f) P. = 3, + ($, - 3,3 / 3, but $ame as components

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F?1324 6 ($.-$,)/3 (Note (11] Same as componeau

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• Use greater of litrits specified.

. " Use lesser of limits .pecified.

t ft0TES:

• 1 (1) 3, value at tempersture shall be specified and justified in Design Report.

(2) P, denotes the coltaose load based on lower bound theorem of limit analyses or as defined in F-1321.1(d).

(3) The Design Limits selected from this Table shall be used in conjunction with F.1323 and F.1324, as applicable, in order to determine the

, limits for P., P,, arkt P,.

l' (4) Higher limi's for $,, may be used as specified in A.9000, where the type of stress field is taken into account.

, ($) 3, is the true effective stress associated with plastic instability IF.1324 4).

(6) For compressive loads sr stresses, the stability requirements of F.1325 shall te rnet.

(7) This enethod is not rermFted if deformation limits are stated in Design Specifications.

' (8) P sdenotes the loti associated with the strain limit placed on the cornponent [F.1321.1(f)l.

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STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION eee.e 4.2 FUEL SYSTEM DESIGN ,

REVIEW RESPONSIBILITIES Primary - Core Performance Branch (CPB) e Secondary - None .

I. AREAS OF REVIEW The thermal, mechanical, and materials design of the fuel system is evaluated by CPB. The fuel system consists of arrays (assemblies or bundles) of fuel rods including fuel pellets, insulator pellets, springs, tubular cladding, end closures, hydrogen getters, and fill gas; burnable poison rods including com-ponents similar to those in fuel rods; spacer grids and springs; end plates; channel boxes; and reactivity control rods. In the case of the control rods, this section covers the reactivity control elements that extend from the coupling interface of the control rod drive mechanism into the core. The Mechanical Engineering Branch reviews the design of control rod drive mechanisms in SRP Section 3.9.4 and the design of reactor internals in SRP Section 3.9.5.

The objectives of the fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and antic-fpated operational occurrences, (b) ft.ol system damage is never so severe as to prevent control rod insertion where it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained. "Not damaged," as used in the above statement, means that fuel rods do not fail, that fuel system dimensions remain within operational tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis. This objective implements General Design Criterion 10 (Ref. 1), and the design limits that accomplish this are called Specified Acceptable Fuel Design Limits (SAFDLs). " Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (Ref. 2) for postulated accidents.

"Coolability," in general, means that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat even after a severe accident. The general requirements to maintain Control rod l Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN Standard review plans are prepared for the guidance of the office of Nuclear Reactor Regulation staff responsibie for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission e podcy to enform the nuclear industry and the general public of regulatory procedures and poiscies Standard review plane are not substitutes for regulatory guides or the Commission's regulatione and compl6ance with them se not required the standard review plan sections are heyed to the standard Format and Content of Safety Analysis Reporte for Nuclear Power Plante.

Not all sections of the Standard Formet have a corresconding review plan.

Published standard review plans well be revised periodically, se appropriate. to accommodate comments and to reflect new snforme.

tion and emperience commente and suggestions for improvement will be considered and should be sent to the u S. Nuclear Regulatory Commission.

Office ,of Nuclear Meactor Regulation. Wachmgton, o C. 20%6 .

N P/lonaW9

96' insertability and core coolability appear repeatedly in the General Design -

Criteria (e.g., GDC 27 and 35). Specific coolability requirements for the loss-of-ccolant accident are civen in 10 CFR Part 50, $50.46 (Ref. 3).

]

All fuel damage criteria are described in SRP Section 4.2. For those criteria that involve ONBR or CPR limits, specific thermal-hydraulic criteria are given in SRP Section 4.4. The available radioactive fission product inventory in fuel rods (i.e., the gap inventory expressed as a release fraction) is provided l . va t0AB 'h for use in estimating the radiological l

( (. y , %R'

. T.e fuel system review covers the following specific areas.

A. Des kes ,'

g O * \/

• l Design bases for the safety analysis address fuel system damage mechanisms and provide limiting values for important parameters such that damage will be limited to acceptable levels. The design bases should reflect the safety review objectives as described above.

8. Description and Desian Drawings The fuel system description and design drawings are reviewed. In general, l the description will emphasize product specifications rather than process l

specifications.

l l

C. Desian Evtluation The performsnce of the fuel system during normal operation, anticipated 'S l operational occurrences, and postulated accidents is reviewed to determine if all design bases are met. The fuel system components, as listed above, are reviewed not only as separate components but also as integral units such as fuel rods and fuel assemblies. The review consists of an evaluation of operating experience, direct experimental comparisons, detailed mathematical analyses, and other information.

D. Testina. Inspection, and Surveillance Plans Testing and inspection of new fuel is performed by the licensee to ensure that the fuel is fabricated in accordance with the design and that it reaches the plant site and is loaded in the core without damage. On-line fuel rod failure monitoring and postirradiation surveillance should be performed to detect anomalies or confirm that the fuel system is performing as expected; surveillance of control rods containing B C4 should be performed

• to ensure against reactivity loss. The testing, inspection, and surveil-1ance plans along with their reporting provisions are reviewed by CPB to ensure that the important fuel design considerations have been addressed.

II. ACCEPTANCE CRITERIA Specific criteria necessary to meet the requirements of 10 CFR Part 50,650.46; General Design Criteria 10, 27, and 35; Appendix K to 10 CFR Part 50; and

- 10 CFR Part 100 identified in subsection I of this SRP section are as follows:

. q' 4.2-2 Rev. 2 - July 1981

A. Desian Bates The fuel system design bases must reflect the four objectives described in subsection I, Areas of Review. To satisfy these objectives, acceptance criteria are needed for fuel system damage, fuel rod failure, and fuel coolability. These criteria are discussed in the following:

1. Fuel System Damace This subsection applies to normal operation, and the information to be reviewed should be contained in Section 4.2 of the Safety Analysis Report.

To meet the requirements of General Design Criterion 10 as it relates to Specified Acceptable Fuel Design Limits for normal operation, including anticipated operational occurrences, fuel system damage criteria should be given for all known damage mechanisms.

Fuel system damage includes fuel rod failure, which is discussed below in subsection II.A.2. In addition to precluding fuel rod failure, fuel damage criteria should assure that fuel system dimen-sions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analysis. Such damage criteria should address the following to be complete.

(a) Stress, strain, or loading limits for spacer grids, guide tubes, thimbles, fuel rods, control' rods, channel boxes, and other fuel system structural members should be provided.

Stress limits that are obtained by methods similar to those given in Section III of the ASME Code (Ref. 4) are acceptable.

Other proposed limits must be justified.

(b) The cumulative number of strain fatigue cycles on the structural members mentioned in paragraph (a) above should be significantly less than the design fatigue lifetime, which is based on appro-priate data and includes a safety factor of 2 on stress amplitude or a safety factor of 20 on the number of cycles (Ref 5).

Other proposed limits must be justified.

(c) Fretting wear at contact points on the structural members mentioned in paragraph (a) above should be limited. The allowable fretting wear should be stated in the Safety Analysis Report and the stress and fatigue limits in paragraphs (a) and (b) above should presume the existence of this wear.

(d) 0xidation, hydriding, and the buildup of corrosion products (crud) should be Ilmited. Allowable oxidation, hydriding, and crud levels should be discussed in the Safety Analysis Report and shown to be acceptable. These levels should be presumed to exist in paragraphs (a) and (b) above. The effect of crud on thermal-hydraulic considerations is reviewed as described in l SRP Section 4.4 (e) Dimensional changes such as rod bowing or irradiation growth of fuel rods, control rods, and guide tubes need not be limited to 4.2-3 Rev. 2 - July 1981

• set values (i.e., damage limits), but they must be included in the design analysis to establish operational tolerances. m

)

(f) Fuel and burnable poison red internal gas pressures should remain below the nominal system pressure during normal opera-tionunlessotherwisejustified.

(g) Worst-case hydraulic loads for normal operation should not exceed the holddown capability of the fuel assembly (either gravity or holddown springs). Hydraulic loads for this evaluation are reviewed as described in SRP Section 4.4. l (h) Control rod reactivity must be maintained. This may require the control rods to remain watertight if water-soluble or leachable materials (e.g., 8 C) 4 are used.

2. Fuel Rod Failure This subsection applies to normal operation, anticipated operational occurrences and postulated accidents. Paragraphs (a) through (c) addressfailuremechanismsthataremoreIfmitingduringnormal operation, and the information to be reviewed should be contained in Section 4.2 of the Safety Analysis Report. Paragraphs (d) through (h) address failure mechanisms that are more limiting during anticipated operational occurrences and pcstulated accidents, and the information to be reviewed will usually be contained in Chapter 15 of the Safety Analysis Report. Paragraph (i) should be addressed in Section 4.2 of the Safety Analysis Report because it is not addressed elsewhere.

To meet the requirements of (a) General Design Criterion 10 as it relates to Specified Acceptable Fuel Design Limits for normal opera-tion, including anticipated operational occurrences, and (b) 10 CFR Part 100 as it reiates to fission product releases for postulated accidents, fuel rod failure criteria should be given for all known fuel rod failure mechanisms. Fuel rod failure is defined as the loss of fuel rod hermeticity. Although we recognize that it is not possible to avoid all fuel rod failures and that cleanup systems are installed to handle a small number of leaking rods, it is the objective of the review to assure that fuel does not fail due to specific causes during normal operation and anticipated operational occurrences.

Fuel rod failures are permitted during postulated accidents, but they must be accounted for in the dose analysis.

Fuel rod failures can be caused by overheating, pellet / cladding interaction (PCI), hydriding, cladding collapse, bursting, mechanical fracturing, and fretting. Fuel failure criteria should address the following to be complete.

(a) Hydriding: Hydriding as a cause of failure (i.e., primary I hydriding) is prevented by keeping the level of moisture and other hydrogenous impurities very low during fabrication.

Acceptable moisture levels for Zircaloy-clad uranium oxide fuel should be no greater than 20 ppm. Current ASTM specifications (Ref. 7) for UO2 fuel pellets state an equivalent limit of 2 ppm g of hydrogen from all sources. For other materials clad in 4.2-4 Rev. 2 - July 1981

_ m-___ ___m -m___

Zircaloy tubing, an equivalent quantity of moisture or hydrogen can be tolerated. A moisture level of 2 mg H 2O per cm3 of hot void volume within the Zircaloy cladding has been shown (Ref. 8) to be insufficient for primary hydride formation.

(b) Cladding Collapsa: If axial gaps in the fuel pellet column I occur due to densification, the cladding has the potential of collapsing into a gap (i.e., flattening). Because of the large local strains that accompany this process, collapsed (flattened) cladding is assumed to fail.

(c) Fretting: Fretting is a potential cause of fuel failure, but l it is a gradual process that would not be effective during the brief duration of an abnormal operational occurrence or a postulated accident. Therefore, the fretting wear requirement in paragraph (c) of subsection II.A.1, Fuel Damage, sufficient to preclude fuel failures caused by fretting during .nsients.

(d) Overheating of Cladding: It has been traditional practice to assume that failures will not occur if the thermal margin criteria (DNBR for PWRs and CPR for BWRs) are satisfied. The review of these criteria is detailed in SRP Section 4.4. For normal operation and anticipated operational occurrences, violation of the thermal margin criteria is not permitted. For postulated accidents, the total number of fuel rods that exceed the briteria has been assumed to fail for radiological dose calculation purposes.

Although a thermal margin criterion is sufficient to demonstrate the avoidance of overheating from a deficient cooling mechanism, it is not a necessary condition (i.e., DNB is not a failure mechanism) and other mechanistic methods may be acceptable.

There is at present little experience with other approaches, but new positions recommeniing different criteria should address cladding temperature, pressure, time duration, oxidation, and embrittlement.

(e) Overheating of Fuel Pellets: It has also been traditional practice to assume that failure will occur if centerline melting takes place. This analysis should be performed for the maximum linear heat generation rate anywhere in the core, including all hot spots and hot channel factors, and should account for the effects of burnup and composition on the melting point. For normal operation and anticipated operational occurrences, centerline melting is not permitted. For postulated accidents, the total number of rods that experience centerline melting should be assumed to fail for radiological dose calculation purposes. The centerline melting criterion was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to come into contact with the cladding nor produce local hot spots. The assumption that centerline melting results in fuel failure is conservative.

(f) Excessive Fuel Enthalpy: For a severe reactivity initiated accident (RIA) in a BWR at zero or low power, fuel failure is assumed to occur if the radially averaged fuel rod enthalpy is 4.2-5 Rev. 2 - July 1981 L_- -_______- _ _ _ _ _ _ _ _ _ - _ _

~

~

greater than 170 cal /g at any axial location. For full power RIAs in a BWR and all RIAs in a PWR, the thermal margin criteria (DNBR and CPR) are used as fuel failure criteria to meet the guidelines of Regulatory Guide 1.77 (Ref. 6) as it relates to fuel rod failure. The 170 cal /g enthalpy criterion is primarily intended to address cladding overheating effects, but it also indirectly addresses pellet / cladding interactions (PCI). Other criteria may be more appropriate for an RIA, but continued approval of this enthalpy criterion and the thermal margin criteria may be given until generic studies yield improvements.

(g) Pellet / Cladding Interaction: There is no current criterion I for fuel failure resulting from PCI, and the design basis can only be stated generally. Two related criteria should be applied, but they are not sufficient to preclude PCI failures.

(1) The uniform strain of the cladding should not exceed 1%.

In this context, uniform strain (elastic and inelastic) is defined as transient-induced deformation with gage lengths

- corresponding to cladding dimensions; steady-state creepdown and irradiation growth are excluded. Although observing this strain limit may preclude some PCI failures, it will not preclude the corrosion-assisted failures that occur at low strains, nor will it preclude highly localized overstrain failures. (2) Fuel melting should be avoided. The large volume increase associated with melting may cause a pellet with a molten center to exert a stress on the cladding. Such a PCI is avoided by avoiding fuel melting. Note that this same criterion was invoked in para-graph (e) to ensure that overheating of the cladding would not occur.

(h) Bursting: To meet the requirements of Appendix K of 10 CFR Part 50 (Ref. 9) as it relates to the incidence of rupture during a LOCA, a rupture temperature correlation must be used in the LOCA ECCS analysis. Zircaloy cladding will burst (rupture) under certain combinations of temperature, heating rate, and differential pressure. Although fuel suppliers may use different rupture-temperature vs differential pressure curves, an acceptable curve should be similar to the one described in Ref. 10. l (i) Mechanical Fracturing: A mechanical fracture refers to a l defect in a fuel rod caused by an externally applied force such as a hydraulic load or a load derived from core plate motion. I Cladding integrity may be assumed if the applied stress is less than 90% of the irradiated yield stress at the appropriate temperature. Other proposed limits must be justified. Results from the seismic and LOCA analysis (see Appendix A to this SRP section) may show that failures by this mechanism will not l occur for less severe events.

3. Fuel Coolability This subsection applies to postulated accidents, and most of the information to be reviewed will be contained in Chapter 15 of the ,

Safety Analysis Report. Paragraph (e) addresses the combined effects y 4.2-6 Rev. 2 - July 1981

of two accidents, however, and that information should be contained in Section 4.2 of the Safety Analysis Report. To meet the require-ments of General Design Criteria 27 and 35 as they relate to control rod insertability and core coolability for postulated accidents, fuel coolability criteria should be given for all severe damage mechanisms. Coolability., or coolable geometry, has traditionally implied that the fuel assembly retains its red-bundle geometry with adequate coolant channels to permit removal of residual heat.

Reduction of coolability can result from cladding embrittlement, violent expulsion of fuel, generalized cladding melting, gross structural deformation, and extreme coplanar fuel rod ballooning.

Control rod insertability criteria are also addressed in this j subsection. Such criteria should address the following to be complete: 1 (a) Cladding Embrittlement: To meet the requirements of 10 CFR Part 50, $50.46, as it relates to cladding embrittlement for a LOCA, acceptance criteria of 2200 F on peak cladding temperature and 17% on maximum cladding oxidation must be met. (Note: If the cladding were predicted to collapse in a given cycle, it would also be predicted to fail and, therefore, should not be irradiated in that cycle; consequently, the lower peak cladding temperature limit of 1800 F previously described in Reference 11 is no longer needed.) Similar temperature and oxidation criteria may be justified for other accidents.

(b) Violent Expulsion of Fuel: In severe reactivity initiated accidents, such as rod ejection in a PWR or rod drop in a BWR, the large and rapid deposition of energy in the fuel can result in melting, fragmentation, and dispersal of fuel. The mechanical action associated with fuel dispersal can be sufficient to destroy the cladding and the rod-bundle geometry of the fuel and to pro-duce pressure pulses in the primary system. To meet the guide-lines of Regulatory Guide 1.77 as it relates to preventing wide-spread fragmentation and dispersal of the fuel and avoiding tne generation of pressure pulses in the primary system of a PWR, a radially averaged enthalpy limit of 280 cal /g should be observed.

This 280 cal /g limit should also be used for BWRs. i (c) Generalized Cladding Melting: Generalized (i.e., non-local) melting of the cladding could result in the loss of rod-bundle fuel geometry. Criteria for cladding embrittlement in paragraph (a) above are more stringent than melting criteria would be; therefore, additional specific criteria are not used.

(d) Fuel Rod Ballooning: To meet the requirements of Appendix K of 10 CFR Part 50 as it relates to degree of swelling, burst strain and flow blockage resulting from cladding ballooning (swelling) must be taken into account in the analysis of core flow distribution. Burst strain and flow blockage models must be based on applicable data (such as Refs. 10, 12, and 13) in such a way that (1) the temperature and differential pressure at which the cladding will rupture are properly estimated (see paragraph (h) of subsection II.A.2), (2) the resultant degree of cladding swelling is not underestimated, and (3) the asso-ciated reduction in assembly flow area is not underestimated.

4.2-7 Rev. 2 - July 1981

The flow blockage model evaluation is provided to the Reactor Systems Branch for incorporation in the comprehensive ECCS evaluation model to show that the 2200 F cladding temperature )

and 17% cladding oxidation limits are not exceeded. The reviewer should also determine if fuel rod ballooning should be included in the analysis of other accidents involving system depressurization.

(e) Structural Deformation: Analytical procedures are discussed in Appendix A. " Evaluation of Fuel Assembly Structural Response to Externally Applied Forces."

8. Description and Design Drawings The reviewer should see that the fuel system description and design drawings are complete enough to provide an accurate representatioq and to supply information needed in. audit evaluations. Completeness is a matter of judgment, but the following fuel system information and associated toleranc'es are necessary for an acceptable fuel system description:

Type and metallurgical state of the cladding Cladding outside diameter Cladding inside diameter Cladding inside roughness Pellet outside diameter Pellet roughness Pellet density Pellet resintering data -

.ra llet length Pellet dish dimensions -

Burnable poison content J Insulator pellet parameters Fuel column length Overall rod length Rod internal void volume Fill gas type and pressure Sorbed gas composition and content Spring and plug dimensions Fissile enrichment -

Equivalent hydraulic diameter Coolant pressure The following design drawing have also been found necessary for an acceptable fuel system description:

Fuel assembly cross section Fuel assembly outline Fuel rod schematic Spacer grid cross section Guide tube and nozzle joint Control rod assembly cross section Control rod assembly outline Control rod schematic Burnable poison rod assembly cross section Burnable poison rod assembly outline Burnable poison rod schematic Orifice and source assembly outline 4.2-8 Rev. 2 - July 1981

. C. Design Evaluation The methods of demonstrating that the design bases are met must be reviewed. Those methods include operating experience, prototype testing, and analytical predictions. Many of these methods will be presented generically in topical reports and will be incorporated in the Safety Analysis Report by reference.

1. Operating Exoerience Operating experience with fuel systems of the same or similar design should be described. When adherence to specific design criteria can -

be conclusively demonstrated with operating experience, prototype testing and design analyses that were performed prior to gaining that experience need not be reviewed._ Design criteria for fretting wear, oxidation, hydriding, and crud buildup might be addressed in this manner.

2. Prototype Testing When conclusive operating experience is not available, as with the introduction of a design change, prototype testing should be reviewed.

Out-of-reactor tests should be performed when practical to determine the characteristics of the new design. No definitive requirements have been developed regarding those design features that must be tested prior to irradiation, but the following out-of-reactor tests have been performed for this purpose and will serve as a guide to the reviewer:

Spacer grid structural tests Control rod structural and performance tests Fuel assembly structural tests (lateral, axial and torsional stiffness, frequency, and damping)

Fuel assembly hydraulic flow tests (lift forces, control rod wear, vibration, and assembly wear and life)

In-reactor testing of design features and lead-assembly irradiation of whole assemblies of a new design should be reviewed. The following phenomena that have been tested in this manner in new designs will serve as a guide to the reviewer:

Fuel and burnable poison rod growth Fuel rod bowing Fuel assembly growth Fuel assembly bowing Channel box wear and distortion Fuel rod ridging (PCI)

Crud formation . -

Fuel rod integrity Holddown spring relaxation Spacer grid spring relaxation Guide tube wear characteristics In some cases, in-reactor testing of a new fuel assembly design or a new design feature cannot be accomplished prior to operation of a full core of that design. This inability to perform in-reactor .

4.2-9 Rev. 2 - July 1981

testing may result from an incompatability of the new design with the previous design. In such cases, special attention should be given to the surveillance plans (see subsection II.D below).

3. Analytical Predictions Some design bases and related parameters can only be evaluated with calculational procedures. The analytical methods that are used to make performance predictions must be reviewed. .Many such reviews have been performed establishing numerous examples for the reviewer.

The following paragraphs discuss the more established review patterns and provide many related references.

(a) Fuel Temperatures (Stored Energy): Fuel temperatures and stored energy.during normal operation are needed as input to ECCS performance' calculations. The temperature calculations require complex computer codes that model many different phenomena. Phenomenological models that should be reviewed include the following:

Radial power distribution Fuel and cladding temperature districution Burnup distribution in the fuel Thermal conductivity of the fuel, cladding, cladding crud, and oxidation layers

Densification of the fuel i Thermal expansion of the fuel and cladding Fission gas production and release '

Solid and gaseous fission product swelling )

Fbel restructuring and relocation /

Fuel and cladding dimensional changes

, e Fuel-to-cladding heat transfer coefficient Thermal conductivity of the gas mixture TfDrmal conductivity in the Knudsen domain FQel-to-cladding contact pressure H ec capacity of the fuel and cladding G OWth and creep of the cladding Ro'd internal gas pressure and composition Sorption of helium and other fill gases Cladding oxide and crud layer thickness Cladding-to coolant heat transfer coefficient

• Because of the strong interaction'between these models, overall code behavior must be checked against data (standard problems or benchmarks) and the NRC audit codes (Refs. 14 and 15).

Examples of previous fuel performance code reviews are given in References 16 through 20.

(b) Densification Effects

In addition to its effect on fuel temperatures (discussed above), densification affects (1) core Although needed in fuel performance codes, this model is reviewed as described in SRP Section 4.4.

J 4.2-10 Rev. 2 - July 1981

power distributions (power spiking, see SRP Section 4.3),

(2) the fuel linear heat generation rate (LHGR, see SRP Section 4.4), and (3) the potential for cladding collapse.

Densification magnitudes for power spike and LHGR analyses are discussed in Reference 21 and in Regulatory Guide 1.126 (Ref. 22).

To be acceptable, densification models should follow the guide-lines of Regulatory Guide 1.126. Models for cladding-collapse times must also be reviewed, and previous review examples are given in References 23 and 24.

(c) Fuel Rod Bowing: Guidance for the analysis of, fuel rod bowing is given in Reference 25. Interim methods that may be used prior to compliance with this guidance are given in Reference 26.

At this writing, the causes of fuel rod bowing are not well understood and mechanistic analyses of rod bowing are not being approved.

(d) Structural Deformation: Acceptance Criteria are discussed in Appendix A, " Evaluation of Fuel Assembly Structural Response to Externally Applied Forces."

(e) Rupture and Flow Blockage (Ballooning): Zircaloy rupture and-flow blockage models are part of the ECCS evaluation model and should be reviewed by CPB. The models are empirical and should be compared with relevant data. Examples of such data and previous reviews are contained in References 10, 12, and 13. l (f) Fuel Rod Pressure: The thermal performance code for calculating temperatures discussed in paragraph (a) above should be used to calculate fuel rod pressures in conformance with fuel damage criteria of Subsection II.A.1, paragraph (f). The reviewer should ensure that conservaxisms that were incorporated for calculating temperatures do not introduce nonconservatisms with regard to fuel rod pressures.

(g) Metal / Water Reaction Rate: To meet the requirements of Appendix K of 10 CFR Part 50 (Ref. 9) as it relates to metal / water reaction rate, the rate of energy release, hydrogen generation, and cladding oxidation from the metal / water reaction should be calculated using the Baker-Just equation (Ref. 27). For non-LOCA l applications, other correlations may be used if ju3tified.

(h) Fission Product Inventory: To meet the guidelines of Regulatory Guides 1.3, 1.4, 1.25 and 1.77 (Refs. 6, 28-30) as they relate to fission product release, the available radioactive fission product inventory in fuel rods (i.e. , the gap inventory) is presently specified by the assumptions in those Regulatory Guides. These assumptions should be used until improved calculational methods are approved by CPB (see Ref. 31). -

D. Testing, Inspection, and Surveillance Plans Plans must be reviewed for each plant for testing and inspection of new fuel and for monitoring and surveillance of irradiated fuel.

4.2-11 Rev. 2 - July 1981

. j 1

1. Testing and Inspection of New Fuel  !

\

Testing and inspection plans for new fuel should include verification  !

of cladding integrity, fuel system dimensions, fuel enrichment, l burnable poison concentration, and absorber composition. Details of )

the manufacturer's testing and inspection programs should be documented  ;

in quality control reports, which should be referenced and summarized  !

in the Safety Analysis Report. The program for onsite inspection of new fuel and control assemblies after they have been delivered to the plant should also be described. Where the overall testing and inspection programs are essentially the same as for previously approved plants, a statement to that effect should be made. In that case, the details of the programs need not be included in the Safety Analysis Report, but an appropriate reference should be cited and a (tabular) summary should be presented.

2. On-line Fuel System Monitoring The applicant's on-line fuel rod failure detection methods should be.

reviewed. Both the sensitivity of the instruments and the applicant's commitment to use the instruments should be evaluated. References 32 and 33 evaluate several common detection methods and should be utilized in this review.

Surveillance is also needed to assure that B4 C control rods are not losing reactivity. Boron compounds are susceptible to leaching in the event of a cladding defect. Periodic reactivity wor;h tests such as described in Reference 34 are acceptable. ~

)

3. Post-irradiation Surveillance . /'

A post-irradiation fuel surveillance program should be described for each plant to detect anomalies or confirm expected fuel performance.

The extent of an acceptable program will depend on the hisXory of the fuel design being considered, i.e., whether the propose.1 fuel design is the same as current operating fuel or incorporates new design features. -

For a fuel design like that in other operating plants, a minimum acceptable program should include a qualitative visual examination of some discharged fuel assemblies from each refueling. Such a program should be sufficient to identify gross problems of structural integrity, fuel rod failure, red bowing, or crud deposition. There should also be a commitment in the program to perform additional surveillance if unusual behavior is noticed in the visual examination' or if plant instrumentation indicates gross fuel failures. The surveillance program should address the disposition of failed fuel.

In addition to the plant-specific surveillance program, there should exist a continuing fuel surveillance effort for a given type, make, or class of fuel that can be suitably referenced by all plants using similar fuel. In the absence of such a generic program, the reviewer should expect more detail in the plant-specific program.

For a fuel design that introduces new features, a more detailed surveillance program commensurate with the nature of the changes M 4.2-12 Rev. 2 - July 1981

should be described. This program should include appropriate qualitative and quantitative inspections to be carried out at interim and end-of-life refueling outages. This surveillance program should be coordinated with prototype testing discussed in subsection II.C.2.

When prototype testing cannot be performed, a special detailed surveillance program should be planned for the first irradiation of a new design.

III. REVIEW PROCEDURES For construction permit (CP) applications, the review should assure that the design bases set forth in the Preliminary Safety Analysis Report (PSAR) meet the acceptance criteria given in subsection II.A. The CP review should further determine from a study of the preliminary fuel system design that there is reasonable assurance that the final fuel system design will meet the design bases. This judgment inay be based on experience with similar designs.

For operating license (OL) applications, the review should confirm that the design bases set forth in the Final Safety Analysis Report (FSAR) meet the acceptance criteria given in Subsection II.A and that the final fuel system design meets the design bases.

Much of the fuel system review is generic and is not repeated for each similar plant. That is, the reviewer will have reviewed the fuel design or certain aspects of the fuel design in previous PSARs, FSARs, and licensing topical reports. All previous reviews on which the current review is dependent should be referenced so that a completely documented safety evaluation is contained in the plant safety evaluation report. In particular, the NRC safety evaluation reports for all relevant licensing topical reports should be cited.

n Certain generic reviews have also been performed lqr CPB reviewers with findings issued-as NUREG- or WASH-series reports. At the present time these reports include References 9, 11, 21, 31, 32, 35, and 36, and they should all be I appropriately cited in the plant safety evaluatipen report. Applicable Regulatory Guides (Refs. 6, 22, 28-30, and 41) should also be mentioned in the plant l safety evaluation reports. Deviation from these guides or positions should be explained. After briefly discussing related previous reviews, the plant safety evaluation should concentrate on areas where the application is not identical to previously reviewed and approved applications and areas related to newly discovered problems.

Analytical predictions discussed in Subsection II.C.3.will be reviewed in PSARs, FSARs, or licensing topical reports. When the methods are being reviewed, calculations by the staff may be performed to verify the adequacy of the analytical methods. Thereafter, audit calculations will not usually be performed to check the results of an approved method that has been submitted in a Safety Analysis Report. Calculations, benchmarking exercises, and additional reviews of generic methods may be undertaken, however, at any time the clear need arises to reconfirm the adequacy of the method.

IV. EVALUATION FINDINGS The reviewer should verify that sufficient information has been provided to satisfy the requirements of this SRP section and that the evaluation supports conclusions of the following type, to be included in the staff's safety evaluation report:

4.2-13 Rev. 2 - July 1981

. .o .

The staff concludes that the fuel system of the plant has l been designed so that (a) the fuel system will not'be damaged as a result of normal operation and anticipated operational occurrences, (b) fuel damage 3 during postulated accidents would not be severe enough to prevent control rod ')

insertion when it is required, and (c) core coolability will always be main-tained, even after severe postulated accidents and thereby meets the related requirements of 10 CFR Part 50, $50.46; 10 CFR Part 50, Appendix A, General Design Criteria 10, 27 and 35; 10 CFR Part 50, Appendix K; and 10 CFR Part 100.

This conclusion is based on the following:

1. The applicant has provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions. Those analytical predictions dealing with structur*1 response, control rod ejection (PWR) or drop (BWR), and fuel densitication have been performed in accordance with *

(a) the guidelines of Regulatory Guides 1.60, 1.77, and 1.126, or methods that the staff has reviewed and found to be acceptable alternatives to those Regulatory Guides, and (b) the guidelines for

" Evaluation of Fuel Assembly Structural Response to Externally Applied Forces" in Appendix A to SRP Section 4.2.

2. The applicant has provided for testing and inspection of new fuel to ensure that it is within design tolerances at the time of core loading. The applicant has made a commitment to perform on-line fuel failure monitoring and postirradiation surveillance to detect anomalies or confirm that the fuel has performed as expected.

The staff concludes that the applicant has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated and thereby meets the related requirements of ',

10 CFR Part 100. In meeting these requirements, the applicant has (a) used the fission product release assumptions of Regulatory Guides 1.3 (or 1.4),

1.25, and 1.77 and (b) performed the analysis for fuel rod failures for the

rod ejection accident in accordance with the guidelines of Regulatory Guide 1.77 or with methods that the staff has reviewed and found to be an acceptable alternative to Regulatory Guide 1.77.

V. IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.

Except in those cases in which the ap;.licant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.

Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides and NUREGs.

VI. REFERENCES

1. 10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants."

]

4.2-14 Rev. 2 - July 1981

2. 10 CFR Part 100, " Reactor Site Criteria."

3. 10 CFR Part 50, 650.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

4. '! Rules for Construction of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code,Section III, 1977.

5. W. J. O'Donnel and 8. F. Langer, " Fatigue Design Basis for Zircaloy Components," Nucl. Sci. Eng. 20, 1 (1964).

6. Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors."

7. " Standard Specification for Sintered Uranium Dioxide Pellets," ASTM Standard C776-76, Part 45, 1977.

8. K. Joon, " Primary Hydride Failure of Zircaloy-Clad Fuel Rods," Trans. Am.

Nucl. Soc. 15, 186 (1972).

9. 10 CFR Part 50, Appendix K, "ECCS Evaluation Models."

10. D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," USNRC Report NUREG-0630, April 1980.

11. " Technical Report'on Densification of Light Water Reactor Fuels," AEC Regulatory Staff Report WASH-1236, November 14, 1972.

12. F. Erbacher, H. J. Neitzel, H. Rosinger, H. Schmidt, and K. Wiehr, " Burst s

Criterion of Zircaloy Fuel Claddings in a LOCA," ASTM Fifth International Conference on Zirconium in the Nuclear Industry, August 4-7, 1980, Boston, Massachusetts.

13. R. H. Chapman, "Multirod Burst Tcst Program Progress Report for January-June 1980," Oak Ridge National Laboratory Report NUREG/CR-1883, March 1981.

14. C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko and L. J. Parchen,

" User's Guide for GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Red," Battelle Pacific Northwest Laboratory Report BNWL-1897, November 1975.

15. C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko and L. J. Parchen, "GAPCON-THERMAL-2: A Computer Program for Caculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratory Report BNWL-1898, November 1975.

16. R. H. Stoudt, D. T. Buchanan, 8. J. Buescher, L. L. Losh, H. W. Wilson and P. J. Henningson, " TACO - Fuel Pin Performance Analysis, Revision 1,"

Bacock & Wilcox Report BAW-10087A, Rev. 1, August 1977.

17. " Fuel Evaluation Model," Combustion Engineering Report CENPD-139-A, July 1974 (Approved version transmitted to NRC April 25, 1975).

18. " Supplement 1 to the Technical Report on Densification of General Electric Reactor Fuels," AEC Regulatory Staff Report, December 14, 1973.

4.2-15 Rev. 2 - July 1981

19. " Technical Report on Densification of Exxon Nuclear PWR Fuels," AEC Regulatory Staff Report, February 27, 1975.

20. Letter from J. F. Stolz, NRC, to T. M. Anderson, Westinghouse,

Subject:

Safety Evaluation of WCAP-8720, dated February 9, 1979.

21. R. O. Meyer, "The Analysis of Fuel Densification," USNRC Report NUREG-0085, July 1976.

22. Regulatory Guide 1.126, "An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification."

23. Memorandum from V. Stello, NRC, to R. C. DeYoung,

Subject:

Evaluation of Westinghouse Report, WCAP-8377, Revised Clad Flattening Model: dated January 14, 1975.

24. Memorandum from D. F. Ross, NRC, to R. C. DeYoung,

Subject:

CEPAN --

Method of Analyzing Creep Collapse of Oval Cladding, dated February 5, 1976. -

25. Memorandum from D. F. Ross, NRC, to D. B. Vassallo,

Subject:

Request for Revised Rod Bowing Topical Reports, dated May 30, 1978.

26. Memorandum from D. F. Ross and D. G. Eisenhut, NRC, to D. B. Vassallo and K. R. Galler,

Subject:

Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors, dated February 16, 1977.

27. L. Baker and L. C. Just, " Studies of Metal-Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium - /

Water Reaction," Argonne National Laboratory Report ANL-6548, May 1962.

28. Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors."

29. Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of Loss-of-Coolant Accident for Pressurized Water Reactors."

30. Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."

31. "The Role of Fission Gas Release in Reactor Licensing," USNRC Report NUREG-75/077, November 1975.

32. B. L. Siegel and H. H. Hagen, "Fue! Failure Detection in Operating Reactors,"

USNRC Report NUREG-0401, March 1978.

33. W. J. Bailey, et al., " Assessment of Current Onsite Inspection Techniques for LWR Fuel Systems," Battelle Pacific Northwest Laboratory Report NUREG/CR-1380, Vol.1, July 1980, Vol. 2, January 1981.

34. " Safety Evaluation Report Related to Operation of Arkansas Nuclear One, j Unit 2," USNRC Report NUREG-0308, Supp. 2, September 1978. -s 4.2-16 Rev. 2 - July 1981

35. B. L. Siegel, " Evaluation of the Behavior of Waterlogged Fuel Rod Failures

- in LWRs," USNRC Report NUREG-0308, March 1978.

36. R. O. Meyer, C. E. Be Fuel at High Burnup,"yer and NUREG-0418, USNRC Report J. C. Voglewede, " Fission Gas Release from March 1978.

37. R. L. Grubb, " Review of LWR Fuel System Mechanical Response with Recommendations for Component Acceptance Criteria," Idaho National Engineering Laboratory, NUREG/CR-1018, September 1979.

38. R. L. Grubb, " Pressurized Water Reactor Lateral Core Response Routine, FAMREC (Fuel Assembly Mechanical Response Code)," Idaho National Engineering Laboratory, NUREG/CR-1019, September 1979.

39. R. L. Grubb, " Technical Evaluation of PWR Fuel Spacer Grid Response Load Sensitirity Studies," Idaho National Engineering Laboratory, NUREG/CR-1020, -

Septeabar 1979. ,' ,

40. S. B. Hosford, et al., " Asymmetric Blowdown Loads on PWR Primary Systems,"

USNRC Report NUREG-0609, January 1981.

41. Regulatory Guide 1.60 Nuclear Power Plants.', " Design Response Spectra for Seismic Design of 4.2-17 Rev. 2 - July 1981

. /

i  !

U.S. Nuclear Regulatory Commission -

• Office of Nuclear Reactor Regulation j

APPENDIX A EVALUATION OF FUEL ASSEMBLY STRUCTURAL RESPONSE TO EXTERNALLY APPLIED FORCES T0 t

STANDARD REVIEW PLAN SECTION 4.2 A. BACKGROUND Earthquakes and postulated pipe breaks in the reactor coolant system would result in external forces on the fuel assembly. SRP Section 4.2

• s tates that fuel system coolabil'ty should be maintained and that damage should not be so severe as to prevent control rod insertion when required

i. during these low probability accidents. This Appendix describes the review that should be performed of the fuel assembly structural response to seismic and LOCA loads. Background material for this Appendix is

! given in References 37-40.

B. ANALYSIS OF LOADS

. 1. Input Input for the fuel assembly structural analysis comes from results l of the primary coolant system and reactor internals structural 4

analysis, which is reviewed by the Mechanical Engineering Branch.

Input for the fuel assembly response to a LOCA'should include (a) motions of the core plate, core shroud, fuel alignment plate, or other relevant structures; these motions should correspond to the .

break that produced the peak fuel assembly loadings in the primary coolant system and reactor internals analysis, and (b) transient i pressure differences that apply loads directly to the fuel assembly. ,

i If the earthquake loads are large enough to produce ~a non-linear

fuel assembly response, input for the seismic analysis should use

! structure motions corresponding to the reactor primary coolant

system analysis for the SSE; if a linear response is produced, a

' spectral analysis may be used in accordance with the guidelines of Regulatory Guide 1.60 (Ref. 41).

l

2. Methods .

-- Analytical methods used in performing structaral response analyses i should be reviewed. Justification should be supplied to show that i the numerical solution techniques are appropriate.

i i Linear and non-linear structural representations (i.e., the modeling) should also be reviewed. Experimental verification of the analytical representation of the fuel assembly components should be provided ..

4 when practical. y 4.2-18 Rev. O L

m

^

A sample problem of a simplified nature should be worked by the applicant and compared by the reviewer with either hand calculations 1 or results generated by the reviewer with an independent code (Ref. 38).

Although the sample problem should use a structural representation that is as close as possible to the design in question (and, therefore, would vary from one vendor to another), simplifying assumptions may be made (e.g. , one might use a 3-assembly core region with continuous sinusoidal input).

The sample problem should be designed to exercise various features of the code and reveal their behavior. The sample problem comparison is not, however, def.igned to show that one code is more conservative than another, but rather to alert the reviewer to major discrepancies so that an explanation can be sought.

3.' Uncertainty Allowances The fuel assembly stru tural models and analytical methods are probably conservative and input parameters are also conservative.

However, to ensure that the fuel assembly analysis does not introduce any non-convervatisms, two precautions should be taken: (a) If it is not explicitly evaluated, impact loads from the PWR LOCA analysis

' should be increased (by about 30%) to account for a pressure pulse, which is associated with steam flashing that affects only the PWR fuel assembly analysis. (b) Conservative margin should be added if any part of the analysis (PWR or BWR) exhibits pronounced sensitivity o input variations.

Variations in resultant loads should be determined for 210% variations in input amplitude and frequency; variations in amplitude and frequency should be made separately, not simultaneously. A factor should be developed for resultant load magnitude variations of more than 15%.

For example, if 110% variations in input magnitude or frequency produce a maximum resultant increase of 35%, the sensitivity factor would be 1.2. Since resonances and pronounced sensitivities may be plant-dependent, the sensitivity analysis should be performed on a plant-by plant basis until the reviewer is confident that further sensitivity analyses are unnecessary or it is otherwise demonstrated that the analyses performed are bounding.

4. Audit V

Independent audit calculations for a typical full-sized core should be performed by the reviewer to verify that the overall structural representation is adequate. An independent audit code (Ref. 38) should be used for this audit during the generic review of the analytical methods.

5. Combination of Loads To meet the requirements o# General Design Criterion 2 as it relates to combining loads, an app opriate combination of loads from natural phenomena and accident conditions must be made. Loads on fuel assembly components should be calculated for each input (i.e.,

seismic and LOCA) as described above in Paragraph 1, and the resulting loads should be added by the square-root-of-sum-of-squares (SRSS) 4.2-19 Rev. O

Y method. These combined loads should be compared with the component .

strengths described in Section C according to the acceptance criteria s in Section D. i C. DETERMINATION OF STRrdGTH

1. Grids All modes of loading (e.g., in grid and through grid loadings) should be considered, and the most damaging mo'de should be represented in the vendor's laboratory grid strength tests. Test procedures and results should be reviewed to assure that the appropriate failure mode is being predicted. The review should also confirm that (a) the testing impact velocities correspond to expected fuel assembly velocities, and (b) the crushing load P(crit) has been suitably selected from the load-v.s-deflection curves. 3ecause of the potential for different test rigs to introduce measurement variations, an evaluation of the grid strength test equipment will be included as part of the review of the test procedure.

The consequences of grid deformation are small. Gross deformation of grids in many PWR assemblies would be needed to interfere with control rod insertion during an SSE (i.e., buckling of a few isclated grids could not displace guide tubes significantly from their proper location), and grid deformation (with'out channel deflection) would not affect control blade insertion in a BWR. In a LOCA, gross deformation of the hot channel in either a PWR or a BWR would result in only small increases in peak cladding temperature. Therefore, average values are appropriate, and the allowable crushing load )

P(crit) should be the 95% confidence level on the true mean as taken from the distribution of measurements on unirradiated production grids at (or corrected to) operating temperature. While P(crit) will increase with irradiation, ductility will be reduced. The extra margin in P(crit) for irradiated grids is thus assumed to offset the unknown deformation behavior of irradiated grids beyond P(crit).

2.

Strengths of fuel assembly components other than spacer grids may be deduced from fundamental material properties or experimentation.

Supporting evidence for strength values should be supplied. Since structural failure of these components (e.g. , fracturing of guide tubes or fragmentation of fuel rods) could be more serious than grid deformation, allowable values (about 95%) of the distribution of component strengths. Therefore,Jg3EEb alues and procedures may be used where appropriate for determining yield and ultimate strengths.

Specification of allowable values may follow the ASME Code require-ments and should include consideration of buckling and fatigue effects.

I 4.2-20 Rev. O

1

. e l

~

0. ACCEPTANCE CRITERIA .

1. Loss-of-Coolant Accident Two principal criteria apply for the LOCA: (a) fuel rod fragmentation must not occur as a direct result of the blowdown loads, and (b) the 10 CFR Part 50, $50.46 temperature and oxidation limits must not be exceeded. lhe first criterion is satisifed if the combined loads on the fuel rods and components other than grids remain below the allowable values defined above. The second criterion is satisfied by an ECCS analysis. If combined loads on the grids remain below P(crit), as defined above, then no significant distortion of the fuel assembly would occur and the usual ECCS analysis is sufficient. If combined grid loads exceed P(crit), then grid deformation must be assumed and the ECCS analysis must include the effects of distorted fuel assemblies. An assumption of maximum credible deformation (i.e., fully collapsed grids) may be made unless other assumptions s

are justified.

Control rod insertability is a third ' criterion that must be satisfied.

Loads from the worst-case LOCA that requires control rod insertion must be combined with the SSE loads, and control rod insertability must be demonstrated for that combined load. For a PWR, if combined loads on the grids remain below P(crit) as defined above, then significant deformation of the fuel assembly would not occur and con-trol rod insertion would not be interfered with by lateral displacement l of the guide tubes. If combined loads on the grids exceed P(crit),

then additional analysis is needed to show that deformation is not severe enough to prevent control rod insertion.

\

For a BWR, several conditions must be met to demonstrate control blade insertability: (a) combined loads on the channel box must ,

remain below the allowable value defined above for ccmponents other than grids; otherwise, additional analysis is needed to show that deformation is not severe enough to prevent control blade insertion, and (b) vertical liitoff forces must not unseat the lower tieplate from the fuel support piece such that the resulting loss of lateral fuel bundle positioning could interfere with control blade insertion.

2. Safe Shutdown Eartneuake Two criteria apply for the SSE: (a) fuel rod fragmentation must not occur as a result of the seismic loads, and (b) control rod inserta- '

bility must be assured. The first criterion is satisfied by the .

. criteria in Paragraph 1. The second criterion must be satisfied for

• SSE loads alone if no analysis for comoined loads is required by ,

Paragraph 1. ,*

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• ANSI /ANS 57.51981

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American National Standard for Light Water React 6rs Fuel Assembly Mechanical Design and Evaluation For more

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information on this and other ,

ANS Standards.

fillout the attached cSd.

i i

! etariat trican Nuclear Society Prepared by the American Nuclear Society Standards Committee Working Group ANS-57.5 Published by the American Nuclear Society 555 North Kensington Avenue La Grange Park Illinois 60525 USA Approved May 14,1981

) by the American Nuclear Standards Institute, Inc.

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1 A m 9rlCa D An American National Standard implies a consensus of those substantially con-011000I cerned with its scope and provisions An American Netional Standard is intended

()

as a guide to aid the manufacturer, the consumer, and the general public. The Standard existence of en Arnetican 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 cautioned 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 la'tcr than five years from the date of publication. Purchasers of this standard may receive current hformation, including interpretation, on all standards published by the American Nuclear Society by calling or writing to the Society.

i J

Published by American Nuclear Society l 555 North Kensington Avenue, La Grange Park, Illinois 60525 USA t

, Price: S24.00 i

l Copyright O 1981 by American Nuclear Society.

Any part of this Standard may be quoted. Credit lines should read " Extracted from American National Standard ANS!IANS-57.5-1991 with permission of the publisher, the American Nuclear Society.** Reproduction prohibited under copyright convention unless written per-mission is granted by the American Nuclear Society. ,

e Printed in the United States of America L

l 4  %

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. . _. . .. l Foreword IThis Foreword is not a part of Arnerican National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluation. ANS!!ANS-57.51981.)

(3)

This American National Standard provides a procedure for determining the mechanical adequacy for 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.

The working group responsible for this standard ANS.57.5, was originally (August 1973) organized as Working Group ANS-13.1 under Subcommittee ANS-13. The sub.

committee voted to dissolve when work was sufficiently advanced on this standard and a companion quality assurance standard developed by Working Group ANS-13.2. The effort was then placed under the auspices of Subcommittee ANS-37, now ANS-55, Fuel and Waste Management.

. The initial scope of the project was to consider only upset, emergency and faulted plant conditions as they affect fuel. The werking group felt that a proper understand-ing of transiant behavior depended on a knowledge of how the plant has been operating undar normal conditions, and so the scope of the standard was expanded accordingly. /

t-i Suggestions for improvement of this standard are welcome. They should be sent to El the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60525.

l The membership of Working Group ANS 57.5, at the time it submitted Revision 1 of this standard, was as f 91ows:

T. G. Danlap. General Electric Compa ty l R. C. Schreiber. Chairman. Buttelte Pacific North. J. W. Heard. Yankee Atomic Electnc Company teest Laboratories H. B. Meieran, Consultant M. P. Bahn, EGaG Idaho. Inc. W. S. Neckodom Erzon Nuclear Company, fr.c.

F. M. Bordelon. Westinghouse Electric Corpontion W. D. Wohlsen. Cc> Chairman. Combustion Engi.

E. J. Brown. U. S. Nuclear Regulatory Commission C. G. Dideon. Babcock and Wilcox Compa"y neering. Inc.

We also wish to thank for their participation:

G. Anderson. General Electric Company P. St.en. Southern California Edison Company l

(now Director. Joint Center for Graduate Study

\

p A, Deajen. EG4G Idaho, Inc.* in Richland, Washington)

} S. B. Kim. U. S. Nuclear Regulatory Commission L. Walton. Babcock and Wilcox Compa ty W. Moore. Southern California Edison Company l

(now with Washington Public Power Supply W. Willoughby. Stone and Webster Corporation *

( A. Roberts, Electric Pouer Research Institute System) l l

0

• Members of the working group during the development of the original standard.

l g, ,

The American Nuclear Society's Nuclear Power Plant Standards Committee A (NUPPSCO) had the following membership at the time of its approval of this standard. V J. F. Mallay, Chairman M. D. Weber, Secretary Name of Representative Organitation Rep'esen ted G.A.htlotto. . U.S. Nuclear Regula rary Com mission R. E. Basso . . . . Cate.:ytic. inc.

R. G. Denham . . ....... ..... . ... . General A tomic Company (for the Institute of Electrical and Electronics Enginsers Irc1 R. E. AUen(Alt.) ... ..... . . . . . . United Engineers & Constructors. Inc.

(for the institute of Electrical and Electronics Engineers Inc.I R. V. Bettinger . ... . Pacific Gas and Electric Company V. Bradbury . . . ., . Westinghouse AdvanecdReactorDivisia s D. A. Campbell . . Wesringhouse Electric Corporation

'*j C. O. Coffer . . .. .KaiserEngineers L. J. Cooper . . Nebraska Pu blic Pou erDistrict W. H. D'Ardenne . . . GeneralElectric Company F. X. Gavige n . . U.S. Depa rtmen t of Energy C. J. Gill . . . Bechtel Porcer curporation H. J. Green . . . Tennessee Va!!ey Authority A. R. Kasper . Combustion Engineering. inc.

W. John son ........ . Cata!) ric. Inc.

R. W. Keaten . . . . .GPUServices Corporation J. W. Lentsch . . Portland GeneralElectric Company D. M. Leppke . . . Fluor Pou er Services. Inc.

J. F. Mallay . . . . . Babcock & Wilcox Company

'(far the American Nuclear Societyl A. T. Mclin . . United Engineers and Constructors J. H. Noble . . . . . . . Cha s T. Main. Inc.

E. P. O'Donnell. . .. ... ... Eba sco Scrices. Inc.

T.J.Pashos. ...

Ifor the Atomic Industrial Forum)

.QuadrextNuclear Services Corpcra tion

(

M. E. Hemley .

....... . . Rock u eillnrernational 3.Stacey . . . Yank er A tomic Electric company S. L. Stamm . . Stone & WebsterEngineering Corpnration J. D. Stevenson . . . . . . . . . . . . . . Structura! Mechanics Associate.*

(for the American Society of CivilEngineers)

G. Wsgner .Commanu ealth Edison Company

3. E. Ward . .. . Sargent and I.u ndy G. L. Wessman . ... . . . . Torrey Pines Tech nology J. E. Windhorst . ... ... . Southern Co:npany Services. Inc.

(for the American Society of Mecha,icalEngineers)

E. R. Wiot . .NUS Corporation G

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Av a. s as....w.~r-s -,. -

Page S

Q Contents ection

.1

1. Scope.. . . .

.1

2. Purpose . ..

.1

3. Definitions .

.2

4. Compliance .

Design and Evaluation .2

5. .

_1 Design Conditions. .

.2 2 FunctionalRequirements . .

.2 5.3 Design Parameters .3 5.4 Limits and Margins. .5 5.5 Specific Requirements for Design . .

.6 Documentation Requirements .8 6.

6.1 Objectives .

.8 6.2 Format. . . .

.9

.9 6.3 Content . .

.9

7. References . . .

Appendix A ANS Design Conditions .10 Appendix B - Illustrationof the Useof th Standard . .14 g)

• .15 Table 1 - Matrix . .

O

m

~3 c) Light Water Reactors Fuel Assembly Mechanical Design and Evaluation which ensures that some aspect of a functional

1. Scope requirement for that event is met.

This standard establishes a procedure for per- (5) A procedure whereby the designer is re-forming an evaluation of the mechanical design quired to dc:ument that the fuel assembly of fuel assemblies for light water cooled com- design has been evaluated in accordance with the limits discussed in 2.4, and has been shown mercial power reactors. It does not address the various aspects of neutronic or thermal- to fulfill each functional requirement for each hydraulic performance except where these event.

factors impose loads or constraints on the mechanical design of the fuel assemblies. The 3. Definitions standard provides definition of design condi- This standard includes a number of terms that, tions. It also presents a list of functional re. while they have several meanings in common quirements representing the general attributes usage, have only one meaning within the stan-of fuel assemblies. This standard also~ includes a dard. Therefore, in order to reduce the possibil-set of specific requirements for design, various ity for misinterpretation of the standard, the potential performance problems and criteria following definitions are provided:

aimed specifica!!y at averting them.

designer. The organization that has the respon-NOTE: As used in the conte.mt of this standard, sibility for preparing the fuel assembly design.

the term " mechanical" is best typified by the parameter list _ presented in 5.3 accounted for in design.

E 2. Purpose fuel assembly. The smallest modular unit com-The purpose of this standard is to establish the prised of individual fuel rods, and associated following requirements for the mechanical integral component parts for handling, control, design of initial core or reload fuel assemblies and the evaluation of that design: support, and maintenance of geometry. For boil-ing water reactors (BWRs), the channel that (1) A comprehensive set of functionalrequire- encloses the fuel bundle is included as part of ments for fuel assemblies. the fuel assembly for design purposes.

(2) A procedure whereby "the designer is I required to select the specific events in each of functional requirer ent. A statement of the the four ANS Design Conditions'.

i necessary capability of a fuel assembly.

l (3) A comprehensive list of considerations,in, cluding material properties, chemical reactions, limit. A bounding value of a variable or l irradiation effects, and failure modes, which are known to affect the capability of fuel assemblies parameter in design, which is established to ensure that one or more aspects of a functional to satisfy one or more functional requirements.

requirement are satisfied.

l (4) A procedere whereby the designer is re.

quired to 11 define which considerations affect margin. A quantitative relationship between a the capability of the fuel cssembly to fulfill each design evaluation result for a given event and a I functional requirement under each postulated limit associated with a functional requirement.

l event, and 2) establish an appropriate limit for I each of these considerations, the meeting of shall, should and may.The word "shall" is used i

j to denote a requirement; the word "should" to denote a recommendation; and the word "may" l i Appendix A presents the ANS Design Conditions together with a suggested bst of events associated with each condi. to denote permission, neither a requirement nor l a recommendation.

i tion.

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8 a Am, nca i Ltn nal Standard ANSirANS 5".5 1991

4. Compliance 5.2 Functional Requirements. This section in-Design documentation shall be prepared to cludes the functional requirements that shall be considered for all design conditions discussed O

show how the provisions of this standard are above. The requirements are the general capa-satisfied. Provisions for dissemination of and bilities a fuel assembly shall have. The designer access to such design documentation are beyond shall identify the specific functional require-the scope and authority of this standard. ments in terms of their significance to each of the design conditions.' The specific features of a

5. Design and Evaluation design that fulfill the functional requ:rements may be different for different designs. Likewise, 5.1 Design Conditions. For the purpose of this the specific methods by which the fuel assembly standard, the design conditions for fuel is shown to fulfill its functional requirements assemblies for light water-cooled reactors are may vary for different designs or designers. The the following set of ANS Design Conditions. fuel assembly shall be designed to fulfill the These conditions are deived from American specific functional requirements throughout its National Standards Nuclear Safety Criteria for expected and postulated operating history (i.e.,

the Design of Stationary Pressurized Water during each event that is considered in the Reactor Plants, N19.21973 (ANS-51.1) and design). A fuel assembly shall be designed to Nuclear Safety Criteria for the Design of satisfy the following requirements:

Stationary Boiling Water Reactor Plants, (1) Provide and maintain acceptable fuel ANSI / ANS-52.1 1978. [1, 2? ' geometry and position axially and radially, i.e.,

5.1.1 Condition I - Normal Operation and locate the fuel rods within the fuel assembly and Operational Transients. Condition I events are the fuel assembly within the core, those that are expected frequently or regularly (2) Provide for acceptable coolant flow and in the normal ccurse of power operation, refuel. heat transfer.

inc. maintenance, or maneuvering of the plant. (3) Provide a barrier to separate fuel and fis-51.2 Condition II- Events of Moderate Fre. sion products from the coolant. f qm ncy. Condition II events are those that are (4) Allos. fur axial and radial expansion of \

en ected during the life of a fuel assembly and the fuel roda, assembly, and contiguous reactor t%t may result in reactor shutdown. The reac. internals, i.e., all sources of dimensional change.

ter is expected to be capable of a return to power (5) Provide self pport,i.e., be free standing without special fuel inspection or repair pro. when required and offer well-defined resistance cedures being required, to distortion by lateral and axial loads.

5.1.3 Condition III - Infrequent Events. (6) Resist action by fluid forces, i.e., accom-Condition III events are those that may happen modate the effects of vibration, wear, lift, infrequently, if at all, during the life of a fuel cavitation, pressure pulse, and flow instabilities.

assembly. It is expected that such an event may (7) Provide for control of the fission process, result in some damage to a fuel assembly that i.e., provide physical guidance to control rods or might necessitate repair or replacement of the blades; tolerate the presence of burnable poison assembly before normal operation could be rods or chemical shim environment; accom-resumed. modate effects of flux, temperature and 5.1.4 Condition IV - Limiting Faults. Condi. pressure gradients, and transients; endure wear tion IV events are those that are not expected to and impact associated with control element occur during the life of a fuel assembly but are mo tion.

postulated because their consequences include (8) Provide for in-core instrumentation and the potential for release of significant amounts other components associated with fuel of radioactive material. assemblies. This includes such items as burn-able poisons, sources, and plugs, as wc!! as instrumentation.

(9) Accommodate chemical, thermal.

SNumbers in brackets refer to corresponding numbers in mechanical, and irradiation effects on materials, Section 7. References.

8 Appendix A presents typical lists of design condition

• Appendix D presents suggested methods for identifying events. these speedic functional requirements.

2

I 1 -

1 Amnice, Nstier.d Standud A NSPANSC iP.N!

j 1

(4) Coolant Chemistry e.g., corrosion, hydriding, Irradiation embrittle-(5) Neutron Flux h ment, expected interactions, fuel densification, in reactor creep and relaxation. (6) Flow Variations (10) Provide for handling, shipping and core (7) Core Internals hiotion.

loading, i.e., have gripping and contact loca. 5.3.2 Fuel and Control Material tions, holddown springs or other neceseary 5.3.2.1 Physical Features hardware, including provision for loads and (1) Dimensions (2) Geometry compatibility with interfacing equipment.

(11) Provide mutual compatibility for all fuel (3) Density assemblies within the core, i.e., reload (4) Surface roughness, 5.3.2.2 Chemical Composition and partially spent fuel assemblies. Compatibility includes fitup and cross flow (open lattice 5.3.2.3 Material Properties designs). Nuclear compatibility is beyond the (1) Thermal Parameters (a) Thermal conductivity coefficients scope of this standard.

(b) Thermal expansion coefficients (c) Specific heats 5.3 Design Parameters. Design parameters developed for demonstration of a fuel assembly (d) Phase structure transformations (e) Melting temperatures.

design shall be identified and justified. These (2) Mechanical / Physical Parameters parameters are usually in the form of material (a) Young's modulus properties, dimensional characteristics or physical response phanomena that are necessary (b) Poisson's ratio.

to describe or evaluate fuel assembly behavior. (3) Ceramic / Metallurgical Parameters These parameters shall be developed by generally (a) Grain size and distribution accepted engineering methods such as reference (b) Pore size to test and experimental data, experience, (c) Pore size distribution.

analysis, use of reference material and correla. 5.3.2.4 Models or Theories or Ccerelations or Mechanisms q tior s. (1) Pellet cracking 2)

(2) Fission and sorbed gas release Some of these parameters are incorporated into (3) Creep specific models or correlations to permit (4) Irradiation induced swelling designation of performance or mechanical limits. Other parameters are incorporated into (5) Densification (6) Thermal conductivity, including por-specific models or correlations to permit evalua-tion of fuel assembly behavior. osity factors (7) Thermal expansion The designer shall identify parameters and (8) Melting.

Justify their application as employed in his 5.3.2.5 Performance and Mechanical Limits.

evaluation. *It is recognized that not all the See 5.4.

parameters, models, etc., are necessarily treated 5.3.3 Fuel Rod. The fuel rod is treated as a explicitly in design. For example, interaction system addressing all subcomponents except between fuel and cladding may not involve a fuel and control materials, which are covered in calculation of pellet cracking. Likewise, some 5.3.2.

5.3.3.1 Phyelcal Features portion of fuel swelling may be implicit in the densification model. Wherever parameters are (1) Length implicitly handled in design, it is sufficient for (2) Diameters the designer to point this out. The parameters (3) Wall thickness considered by the designer shallinclude, but not (4) Ovality necessarily be limited to, the items listed below. (5) Fuel stack heights (6) Surface roughness including scratches (7) Void and plenum volumes 5.3.1 General Reactor Core Environmental (8) Initial internal pressure Conditions (1) Coolant Temperatures (9) Fill gas composition (10) Inclusion of other nonfuel com-i l

h (2) Coolant Pressures (3) Coolant Flow Rates ponents (i.e., spacer pellets, getters, springs) 3 l

e Awr ien Natiod 5t twtird ANS! ANS-57.51941 (11) Surface condition, including crud (19) Means of plenum support and limiting (12) Closures. of axial fuel redistribution during handling 5.3.3.2 Chemical Composition. Material (springs, sleeves, etc.)

designation of tile fuel rod subcomponents. (20) Fretting 5.3.3.3 Mater'al Properties for Cladding (21) Stress rupture.

and Other Subcomponents as Appropriate 5.3.3.5 Performance and Mechanical (1) Thermal Parameters Limits. See 5.4.

(a) Therrnal conductivity coefficients 5.3.4 Fuel Assembly (b) Thermal expansion co=fficients 5.3.4.1 Physical Features (c) Specific heats (1) Dimensional characteristics required (d) Phase structurcl transformations. for interface between fuel assemblies, reactor in-(2) Mechanical! Physical Parameters ternals and other components (a) Young's modulus (2) Means of radial support (b) Vield strength (3) Means of axial support (c) Ultimate strength (4) Grid and channel dimensions (d) Ductility (5) Means of positioning.

(e) Density 5.3.4.2 Chemical Compositions. Alloys for (f) Poisson's ratio. all assembly structural components.

(3) Metallurgical Parameters 5.3.4.3 Material Properties (a) Grain size (1) Thermal Parameters (b) Anistropy factors (a) Thermal conductivity coefficients (c) Texture coefficients (b) Thermal expansion coefficients (d) flydride orientation. (c) Microstructural transformations (4) Chemical Parameters (d) Specific heat.

(a) Corrosion rates (2) Mechanical / Physical Parameters (b) liydrogen pickup rates (a) Ultimate strength (c) Serface preparation prior to irra- (b) Poisson's ratio diation, incide and outside. ,

(c) Young's modulus {

5.3.3.4 Models or Theories or Correlations (d) Yield strength or l'.!cchanisms (e) Ductility (1) Void volume used for gas accom- (f) Fatigue strength modation (g) Density (2) Creep, creep / collapse, creepirupture (h) Impact strength.

(3) Thermal performance (3) Metallurgical Parameters (4) Water / clad heat transfer coefficient (a) Grain size (5) Thermal expansion (radial, cir- (b) Anisotropy factors cumferential, and axial) (c) Texture coefficients.

(6) Fuel'elad gap conductance, including (4) Chemical Parameters gas compositions and gas thermal conduc- (a) Corrosion rates tivities, and the contributions from non-volatile (b) Ilydrogen pickuplembrittlement fission products rates.

(7) Bowing 5.3.4.4 Models or Theories or Correlations (S) Irradiation growth, including aniso- or Mechanisms tropic correlations (1) Wear (9) Stress relaxation (2) Vibration (frequency and arnplitude)

(10) Fatigue (3) Support stress relaxation (11) Waterlogging (4) Deformation (permanent)

(12) Fuel / cladding interaction (5) Irradiation and temperature induced (13) Stress corrosion cracking materials growth and property changes (14) Corrosion rates (6) Cladding / support compressive forces (15) Ilydriding rates (7) Assembly hold down forces (16) Axial gap formation in the fuel stack (8) Creep (9) Crud Buildup (17) Plastic deformation (18) Stored energy (10) Fracture Mechanics.

(*

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i '* American Lt;na i M 14:.M GSla ' 'i

. 5. M 21 5.3.4.5 Performance and Mechanical N n

ao+ 7d /da\

V Limits. See 5.4. \dN) dN < a e No]

5.4 Limits and Margins 53.1 Limits. Limits shall be established by where the designer for the purpose of demonstrating that the functional requirements (5.2) pertinent imtial length of maximum acceptable

= ..

ao

'.o a given design condition (5.1) are satisfied. defect for growth These limits are establ.shed for the purpose of = number of load cycles necessary to m, -

ensuring that a sufficiently high probability ex- No 1 itiate growth of defect ists of meeting the functional requirements.

d_a_ = crack growth rate, extension per load dN cycle The following limits for structural components shall be addrzssed as a m:nimum for design con- Nd = design number of load cycles ditions I, II, III and IV. dc = critical crack length for sudden fracture.

5.4.1.1 For time-independent effects, either the peak stress (S) or the plastic strain (c) she'l 5.4.1.6 Tha calculated maximum stress in-be demonstrated to be less than its ultimato tensity (Kr) from a single tensile loading of a vale, Su or tu.

component shall be lets than the critical strees intensity (Kg) calculated fer the same cer.di-S < Su or s < tu tions.

5.4.1.2 For components subjected to cyclic loads, thn sm of the terms defined by the Kr<K m mtmber cf cycles in each stress range or strain range (N) divided by the corresponding number p of cycles to failure (N 1) shall be less than 1.0. 5.4.2 Margin. A sufficient vabu ci mam:n -

' shall be demonstrated such that irJterent u :cer- '

Ni taintics in experimental or enalydeal pre W-

- <1.0 tion,: do not result in fcilure to meet tbn tur-ticular functional requirement. The perticblar m thod a designer may use to catablish that na 5.4.1.3 For time-deper. dent effects, the sum- adequate margin exirts is optional, but it shcl.

mation of actual time (t) at a given stress level be selected from one or more of the ioitowmg:

divided by the time to failure (t t) at that stress (1) Probability analpes m which the level shall be lets than 1.0. Also, the summat:on .

vanances of mppen&nt paramders are of creep strain incurred (cc) divided by the creep stadsucaHy comumed. ,

strain to failure (({} shall be less than 1.0. (2) Sensitivity analyses m wFeh tha c

variance of the dependent parameter is t

predicted as a function of the tolerance ranges cf

.i. < 1.0 and < 1.0 (t,); w(cy), the m, dependent variables.

(3) Worst-case analyses in which er.ch :n-5.4.1.4 Any load (P) with potential for cars.'ng dependent variable is deliberatdy biased t "ro-structural instability shall be demonstrated to duce the most adverse predicted depen. nt be less than the critical value of that load (Pc), parameters.

which wo,uld result in crippling of the structure. (4) Combined analyses in which certain in-dependent varial2es are worst-case and otb ra .

P<P, are statistically determined or nominally cheren and weighted for sensitivity.

5.4.1.5 The integral of cyclic crack growth (5) Reference to exp trimental data er o;wa-tional performance widch clearly verifies t's rate (da\gg in a component shallbe demonstrated adequacy of the dcsi:m f ar fulfilling a specific to result in a final crack size that is less than the functional requirement for a given design condi-Q critical value of crack size (ac) for sudden fracture. tion.

5

h

, , Am Hn n .htional Str. brd ANSI ANS 57.5 1931 5.5 Sp cific Requirements for Design. A com- designer shall determine the maximum accep-prehensive set of fuel assembly functional re- tab!c hydrogen content in the fuel and burnable -

quirements and design considerations is poison rods.

presented in 5.2 through 5.4. These are arranged without specific instructions as to how the The designer shall evaluate the effect of a "get-various considerations should be applied so as ter" if it is added to the fuel for reduction of not to place unnecessary constraints upon the hydriding, including the potential for design. However, certain considerations for mechanical and chemical interactions with other designing fuel a::semblies are so basic that they fuel rod components and with primary coolant.

have been included as mandatory requirements.

This portion of the section sets dmm these re- A " getter" is material added to a fuel rod that quire:nents. Although these items are regarded effectively competes with the cladding for free as necessary to acceptable fuel design, they are hydrogen that may be present in the rod.

not regarded as sufficient bases for design. 5.5.4 Fretting Corrosion. Support of the fuel 5.5.1 Material Properties. The following rods by the spacer grids shall control relative geeral criteria shall be met in the selection of motion between the rods and the support sur-material properties for specific evaluations: faces, so that excessive wear of the cladding at (1) The mechanical properties of fuel these surfaces does not penetrate the cladding assembly materials are affected by irradiation. or significantly reduce the capability of the clad-The designer shall evaluate component perfor- ding to withstand operating loads.

mance considering both the nonirradiated and irradiated material properties. The adequacy of the spacer grid design and posi-(2) Irradiation-induced densification shall tioning within the fuel assembly shall be be considered from the standpoint of fuel established by test or analyais under conditions redistribution, power peaking and stored representative of coolant temperature, pressure, energy. flow rate and chemistry.

(3) The effect of temperature on individual (

material properties shall be taken into account. For the establishment of an initial design, the The designer shall use properties appropriate to following criteria apply:

tha expected component temperature. (1) The designer shall account for known or (4) For material property data that are cor- predicted excitation frequencies when related against neutron flux or fluence, care establishing fuel rod grid spacing.

shall be taken to ensure that the neutron flux (2) The test or analysis used to demonstrate energy r.pectrum is addressed. the adequacy of grid design shall include grid 5.5.2 Corrosion.The following criteria shall be cells whose interference with the cladding out-epplied in evaluating the effect corrosion of ruel side diameter is adjusted to account for sources components has upon performance. of the predicted reduction of restraint,i.e., grid (1) Corrosion behavior characteristics of tab stress relaxation, chdding diametral creep, fuel a*embly materials shall be obtained under differential thermal expansion, and lateral grid l conditions representative of the reactor environ- irradiation induced growth. The range of initial I

ment. preset force specified for new assemblies shall be l (2) The effect cf corrosien and crud film considered.

'mildup on heat trarsfer surfaces shall be ad- (3) The designer shall consider redistribu-l drosed in the calculation of pe!!et and cladding tion of flow w 4 thin and between fuel assemblies. .

temr.eratures. examples of regions of redistribution are end fit-(3) The design - shall account for the ef- tings and grids. Cross flow caused by jetting of fects of fabrication processes (i.e., cold work, fluid from core baffle joints shall be considered.

heat treatmeat, stress relief, welding) on corro- (4) The designer shall analyze or test for sion behavior. cladding wear and its effect on related analyses.

' 5.5.3 Control of Ilydriding (Zircaloy Clad- 5.5.5 Fuel Assembly Hold down Force. For ding). In order to minimize clad perforations designs which utilize hold down mechanisms from the mechanism of primary hydriding, it is (e.g., springs) to accommodate hydraulic loads.

h I necessary to limit the total internal inventory of the designer shall show that an cdequate hold-hydrogen (or hydrogenous material). The down force exists considering the following:

l 6

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American National Standnd ANSUA.M+57.5 IM t (1) Stress relaxation for hold down springs. (7) Irradiation-induced densification in the h rate.

(2) Maximum expectcd Condition I flow fuel material.

i8) Irradiation-induced growth and creep of (3) Fuel assembly pressure drop, including the cladding, as well as thermal creep of the possible increases due to crud deposition within cladding.

the assembly.

~

(9) Expected variations in initial fill gas (4) Combinations of fuel assembly and sup- pressure and component dimensions.

porting structures dimensional tolerances. (10) Release of sorbed gases from the fuel (5) Differential thermal expansion between material.

5.5.8 Cladding Collapse. Cledding collapse the fuel assembly and reactor internals.

(6) Fuel assembly irradiation induced refers to the dimpling of cladding into short, un-supported gaps in the fuel column. The fuel rod growth, shall be designed such that the cladding is not 5.5.G Fuel Rod Axial Growth Allowance.

susceptible to collapse from the long term ef-There shall be sufficient axial clearance between the fuel rods and the fuel assembly structure to fects of cladding creep. The designer shall state the criterion for collapse. Demonstration of com-accommodnte expected dimensional changes of the components during the design life of the fuel pliance with this requirement shall reflect the following specific considerations:

assembly, (1) The effect of fuel rod burnup and power level on internal pressure, including the effect of Demonstration of compliance with this require-a conservatively low assessment of fission gas ment shall reflect the following specific con-release.

iderations. (2) Irradiation-induced densification of fuel (1) Pifferential thermal expansion between pellets and the solubility of the fill gas in fuel.

the fuel rods and the fuel assembly structure. (3) The range of power histories to which (2) The effect of tolerances. the, rod is likely to be subj,ected.

(3) Differential irradiation-induced growth between the fuel rods and fuel assembly struc- . ' (4) Combmations of component dirnen-

p. si nal tolerance and fill gas pressure tolerances.

h ture. This should include consideration of axial extens. ion of claddm.g mduced by m. teraction 5.5.9 Rod Bow. It has been observed that some fuel rods (and burnable poison rods) bow during between fuel and cladding. operation. As a result, the lateral spacing be-

, p) The effect of fuel assembly structure tween fuel rods will vary. The designer shall axiar compression and creep. qeantify the amount of rod bow that is 5.5.4 Fuel Rod Internal Pressure. The fuel rod predicted to occur during the life of the fuel rod.

performance is signtftcantly affected by mternal The designer shall consider the following effects pressure in such areas as creep, balloonmg and cf bowing in the evaluation:

coUapse. The designer shall consider the varia- (1) The local variation of the fuel moderator tions in pressure that occur over the life of the volume ratio on the peak local fuel rod power fuel rod. Calculations of fuel rod mtunal pressure shall take into account the following g,ygg')(2 Variations in coolant subchannel diver-effects: sions on the departure from nucleate boiling (1) Differential thermal expansion (axial (DNB) margin.

and radial) between pe!! cts and cladding. (3) Control element operation for designs in (2) Irradiation-induced swelling of the fuel which the expected magnitude of fuel rod bow pellets. would be sufficient to cause the fuel rod to in-(3) The accumulation of nonvolatile fission trude into a control element path, products. 5.5.10 Fuel Assembly Bow. It has been (4) Solubility of the fil! gas in the fuel observed that some fuel assemblies exhibit a material. slight bow upon removal from the core. The ex-(5) The higher temperature of gas contained pected fuel assembly bowing shall be accounted in pellet end dishes, pellet cracks and pellet open for in the design in acccrdance with the following-porosity, than is in the annulus between pellets (1) The designer shall show that the max-and cladding.

imum expacted bow can be accommodated by (6) The release of gaseous fission products the handling equipment and fuel storage 0 from the fuel material. facilities.

- . _ ~ _ __ _

e b-Arn.-rian National Standant ANSilAN%57.54Mt (2) The effect of the fuel assembly bow on tain fission products. It is beyond the scope of ,

control rod motion (e.g., through friction drag) this standard to prescribe the means by which a shall be assessed. the fuel designer assesses the failure (3) The effect of fuel assembly bow on local mechanisms, or to require that fuel be designed povier and coolant flow distribution shall be to limit failures to a particular number of rods.

assessed.

5.5.11 Component Cooling Flow. Where other However, the designer shall state the method cornponents (e.g., control rods, poison rods, used in the assessment of these failure neutron sources, instrumentation) are included mechanisms (e.g., the design features, reactor within the fuel assembly, the design shall allow loading and power maneuvering limitations, fuel for sufficient cooling of these components. The duty) that lead to an acceptably low probability following shall be considered in establishing of failure.

cooling adequacy: 5.5.14 Analytical Evaluation of Stresses and (1) Minimum expected pressure head. Strain. For those components in a fuel assembly (2) Contributors to maximum flow for which mechanical integrity is to be resistance. demonstrated by means of stress analysis, the (a) Component dimensional tolerances following requirements shall apply:

(b) Differential thermal expansion and (1) For evaluating the response of com-irradiation induced dimensional changes ponents subject to multi axial stress conditions, (c) Component heat generation the analysis shall use one of the recognized (d) Surface roughness, component corro- methods for combining such stresses (e.g., man-sion and crud deposition. imum strain energy, maximum resolved shear 5.5.12 Fuel llandling. The expected fuel stress) and shall define the criteria for accept-handling shall be provided for in fuel assembly ability of the result.

design in accordance with the following: (2) For components that are subjected to (1) Each fuel assembly shall be marked in cyclic loading, the cumulative effect shall be such a way that its identity and orientation in determined. The designer shall identify the f the core can be verified. American National method used. \.

Standard for Fuel Assembly Identification, (3) For structural mmponents that may be ANSI /ANS 57.81978, defines a system of subject to significant creep strains as a result of serialization which may be used to satisfy this operationalloadings, the magnitude of the resul-requirement. [3] tant creep strain shall not be sufficient to pro-(2) The fuel assembly shall be capable of duce rupture of the component.

sustaining the loads induced by normal han. (4) For components that are subjected to dling operations. both cyclic loadings and creep strains, tho (3) The fuel assembly shall be designed to designer shall establish an acceptance criterion facilitate core loading without damage. that takes both constant and cyclic loads into (4) The fuel assembly shall be capable of account.

sustaining the expected loads induced by possi-ble reconstitution operations (i.e., the partial 6. Documentation Requirements disas,sembly and subsequent reassembly of ir-radiated fuel assemblies for purposes of inspec. 6.1 Objectives. The objective of this portion of tion, repair, or other reason), the standard is to establish requirements for 5.5.13 Pellet Cladding Interactions and design documentation to satisfy Section 4, Com-Strens Corrosion Cracking of the Cladding, pliance. The documentation shall demonstrate Experience has shown that the cladding of fuel that the fuel assembly design meets this stan-rods might be breached during normal modes of dard with respect to:

reactor operation prior to the designed end of- 6.1.1 Defining the function and desired perfor-life (EOL) condition. In some cases these mance under stated conditions for both the fuel failures have been attributed to stress corrosion assembly as a unit and for individual com-cracking initiated on the inner surface of the ponents as appropriate.

clad due to a combination of local clad stress, 6.1.2 Defining and documenting the set of produced by mechanicalinteraction between the criteria that provide assurance of achievmg the fuel pellet and the clad, and the presence of cer- stated function and performance requirements. {'

8

L, ,

American NatMnal Star.b ril A.*Gi%NF. r 7. .19xl 6.1.3 Demonstrating that the design does, in assure meeting the fuel assembly functional re-O fact. m eet the criteria. 2eiremente ef e ch deelen ce ditien =haii ne established and appropriately recorded in the G.2 Format. The design documentation shall design documentation. A discussion of limits is follow a logical pattern and be free from presented in 5.4.1.

ambiguity. 6.3.6 Design Evaluntion. The documentation shall present the' detailed evaluation and the G.3 Content resulting margin to the limits established. The 6.3.1 Design Conditions. The specific condi- margin shall be established, as per 5.4.2, to tions considered for fuel assembly design shall assure that the functional requirements are be identified according to the ANS Design Con- satisfied. The methods for design evaluation ditions presented in 5.1. Specific events, or com- shall account for appropriate parameters, binations of events for each condition, shall be analysis techniques, experimental testing or described in terms of cause, such as plant pro- operational results. A listing of all reference cess conditions or postulated accident. material utilized in the evaluation shall be pro-G.3.2 Functional Requirements. The fuel vided.

assembly functional requirements appropriate to each design condition shall be established and 7. References documented. The functional requirements shall include the requirements in 5.2 where ap- [1] American National Standard Nuclear Safety propriate for the specific fuel assembly design. Criteria for the Design of Stationary 6.3.3 Design Description. A description of the Pressurized Water Reactor Plants, fuel assembly design shall be presented. The im- N18.2-1973 (ANS-51.1). American Nuclear portant components, dimensions, physical Society, La Grange Park, Ill.

features, and chemical and material properties of the fuel assembly relevant to function and [2] American National Standard Nuclear Safety q performance shall be presented. The physical Criteria for the Design of Stationary Boiling 5/ features, chemical composition, materials pro- Water Reactor Plants, ANSI /ANS-52.11978.

perties, and model subsections of 5.3 shall be American Nuclear Society, La Grange Park, consulted as a guide for important items to be Ill.

presented, j ti.3.4 Fuel Assembly Loading. The quan- [3] American National Standard for Fuel titative loading component or response for Assembly Identification. ANSI /ANS 57.8-l specific design conditions shall be established. 1978. American Nuclear Society, La Grange l

Pertinent environmental parameters (in 5.3.1), Park, Ill.

such as temperature, shall be explicitly stated for each loading combination to be evaluated.

The effects of the design conditions considered When the preceding American National Stan-

, shall be identified and included in the loadings dards referred to in this document are sunerseded established. by a revision approved by the Arnerican G.3.5 Design Limits. Specific design limits, National Standards Institute, Inc., the revision l

l together with the associated justification, which shall apply.

l l

l l

0 l

l 9

f-

Amerit an N.itiorial Storidard ANSI ANS 5711981 1

l Appendix A ()

(This Ap,+-ndix is not a part of Arrerican National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluativa. ANS!!ANS-57.51981. but is included for information purposes only.)

ANS Design Conditions Condition I - No mal Operation and Operational Transients 5 Condition I events are those which are expected frequently or regularly in the course of power opera-tion, refueling, maintenance, or maneuvering of the plant. As such, Condition I events are accom-modated with margin between any plant parameter and 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 correspon-ding to the expected mode of plant operation up to the time of the subject event.

A typicallist of Condition I events is given below:

1. Steady state and shutdown operations (a) Power operation (b) Startup (c) Hot shutdown (suberitical, residual heat removal system isolated)

(d) Cold shutdown (suberitical, residual heat removal system in operation)

(e) Refueling, including fuel har.dling (f) 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 fuel with cladding defects.

(c) Activity in the reactor coolant (1) fission products (2) corrosion products (3) tritium.

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

4. Preoperational and operational testing of systems with fuelin place.

Condition II - Events of Moderate Frequency These events at worst result in the reactor shutdown with tha plant being capable of returning to opera-tion. By definition, these events do not propagate to cause a more serious accident, i.e., Condition III or IV category.The fuel assembly should be designed in anticipation of the frequency of occurrence for ex-pected Condition Il events during the full residence time. Margins to fuel design limits (stress, strain, temperature, creep, etc.) should include provisions for the expected Condition 11 events.

8 Specific design and opa.ational changes are hkely to give rise to the need for considering additional events, but these lists can be recommended as providing a minimum of events to consider. h 10

1 i

Ar.vrica i htiw..-l %m% ret A Nef ANW ..;$1 l

PWR Events i For purposes of illustration, the following events have been grouped into this category for pressurized I ,

water reactors: f Uncontrciled 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 flon Startup of an inactive reactor coolant loop I oss of externalload and/or turbine trip Loss of normal feedwater Loss of offsite power to the station auxiliaries (station blackout)

Excessive her.t 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 purpows ofillustration, the following events have been grouped into this category for boiling water reactors:

Turbine trip without bypass Isolation of any or all main steamlines g

Loss of condenser cooling /

- Loss of feedwater heating Inedvertent moderator cooldown Loss of feedwater flow Totalloss of offsite a c 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 By definition, Condition III events occur very infrequently during the life of the plant. They will be ac-commodated 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 er restrict public use of these areas beyond the exclusion area. A Condition Ill 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.

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. . - - - . . . ..= . - _ - _ _ - . - - . - - - _ - - - - . _ - _ _ _ _ . - - - _ . . - - - - . _ . - - _ _ - _ - . - -

Amnicar: National Standard ANSliANS-57.51981 PWR Events For purposes of illustration, the following events have been grouped into this category for pressurized water reactors:

Loss of reactor coolant, from small ruptured pipes or 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, or seizt re of one main coolant pump Fuel handling accident (minor).

UWR 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 functions ofisolation of containment, emergency core cooling, and reac-tor 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 occurrences are faults that are not expected to take place 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. Con-dition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values of 10CFR Part 100. A single Condition IV fault is not to cause a consequentialloss of required functions of systema needed to cope with the fault in-cluding those of the Emergency Core Cooling System and the containment.

PWR Events For purposes of illustration, the following faults hwe been grouped into this category for pressurized water reactorm Major rupture of pipes 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 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.-

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

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Anwrican Ltions' Sten l.+rri ANS!'ANS ",7 % t' 31 Control rod drop accident l

.- Fuel handling accident resulting in major cladding darnage of an irradiated fuel assembly 6g Major rupture of that portion of the steam line that is not a part of the reactor coolant pre::sure 1

boundary up to and including a double ended rupture of the steam line Major rupture of any pipein the reactor coolant pressure boundary larger than that defined as an in-frequent plant process condition (PPC) and including a double-ended rupture of the largest pipe Safe shutdown earthquake.

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n ro. g Amerir-n Natien.it Standard A.'sS!'ANS-57.51981 Appendix B -

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(This Appendis is not a part of American National Standard for Light Water Reactors Fuel Anembly Mect anical Design and Esatuation. ANSilANS 57.51931, but is included for information purpons only.)

Illustration of the Use of the Standard Implen.entation of the standard may be accomplished in four steps:

1. Prepare a matrix of functional requirements versus the ANS Design Condition Events. (See 5.1. 5.2.

and Appendix A. See Table 1 as an example.)

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, a nd 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 (company policy and procedure). .

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

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