ML20213D141

Draft 1 SRP Proposed Revision 2 to Section 4.2, Fuel Sys Design

ML20213D141
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Issue date:	02/13/1980
From:	Office of Nuclear Reactor Regulation
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ML111220504	List: ML20213D135 ML20213D138 ML20213D141 ML20213D145 ML20213D147
References
CON-WNP-0331, CON-WNP-331 NUREG-75-087, NUREG-75-087-04.2, NUREG-75-87, NUREG-75-87-4.2, PSRP-4.2-R2, SRP-04.02-DRFT, SRP-4.02-DRFT, NUDOCS 8003240273
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation PROPOSED REVISION TO STANDARD REVIEW Pl.AN P5RP-4.2, REV!sION 2, DWT 1 SECTION 4.2 FUEL SYSTEM DESIGN REVIEV RESPONSI8fLITTE5 PMeary - Core Perfomance Branc5 (CPS)

Secenaary - None

!. A#EAS OF #EVIEW The thereal, eschanical, and esterials design of the fuel system is evaluated Dy CPS.

The fuel system consists of: arreys (assemolies or bunales) of fuel roes including fuel pellets, insulator pellets, spMngs, tuoular claading, and closures, hyerogen getters, and fill gas; Durnacle poison rods incluaing components steiler to thest in fuel roes; guise tuoes and other non-fueled tunes; soecer gMas and spMngs; and plates; cnennel Domes; and reactivity control mes. In tne case of the control roes, tais section covers the reactivity control elements that entend from the coupling int rface of the control rod arive mechanism into the core. The Mocnenical Engineering Branen reviews the design of control rod CM we mecnenisme in SRP Section 3.9.4 and the eesign of reactor internals in SRP Section 3.9.5.

The oojectives of the fuel system safety review are to provice assurance that (,) the fuel system is not assagee as a result of nomel operation and anticipated operational occurrences, (b) fuel system canage is never so severe as to prevent control rod inser-tion wnen it is reoutred, (c) the neeer of fuel rod failures is not unserestimated for postulated accisents, and (d) coolacility is always maintained. "Not samagea," as used in the aoove statement, means that fuel rods do not fail, that fuel system dimensions remain within operational tp'erences, and that functional capaci11 ties are not reduced below those assLand in the safety analysis. " Fuel rod failure" means that tne fuel rod leets and that the first fission proeuct barMer (the cleading) has, therefore, been breeched. Coolanility, in general, seens that tne fuel assemely retains its roebundle geometry with aceouate coolant channels to nemit removal of restaual Meet after accioents analyzed in Chapter 15.

This orcoosea -evision of one 5tancara deview nan ano sne suoport vaiuanmoect statement nave not received a coeolete staff review ano aooroval anc co not reoresent an official NRC staff oosition. Puolic comments are peing solicited on noth tne revision ano tne value/fseact statement (incluaing any teolementation senecules) prior to a review cy the Regulatory Require-eents Revtew Committee and taetr recommenoation as to wnether tais veiston snould te aooroveo.

onenents snould to sent to the Secretary of the Cosmtssion, U.S. Nuclaar 4egulatory Commission.

'.asnington. 0.C. 20555, Attention: Joctating anc Service Branen. All comments receivec ey util me consic. red oy the Requiatory Reouiremerts Review Committee, a summerf of sne meeting of the Cosmittee at enten tais revision is constoerec, the Committee recommencations ano all of the associatec coctments and comments considerec by the Committee will se saae puolfely availaole prior to a oecision ey the 31 recur, Office of Nuclear teactor l 4ewi s tien , on eetmer to %1ement this evisten.

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Fuel failure cMteria and coelatility criterta that involve thermel-hydraulic considere-tiene are provided by the Core Perfomance trench to the Analysis Brench for feelementa= l tien in SRP 5ection 4.4 The Analysis Branch provides hydraulic leads under SRP 5ection 4.4 to taa Care Pseformance_ Branch for evaluetten (la S W 5ection 4.2).of fuel asseemly anchemical reopense unser nemel and accident conditions. The availatie radio-active fission product inventory fn fuel rods (i.e., the ger inventary empressed as a release fraction) fs provided to the Accieent Analysis Branch for use in estinating the radiological consequences of plant releases.

The fuel system review covers the following specific areas.

A. Desi e 8ases The principles and related assuntions of the fuel system design should be reviesed.

These bases any be expressed as explicit nemmers or as general cMteMa. The meses will inclues traditional fuel design lietts, inaustry codes and staneerde, and limits related to the safety analysis (i.e., related to fuel ammage, red failure, or coolatility reeuirements). Once such limits are approved in the safety evalue-tion report, they become the specified acceptacle fuel design lietta referred to in General Design CM terion 10 (Ref. 1). The ensign bases should reflect the safety review objectives as anscMbed aseve.

B. Oe'scriotion and Desion Orarines The fuel system esecMotion and design erewings are reviewed. In general, the aescMotion will esonesize proeuct soucifications rather than process soecifications.

C. Desfon Evaluation The perfomance of the fuel system during neruel operation, anticipitated opere-tional occurrences, and postulated accidents is reviewed to estermine if all design bases are met. The fuel system components, as listed aseve, are reviamed not only as separete commenents but also as integral units such as fuel rods and fuel assemblies. The review consists of an evaluation of operating experience, direct l expeMeantal compeMeens, esta11ed mathematical analyses, and other infomation.

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D. Testine. Insoection, and Surveillance 81ans Testing and inspection of now fuel is performed by the licensee to ensure that the l fuel is faericated in accareence with the design and that it reacnes the plant site and is loaces in the core without damage. On-line fuel rod failure monitoring and postirradiation surveillance should be performed to ostect anomalies or confim that the fuel system is performing as expected; surveillance of control rods con-taining $ 40 should be performed to ensure against reactivity loss. The testing, inspection, ano surveillance plans along with their reoorting provisions are l

' reviewea by CPS to ensure that the iscortant fuel cesign consioerations have been amoressed.

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!!. ACCEMANCE CRITERIA A. Desion Bases The fuel syntes assign bases must reflect the four oejectives described in Subsection I, Areas of Review. To satisfy these objectives, acceptance criteria are nessed for fuel systas damage, fuel red failure, and fuel ecolatility. These criteria are discussed in the following:

1. Fuel System Damece

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Fuel system assage includes fuel red failure, maich is discussed below in i Sesection II-4-2. In addition to preclueing fuel red failure, fuel damage  !

criteria should assure that fuel system dimensions remain within operational tolerances and that functional cease 111 ties are not reduced below these assumed in the safety analysis. Such esmoso criteria should inclues the following *.o be complete.

(a) Stress, strain, or loading limits for spacer grios, guide tees, thiamles, fuel rees, centrol rees, enannel bones, and other fuel system structurel meneers should be provided. Stress Itaits that are attained Dy metness steilar to these given in Section III of the ASIE Caes (Ref. 2) am acceptable. Other proposed limits met be justified.

(b) The cumulative nemer of strain fatigue cycles on the structural members sentioned in peregraph (a) above should be significantly less than the design fatigue lifetime, which is based on appropriate este and incluses a safety facter of 2 on stress amelitues or a safety factor of 20 on the masser of cycles (Ref. 3). Other proposed Itaits must be justified.

(c) Fretting wear at contact points on the structural semeers sentioned.in paragreen (a) aseve should be limitad. The allaseele fretting wear

• should be stated in the safety analysis resort and the stress and fatigue limits in peregreens (a) and (b) aseve should presmo the existence of this user.

(d) 0xidation, hydriding, and the bailaw of corresten preaucts (crud) should be limited. Allousele oxidation, hydriding, and crud levels sneuld be discussed in the safety

• analysis remert and shown to be accostacle.

These levels sneuld be presumed to exist in paragreens (a) and (b) aseve.

The effect of crud on theruel- p ulic considerations is revisued by the Analysis arench as described in SRP Section 4.4 '

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(e) Oimensional enanges such as rod bowing or irreciation growth of fuel roes, control rods, and guide tunes need not ne limited to set values (f.e. , canage Itaits), but they sust De incluced in the cesign analysis

• o estaolisn operational tolerances.

4.2-3 Prooosed Rev. 2 3 raft I

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(f) Fuel and Durmaale peisen red intamal gas pressures sneuld recein Delow the nominal system pressure duMag neraal operation unless otherFise justified.

(g) Worst case hydredits lease for nornel operation should not escoed the helddome caess111ty of the fuel assently (either gravity or helddeun springs). Hydraulic leads for this eyelustion are provided lyy the Analysis Rfanen as described in SRP 5ection 4.4.

(h) Centrol rod reactivity must be enf atalped. This any reestre the control rods te rossin watertignt if water-soldle er leechsele seterials (e.g.,

B4 C) am used.

2. Fuel Red Fetture Fuel red fatture is defined es the fees of fuel red hermeticity. Although we recognize that it is not peestble to avoid all fuel red failures and that claene systems are installed to handle a small numer of teeming rede, it is the cejoctive of the review to assure that fuel does not fail eue to kneun failure ascnenises duM ng neraal operation and anticipated operational occwtonces. Fuel red failures can be caused by overnesting, pellet / cladding interaction (PCI), hydHding cladding collapse, bursting, anchenical fractur-ing, and fretting. A fuel fatture cMtarten should De given for each knen failure anchenism. Such criteM a should address the following to be complete.

(a) Overnesting; No useful nochanistic criteria exist at present for fuel rod failure due to evernesting. Homover, to show that overnesting will be aveiaed, it will De sufficient to show that (1) cladding tasserstures ao not greatly exceed the coolant tessereture and (2) fuel melting does not occur.

Adequate cooling is asemed to exist when the theres1 sergin cMterion to limit the deearture free nucleate boiling (DIG) er boiling treneition condition in the care is satisfied. The review of this criteHon is detailed in SAP Section 4.4. ,

For a severe reactivity initiated accident (RIA), Regulatory Guise 1.77 (Ref, 4) relies on a DNS cMteMoe for ootermining failures in PWRs, wrerees a radial average energy density of 170 cal /g is accooted for 9WRs unser zero and low power conditions. Other lielts may De more accurate for an RIA, but continued aggrovel of these lisits say De given until j generic studies yield isprovements.

Althougn a DNS eMterion is sufficient to demonstrate tt.e avo10ance of ,

I overneating free a ceficient cooling secnanise. it is not a necessary condition (i.e., DNB is not a failure SeChanisa) and other teenanistic sethods say to accastacle. Althougn there is at present little exoerience

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with other aeproaches, positions rocameneing different criteria should aderoes cleading temersture, pressure, time durction, exidation, and serittlement.

The second criterien used te assure that the cleeding does not overnest is that fuel amiting will not occur. There would otherwise be concem that metten fuel signt contact the cleading and cause local hotspeta.

This cMterien aise avoids the axial relocation of molten fuel that could cause local overnesting.

(b) Pellet / Cladding Interaction (MI): There is no current cH teH on for fuel failure resulting from MI, and the design bests can only be stated generally. Te mleted cM tem a should be applied, but they are not sufficient te preclude PCI failures: (1) the unifem strain of the cladding should not exceed IL In this context, unifere strain (elastic and inelastic) is defined as transient-insuced deforestion with gage lengths corressending to cleading dimensions; steady-state creeposun and irrestation growth are ancluded. Although observing this strain limit may preclues some MI failures, it will not preclude the corrosion-essistee failures that occur at low strains, nor will it preclude highly localized overstrain failures. (2) Fuel setting should be avoided. The large velee increase associated with molting any cause a pellet with a molten care to exert a stress on the cladding. Sucn a MI is avoided by aveicing fuel aslting. Note that this same criterion uns invenes in peregraph (a) l to ensure that overnesting of the cleading usuld not occur. ,

(c) Hydriding: NyeMdtng as a cause of failure (i.e., primary hyeriding) is prevented by toeping the level of asisture and other hydrogenous tasuM-ties very low ouM ng faeH cation. Acceptable moisture levels for Zircaloy-clad urenfias oxide fuel saould be no greater then 20 pse. Carrent ASTN specifications (Ref. $) for UD 2 fuel pellets state an equivalent limit of 2 ppe of hydrogen free all soumes. For other esterials clad in Zircaloy teing, an equivalent quantity of asisture er hyerogen can be talerated.

3 A asisture level of 2 og HgG por cm of het void volume within the Zircaloy cleeding has been shoun (Ref. 6) to be insufficient fer priesty hyeries forestion.

(d) Claeding Callasse: If axial gaes in the fuel pellet column occur d.m to consification, the cladding has the notantial of collassing into a gao ,

(i.e., flattening). Because of the large local strains that accousany this process, collapsee (flattenea) cleading is assumed to fail.

(e) Sursting: Zircaloy cladding will turst (ructure) unoer certain coastna-tions of tem erature, heating rate, and differential pressure. Althougn i

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fuel suppliers any use different rupture-temperature vs differential-pressure curves, an acceptable curve saould be steilar to the one deter-eined by cea Ridge motional Laserstory (Ref. 7). This cM teH on is included in the ECCS evaluetten aseel roeutrod by Appendix K (Ref. 8).

(f) Mechanical Fracturing: A eschenical fracture refers to a defect in a fuel rod caused by en externally applfed force such as a hyereulic load or a load derived free core-plate estion. Cladding integMty any be assumed if the applied strees is less than 905 of the irradf ated yield strees at the appropMate temperature. Other proposed lietts must be justified.

(g) Fretting: Fretting is a potential cause of fuel failure, but it is a greasel process that would not be effective duMng the bMef duration of an annereal operational occurrence or a postulated accident. Therefore, the fretting weer requirement in peregraph (c) of Subsection II-A-1. Fuel Qaeoge, is sufficient to preclues fuel failures caused Dy fretting auMng transients.

3. Fuel Csolatility Coolatlity has traditionally feelted that the fuel assemely retaira its roe-bundle geometry with aesquete coolant channels to pereit renoval of resid-ual heat. Resuction of coolesility can result free cladding ee M ttlement, ,

violent eagulston of fuel, generalized cladding salting, gros, structurel eefometton, and extries coplanar fuel red ballooning. Coolant 11ty crf *. aria should inclues t k following to be complete:

(a) Claeding Emerittlement: Oxygen contamination and hyorising in Zircaloy cleading are the pMeery casses af cleeding eseMttleennt. For the UICA, Appendia K addresses thace phenomena with a criterion of 2200*1 post cleading temperature and a criteMan of 17I sexismas cladding oxidation.

(Note: If the cladding were precicted to collapse in a given cycle, it  ;

would also be predicted to fall and, therefore, should not be irradiated in that cycle; consequently, the lower poes cladding temperature 11stt of 1800*F previously descM bed in Refergnce 9 it no longer neeend in CP and CL revious.) Specific tassersture amt emidation cMteMa have not been aerived for otner serioents, out snould they be neeeed Aopennin t can be wee as guisance.

(b) Violent Expulsion of Fuel: In severe reactivity initiated accidents ,

(RIAs), such as rod ejection in a NR or rod crop in a WR, the large and roofd coposition of energy in the fuel can result in setting, fregnentaa tion, and dispersal of fuel. The escr.anical action associated with Nel cispersal can be sufficier.t to costroy the eleccing arrs the recuncle

. ;eceetry of the 6.e1 and to 3 ochce pressure pulses 'n the primery system.

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- v Observing the 200 cal /g limit specified by Regulatory Gutes 1.77 prevents viessereed fragmentation and dispersal of the fuel and avoids generating pressure pulses in the pM eery systes during an RIA. This 200 cal /g 11 sit should be used for PWRs and IWRs.

(c) Generalized Claeding Melting: Generalized (f.e., non-local) nelting of the claeding could result 'in the less of Wie fuel goemetry.

CH tem a for cladding ese H ttlement in peregraph (a) aeove are more stringent than eelting criteria would be; therefore, aeditional specific cMteria are not used.

(d) Structural Deforestion: Analytical procedures are discussed in Appendix A,

" Evaluation of Fuel Assamely structural ksponse to Externally Applied Forces."

(e) Fuel Red Ballooning: For the UllCA analysis, Aspendia K requires that flow bloctage resulting free cleeding ballooning (euelltag) be taken into account in the analysis of core flow distribution. Flow blockage aseels must be based on appitcable data (Refs. 7, 11, and 12) in such a way that (1) the tamperature and differential pressere at which the cleeding will reture are prooerly settested (see peregraph (e) of Subsection II-A-2), ,

(2) the retultant degree of cleading smelling is not unserestiested, and (3) the associated reduction in assembly flow area is not unserestimated.

The flow blocaage eseel evaluation is provised to the Analysts Brenen for incorporetton in the comprenensive ECT.3 seeml evaluation to show that the 2100*F classing temperature and 17% classing oxidation Itaits are not

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The reviewer should also determine if fuel red ballooning should be incluesd in the analysts of other accidents involving systes espressurization.

B. DescM otion and Desian Drawines The reviewer smould see that the fuel systas descMption cnd cosign drowings are complete enough to provide an accurate reorosentation and to supply inforention nessed in audit evaluations. Completanees is a setter of juegment, but the follow =

fng fuel systas inforestion and associated tolerances are necessary for an accent-able fuel systas descM otion:

fp e and metallurgical state of the cleading 01aading outstae aiameter Claading Insion diameter Cladding 1.d. rougnness Pellet outslos diameter l *ellet roughness i Pellet sensity l Pallet resintering cata pellet lengtn Pellet disn dimensions Burnacle motson content Insulator pellet parameters Fuel column lengtn Overall roa length 4.2-7 P w sed Rev. 2 3 raft i

s Itod internal void vel ee -

Fill gas type and pressure Sorted gas coneositten and content F Soring and plug dimensten Fissile enricament Eeutvalent hydraulic diameter Coelant pressure The following easign drerings have aise been found necessary for an acceptaale fuel system descM ption:

Fuel assembly crees section Fuel assemely outline Ftel red scnematic Spacer gM d cross section Gutes t es and ne u te jetnt Control red assemoly cross section control red assemely outline Control red schematic

• Surmaale poisen red asseusly cross section Burneele petsen red assemoly outline Burneele petson rod senematic 1

QMfice and source assemely outline C. Desion Evaluation The mothens of demonstrating that the design bases are set must be revisued. These mathees inc1wte operating experience, prototype testing, and analytical predictions. Many of these mecnods will be presented generiemlly in tapical reports and will be incorporated in PSAAs and F5ARs by reference.

1. Doeestire Experience Operating experience with fuel systems of the same or similar assign sneuld be described. When adherence to specific assign cMtaria can be conclusively demonstrated with operating experience, prototype testing and assign analyses that were performed prior to gaining that experience nessed not be reviewed.

Design cMteMa for fretting wear, oxication, Nyeri: ling, and crud builos might be addressed in this menner.

2. PMtotvoe Testine When conclusive operating experience is not available, as with the introduc-tion of a design change, prototype testing sneuld be reviamed. Out-of-reactor testa should be performed when practical to determine the eneracteristics of the new design. No definitive reouirements have been developed regarding those design features that east be tested pHor to irradiation, but the following out-of-reactor tests have been performed for tnis purpose and will serve as a guide to the revisier:

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$04cer grid structural tests Control W structural and performance tests Fuel assemoly structural tests (lateral, 4Atal and torsional stiffness, frecuency, and canning)

Fuel assemoly nyaraulic flow tests (1tft ferees. control rod wear vibration, and asseenly wear and life)

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Irr reactor tasting of design features and lese-assemely freadiatien of ediola assemeltes of a new cosign should be reviewed. The following onenemena that have been tasted in this menner in new assigns will serve as a quien to the l reviewer:

Fuel and burnable poison red growth Fuel red bewing Fuel assembly growth Feiel assaemly bewing Channel ben weer and distortion

. Fuel red ridging (PCI)

Crud focussion Fuel rod integM ty Heleesmes spring relaxation Soacer gMd spring relaxation

. Guise tune wear enaracteristics In some cases, in-reactor testing of a new fuel assembly aosign or a new dr:s'pi feature cannet be accomplished pMor to coerstion of a full core of that design. This inability to perform in-reactor testing may result free an incoenatability of the new design with the previous design. In such cases, special attention should be given to the surveillance plans (see Sasection II-0 below).

3. Analytical Pmdictions Some assign bases and related persw tpes can only be evaluated with calcula-tional procoeures. The analytical methods that are ' sed u to make performanca predictions sust be rerirmed. Many"sucn reviews have been performed estan-lishing numerous exaeoles for the reviewer. The following paragreens aiscusa the mere established review pattams and previos eeny related references.

(a) Fuel Temperetures (Stored Energy): Fuel tesseretures and stored energy

duM nq norw l operation are neer.ad as input to ECCS performance calcula-tions. The temper *%re calculations require complex camouter coces that mesel eeny diff

• N pnenemana. Phenomenological aseels that snculd be l reviour fedkas :, e following:-

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N y ~ er distM bution Pw.4 / cleosing taseerature distribution Burnup asatMbution in the fuel Thermal consuctivity of the fuel, cladding, cleading crud. and oxication layers Densification of the fuel Thornel expansion of the fuel and cladding Fission gas production and release Solfd and gaseous fission orecuct swelling Fuel restructuring and relocation Fuel and cleading disonstenal enanges

' Fuel-to-cladding neat transfer coefficient Thermal conouctivity of the gas eixture Themel conouctivity in the Knuasen acesin Fuel-to-claccing contact pressure Heat c3saCity of the fuel and cladding Growth ano creep of the clascing Rod internal gas pressure and cacosition -

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• A Serption of helfim and other fill gases Cladding oxide and crud layer thickness Claeding-to-coolant heat transfer coefficient

• 8ecause of the strong interaction betmoen these andels, overall code behavier aust be checked against esta (staneart! preeless or benchmerts) and the MAC audit codes (Refs.13 and 14). Ezamples of previous fuel performance caen revious are given in References 15 througe 18.

(b) Denstfication Effects: In addition to its effect on fuel temperatures (discussed above), densification affects (1) core poner distributions (power spiking, see SRP 5ection 4.3), (2) the fuel lineer heet generation reta (LH8R, see SRP Section 4.4), and (3) the potential for cleeding collapse. Osnsification aspittmeos faa power spike and I.NER analyses are discussed in Reference 15 and in Regulatory Guide 1.128 (Ref. 20).

Mosels for cleeding collages tiens aust aise be reviewed, and previous review examples are given in References 21 and 22.

(c) Fuel Rod Bowing: Guidance for the analysis of fuel red bowing is given in Reference 23. Interie methods that any be used prior to compliance with this guidance. are given in Reference 24. At this writing, the causes of fuel red bewing are not well understood and anchenistic analyses of rod bewing are not being approved.

(d) Structural Deformation: Acceptance criteria are discussed in Ascendix A.

" Evaluation of Fuel Assamely Structurel Respo6:se to Externally Applied Forces."

(e) Reture and Flow 81ockage (Ballooning): Zircaloy rupture and flow block-age models are part of the EC 5 evaluation assel and should be reviewed by CPE. The models are amoirical and should be compared with relevant data. Ezameles of such data and a previous review are contained in References 7, 11, 12, and 28.

(f) Fuel Red Pressure: The thereal performance coes for calculating tempere-tures discussed in paragraph (a) above samid be used to calculate fuel roa pressures in conformance with fuel damage criteria of Subsection II-4-1, paragraph (f). The reviamer should ensure that conservatisas that were incorporated for calculating temperat',tres do not introduce nonconservatises with regard to fuel rod pressures.

(g) Metalhater Reaction Rate: The rate of energy release, nycrogen generatier and cladding oxidation from the metal / water reaction should be calculated using the Baker-Just ecuation (Ref. 27) as required by Aeoendix K. For non-LCCA acclications, other correlations say be used if justified.

"Althougn neeoed in fuel performance coces, this model fs reviewed by the Analysis 3ranen as coscribec in SAP 5ection a.a.

Procesed Rev. 2 a.2-13

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l (h) Fission Product Inventary: The availante radioactive fission product inventory in fuel reos (f.e., the gap inventory) is presently specified by assumptions in Regulatory Guides (Refs. 4, 28-30). These ass eptions

._ _ should,be _used until toproved calculational methods are approved by cpg (see ilef. 31).

D. Testine. feepection. and Surveillance plans Plans must be reviamed far each plant for testing and inspec". ten of now fuel and for monitoring and surveillance of irradiated fuel.

1. Testine and Insoection of een suel Testing and inspection plans for new fuel should include verification of '

cleading integrity, fuel systas dimensions, fuel enrichmeet, burnante petson concentr'stion, and absorter coupesition. Details of the menefacturer*s teet*

in'g and inspection programs sneuld be Idecimented in quality centrol reports, wt:fta should be referenced and sumerized in the safety analysis report. The program for on-site inspection of new fuel and control assemblies after they have been delivered to the plant should also be oescribed. W ere the overall testing and inspection programe are essentially the same as for previously approved plants, a statement to that effect should be ande. In that case, the details of the programs need not be included in the safety analysis report, but an appropriate reference snould be cited and a (tapular) summary should be presented.

2. On-line fuel $vstem e spito'rino l The applicant's arr line fuel rod failure detection methods should be revisued.

Both the sensitivity of the instruments and the applicant's commitment to use the instruments sneuld be evaluated. Referenes 32 evaluates several common detection methods and should be utilized in this review.

Surveillance is also needed to assure that 54 C control rods are not losing reactivity. 8eren causennes are susceptible to teacning in the event of a cladding eefect. Periodic reactivity verth tasts such as described in Reference 33 are acceptaale. ._

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3. 8ost-irmdiation Surveillance 4 post-irrestation fuel survefltance progres should be described for each plant to detect anamelfes or confim aw.e4 fuel performance. The extent of an acceptante program will oogend on the history of the fuel design oeing l l considered, f.e., whether the orososed fuel assign is the same as current operating fuel or incorporatas new oesign features. l For a fuel oesign like that in other operating plants, a sinious acceptacle program snculd include a cualitative visual examination of some discharged l fuel assasolies free eacn refueling. Suca a program should be sufficient to icentify gross proolems of structural integrity, fuel and failure, rod bowing, j or c ud deposition. There should also se a commitment in tr.e progree to 4.2 11 Proposed Rev. 2 Oraft 1 l

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e perform aeditional surveillance if unusual benavior is noticed in the visual eseeinetton or if plant instruentation ineicates gross fuel failures. The surveillance program sneuld aderess the disposition of failed fuel.

In addition to the plant-soecific surveillance progree, there should entst a cwtinuing fuel surveillance effort for a given type, enke, or class of fuel that can be suitably referenced by all plants using steiler fuel. In the assence of such a generic program, the revisuer sneuld egoct eere detail in the plant-specific progree.

For a fuel design that introduces nor features, a more detailed surveillance progree esemensurate with the nature of the changes should be eescribed. This progree should inclues appropriate qualitative and quantitative inspections to be carried out at interie and one of-life refueling outages. This surveillance program sneuld be coortinated with prototype testing discussed in Suseection II-C-2. When prototype testing cannot ne performed, a special detailed surveillance program should be planned for the first irradiation of a new design.

III. REYTEW P40CEURES For construction permit (CP) asolications, the review should assure that the design Bases set forth in the preliminary safety analysis report (PSAR) seet the acceptance criteMa given in Supeection II-A. The CP review'should further astermine from a stusy of the pre 11einary fuel system design that there is reasoneele assurance that the final fuel systes design will meet the design bases. This jungsent say be based on expeM once with steilar designs.

For operating Ifconse (OL) applications, the review should confire that the design bases set forth in the final safety analysis report (FSAA) east the acceptance cMteria given in Sesection II-A and that the final fuel system design meets the design bases.

Much of the fuel system review is generic and is not moested for each steiler 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 unich the current review is deoensent should be referorced so that a coeletely cocumented safety evaluation is contained in the plant safety evaluation report. In j

particular, the NRC safety evaluation reports for all relevant licensing topical reports I

should be cited. Cartain geneMc mviews have also been perforced by CPB reviewers with findings issued as MURG- or WASH seMes reports. At the present time these reports include References 9,19, 31, 22, 34 and 35, and they should all be appropMately cited in the plant safety evaluation report. Applicante Regulatory Guides (Refs. 4, 20, 25-30) and Branen Tecnnical Positions (there are none at present) should also be mentioned fn the plant safety evaluation reports. Deviation free these guides or  ;

positions should be exclained. After triefly discussing related =revious reviews, the I l

7ecoosed Rev. 2 4.2-12 i Crsft i I 1

)

1

i plant safety evaluation should concentrate on areas where the application is not iden=

tical to previously reviamed and approved applications and areas related to newly discovered 3reeless.

Analytical pratictions discussed in Subsection II-C-3 will be revissed in PSARs, FSARs, or Itcenstag t. apical reports. idhen the methods are being reviewed, calculations by the staff eer be performed te verify the aseouacy of the analytical methods. Thorgefter, audit calculations will not usually be perfereed to chect the results of an appmved method that has been semitted in a safety anal.,11s report. Calculations, bencieserting exercises, and additional revious of generic eethods may be undertaken, heuever, at any time the clear need arises to reconfirm the adequacy of the eethod.

IV. EVALUATTON FTId3INES The reviewer should verify that sufficient inforestion has been provided to satisfy the requirements of this SRP Section and that the evaluetten semperts conclusions of the following type, to be included in the staff's safety evaluation report:

'The fuel system of the plant has been desfytod so that (a) the fuel systes will not be desaged as a result of normal operation and anticipated operational occurrences, (b) fuel desage during postulated accidents would not be seve=e enough to prevent control rod insertion unen it is required, and (c) core coolanility will always be maintained, even after severe postulated accidents.

"The applicant has provided sufficient evioence that these design oejectives will be est based on operating experience, prototype testing, and analytical predictions.

"The applicant has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not unserestimated.

• The asolicant has also 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 sede a coesiteent to perfore on-line fuel failure monitoring and post-irradiation surveillance to detect aneselfos or confire that the fuel has perfomed as expected.

"On the bests of our review of the fuel system design, we conclude that the applicant has met all tne requirements of the applicaele regulations, current regulatory post-tions, and good engineering practice."

v. REFEsENCES

1. 10 CFR Part 50. Accendix A General Design Criterion 10. "teactor Design."

2. " Rules for Construction of Muclear Power Plant casoonents," ASME Soiler and Pmssure vessel Cooe,Section III, 1977.

3. W. J. O'Donnel and S. F. Langer, " Fatigue Design desis for Zircaloy Conconents "

Mucl. Sci. Eng. {0, 1 (1964).

4. M 3 Prooosed 4ev. 2 3 raft *

4. Regulatory Guide 1.77, " Ass e sions used for Evaluat'ng a control Red Ejection Acciennt for Pressurized water Reactors."

5. *$taneerd Specification for Sintered Urania Dioxide Pellets " ASTM 5taneard C776-75, Part 45, 1977

5. K. Jeon. *Prisery Myeride Failure of Zircaloy-Clad Fuel Reds," Trans. As. Nucl.

Sec.jj,,145(1972).

7. R. N. Cheoman, %1 tired Buret Test Program Quartarly Progress Resort for April -

June 1977,* Oak Ridge National Laserstory Report OWN./NUREG/TM-135, Decommer 1977.

8. 10 CFR Part 50, Appendia K, "ECCS Evaluation Mosels."

9. "Tectmical Resort on Densification of Light Water Reactor Fuels," AEC Regulatory Staff Report WA5M-1236, Nevenner 14, 1972.

10. (Deleted)

11. F. Eroecner, *5fngle and hitired feats, Trenaient and Staesy State Internal Consuction Hosting ' Ff ftn NRC Water Reactor Safety Researcn Information Meeting, Gaithersourg, Maryland, Novemmer 11, 1977.

12. R. H. Chapman, "Some Preliminary Results of Single Red and Multired Tests With Internal Heaters," MRC Zircaloy Classing Review Grow Meeting, Silver Spring, Maryland, January 18, 1978.

13. C. E. Beyer, C. R. Henn, D. D. Lanning, F. E. Paniske and L. J. Perenen, Joer's Gutes for GAPCtNPTHEleut-2: A Casouter Program for Calculating the Theres1 Behavior of an 0mies Fuel Red," Sattelle Pacific Nortnuset Laboratory Resort BlefL-1897, movemeer 1975.

14 C. L Beyer. C. R. Hann, D. D. Lanning, F. E. Penisko and L. J. Perchen, "GAPCDekTHElemL-2: A Computer Program for Calculating the Thornal Benavior of an Qxies Fuel Red,' Battelle Pacific Nortinuset Lasoretory Report SNWI.-1898, Neusener 1975.

15. R. H. Staunt, D. T. Buchanan, 8. J. Buesener, L. L Losn,' N. W. Wfison and P. J. Henningson, "TACD - Fuel Pin Performance Analysis, Revision 1," Saccoct &

W11com Report SAW-10007A, Rev. 1. August 1977.

l l .

15. " Fuel Evaluation Model,' Comeuetion Engineering Rooert CENPD-139-A, July 1974 (Aceroved version transeitted to NRC April 25,1975).

17. 'Seolament 1 to the Tecnnical Resort on Densification of General Electric Reactor Fuels,* AEC Regulatory Staff Report Dae====c 14, 1973.

18. "Tecnnical Resort on Densification of Exxon Nuclear PWR Fuels," AEC Regulatory Staff Report, Fearwary 27, 1975.

19. R.O. Meyer, 'The Analysis of Fuel Densification," USNRC Resort NUREG-0085, July 1976.

20. Reguletary Guide 1.126, "An Accostable Mosel and Relates Statistical Metnoes for the Analysis of Fuel Densification."

21. Memorendum from V. Stallo, NRC, to R.C. Defoung, Suoject: Evaluation of Westing-house Report, WCAP-8377, Revised Clad Flattening Mooel, dated January 14, 1975.

22. Memoranous froe D. F. Ross, NRC. to R. C. DeYoung,

Subject:

CEPAN ~ Metnod of Analyzing Creeo Collapse of Oval Clacaing, cated Feoruary 5,1976.

23. Memorencum free D. F. Ross, NRC, to D. 5. vassallo, Sunject: Recuest for Revised Rod Bowing Tootcal Reports, dated May 30, 1978.

Proposed Rev. 2 4.2-14 Draft 1

24. Mesorenes from O. F. Ross and 3. G. Eisennut, NRC to D. 8. Vassallo and K. R. Galler, Seject: Revised Interie Safety Evaluation Report on the Effects of Fuel Rod Sering in Thereal Margin Calculations for Lignt Water Reactors, dated Feeruary 16, 1977.

25. (Noved te Appendix A)

26. Letter fue D. F. Ross, NAC, to A.E. Senerer, Connustion Engineering, dates Meren 22, 1978.

27. L. Baker and L. C. Just. " Studies of Metal-tdater Reactions at High Tasserstures, III. Emmerimental and Theoretical Studies of the Zirconia - Water Reaction,"

Argonne National Lamoratory Aeoert ANL-6548, May 1962.

28. Regulatory Gutes 1.3, "Assuntions used for Evaluating the Potential Radiological Cae==t- of a Loss of Coolant Accident for Betling Water Anectors."

29. Angulatory Guies 1.4, " Assumptions used for Evaluating the Potential Radielegical Conseeuences of a Loss of Coolant Accident for Pressurized idater Reesters."

30. Regulatory Guies 1.25, "Assu stions used for Evaluating the Potential Radielegical Consequences of a Fuel Handling Accident in the Fuel Handitng and Steroge Facility for Boiling and Pressurized Water Reactors."

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

32. 8. L Siegel and H. H. Hagen. " Fuel Failure Detection in Operating Reactors," USNRC Aegert MUREG-0401, Meren 1978.

33. " Safety Evaluation Neoort aelated te Operation of Artensas Muclear One, Unit 2,*

USNRC Report MUREG-0308, Sep. 2 (t,o be issues).

34. B. L. Siegel, " Evaluation of the Behavior of Waterlogged Fuel Red Failures in LidRs," USNRC Resort NUREG-0303, Merca 1978

35. R. O. Meyer, C. E. Beyer and J. C. Voglemose, " Fission Gas Release fras Fuel at High Burms," USNRC Report NUREG-0418. Meren 1978.

6 a.2*15 Prooosed Rev. 2 3 raft 1

U.S. Nuclear Regulatory Commission Office of Nuclear Reacter Regulation P2 POSED A00! TION OF APPENDIX A EVALUATION OF FUEL A$$8 GLY $TRUCTURAL RESPONSE TO EXTEfteALLY APPLIED PORCES TO STANDA W REVIEnt PLAN P5RP-4.2, REVISION 2. DRAFT 1 A. SActGROUn0 Eartneuenos and postulated pipe broens in the reactor coolant systes m uld result in external forces en the fuel assently. SRP Section 4.2 states tast fuel system coolentlity should se maintained and that demose sneuld not be se severe as to prevent control res insertion during these low prenan111ty accidents. This Aspendiz describes the review that should performed of the fuel assemoly structural response to seismic and LOCA loems. Background meterial for this Appeneix is given in Refs.1-3.

8. ANALYSIS OF LOADS

1. Input Input for the fuel essembly structural analysts comes from results of the primary coolant system structural analysis, which is revisued oy the %cnemical Engineering Sranca. Itout for the fuel assemely ressense to a LOCA should fnclues (a) motions of the core plate, core snegue, fuel altgruent plate, or other relevant structures; these motions should corrosoons to the breet that greeuced the peat fuel assessly loadings in the reactae primary coolant system analysts, saa (b) transtant pressure 61fferences that apply leans directly to the fuel assemely. If the earthoutaa loses are large enougn to preauce a non-11neer fuel assemely response, input for the seismic analysis sneuld use structure motions corresponsing to the reacter primary coelaat systes analysis for the 53E; f f a linoer response is presuced, a spectral analysis eey be used (see Regulatory Gutee 1.50).

This proposeo revision of sne Stansero deview Plan ene ene sunoort valvenmoec. statement have not receives a complete staff review and amorevel and do not reoresent an official NRC staff nosition. Public coments are seing solicited on both the revision and the value/imoect statement (including any f aelementation seneeules) orior to a review by the Regulatory Resutre*

ments Review Committee and their recommenettion as to unether this revision sneulo De aooroved.

Comments snould vesnington, D.C. be sent to the Secretary of the Camission, U.S. Muclear Regulatory Cosmission.

20$55 Attention: Docmeting and Service Sranen. All comments receivee my will ne consioered by the Regulatory Recuirements Review Casmaittee. 4 summary of sne seettng of the Committee at unich this revision is consicered. the Committee recomenoations ano a)) of the associatea oocuments anc comments considereo Dy the Committee will te saae nuolicly availacle prior to a cecision Dy the Director, Cffice of Muclear Reector seculation. on .netmee to %elemoet this aevision.

4.2*Al saccosed Rev. 2 Craft i

v

2. Matheos Analytical methoes used in performing structural response analyses aust be reviamed. Justification should be supplied to show that the neerical solution techniques are appropriate.

Lineer and non-lineer structural representations (f.e., the aseeling) must also be reviewed. Emperimental verification of the analytical roerosentation of the fuel assembly components should be provided when pratical.

A sample proclam of a stealified nature must be wereed oy the applicant and compared by the reviewer with either hand calculations or results generated by the reviewerwithaninessendentcosa(2). Although the samp;a proeles should use a structural representation that is as close as possible to the dosip in question (and, therefore, would very from one veneer to another), stupitfying assmettons any be seen (e.g., one signt use a 3-assembly core region with continuous sinuseisel input).

The sample proelen should be assigned to exercise various features of the code and reveal their benevior. The sample proeles comparison is not, however, designed to show that one coes is more conservative than another, but rather to alert the reviewer to major discrepancies so that an explanation can t>e sought.

3. Uncertainty 411omences The fuel assemoly structural models and analytical methods are proosely conservative and input parameters are also conservative. Mounger, to ensure that the fuel assemely analysis does nct introduce any non-conservatises, two precautions should be tamen: (a) If it is not explicitly evaluated, impact loads free the PWt LOCA analysis should be increased (by aneut 305) to account for a prwssure pulsa, which is associated with steem flashing that affects only the PWt fuel assembly analysis. (b) Conservative margin should be added if any part of the analysis (PWI or SWR) exhibits pronounced sensitivity to input variations.

Variations in resultant loads should be determined for 105 variations in input amplitude and frequency; variations in amplitues and frequency should be ases separately, not simultaneously. A factor should be developed for resultant load eagnituos variatioss of more than 155. For enamole, if :105 variations in input magnituos or frequency proeuce a eexism resultant increase of 255, the sensitivity factor would be 1.2. Since resonances and pronounced sensitivities any be plant-dependent, the sensitivity analysis should be perferised on a plant-by-plant tasis until the reviewer is confident that further sensitivity analyses are unnecessary.

4. Aucit Inceoencent audit calculations for a typical full-sited core must be performed oy the reviewer to verify that the overall structural representation is acecuate. An 8P000$0C tev. 2 3 raft 1 4.2-42

v

?

ineeseneont audit code ((} nould be used for this audit during the generic review of the analytical methoes.

5. Comeinstion of Loses Generel Deafgn Criterien 2 requires an aeoropriate conninetton of loses from natural pnenemone and accident conettions. Loans on fuel assemely camponents should be calculated for each input (f.e., seismic and LOCA) as anscribed aeove in Peregreen 1, and the resulting loses sneuld be aseed by the square *reet-of*sur of*souares (5R15) aothed. These cousined loses sneuld be comoeroe with the component strengths oescribes in Section.C occoreing to the acceptance criteria in Section 3.

C. DETEMhaTTON OF TT1tENGTH L GMa .

All modes of loading (e.g., in grid and through-grid loadings) should be considered, ano the most damaging mede should be representee in the veneor's Iaoeretory grid strength tests. Test proceeures and results snould be reviewed to assure that the approoMate failure sees is being predicted. The review snould also confire that (a) the testing imoect velocities correspone to expected fuel assemely velocities, and (b) the crushing load P has been suitaely selected from the lead- e erit oeflection curves. e== of the potential for different test rigs to introeuce onesurement variations, an oveluation of the grid strength test equipment will se incluoed as part of the review of the test proceeure.

The consequences of gMc deforeetion are saali. Gross deforestion of grias in many PWR assemelfes would be needed to interfere with control red insertion during an

$5E (f.e., buckling of a few isolated grids could not af splace guide tunes signifi-cantly from their proper location), and gMa deformation (without enannel oeflection; muld not affect control blade insertion in a BWR. In a LOCA, gross deforestion of the not enannel in either a PWR or a SWR would result in only small increases in poet classing temperature. Therefore, average values are appreoMate, and the I

alloweele crusning load Pg ,9g should be the M confidence level on the true seen as taaen from the distribution of eessurements on unirreefsted production gMds at (or corrected to) operating temperature. While P ggg will increase with irradiation.

auctility will be roeuced. The extra mergin te Pegg for irrsetated grids is thus assmee to offset the unknown esformation benewfor of irresisted gMas beyond

'cM t-

2. Components Other than Grids

$trengtns of fuel assemely components other than spacer gries eey be deduced from funoamental procerties or experimentation. Succorting evioence for strengtn values snould te suoplise. $1nce structural failure of these coaconents (e.g. , fracturing i

[ of gutes tunes or fragmentation of fuel roas) could be more serious than grid l

l .

4.2*43 Precosed 8ev. !

Oraft 1 l

l

e

./

deforestion, alloweele values should bound a large percentage (anout 95I) of the distribution of component strengths. Therefore, ASME Seiler and Pressure vessel Coos values and procedums any be used unere appropriate for determining yield and

__ __ _ _ ultioste_ strengths Mfication of alloweele values any follow the ASME Code requirements and should include consideration of buckling and fatigue effects.

D. ACCEPfamCE CRITERIA

1. Loss-of-Coelant Accident Two principal criteria apply fir the LOCA: (a) fuel rod fragmentation must not occur as a direct result of the blowenwn loads, and (b) the 10 CFR 50.46 temperature and oxidation Itaits mast not be encoeded. The first criterion is satisfied if the cousined leads on the fuel rods and components other than grids reesin below the alloweele values defined above. The second criterion is satisfied by an ECCS analysis. If conoined leads on the grids remain below P,74g, as defined aeove, then no significant distortion of the fuel assemoly would occur and the usual ECCS analysis is sufficient. If connined grid loads exceed P,79g, then grid oeforestion must be assumed and th6 ECC5 analysis must incluse the effects of distorted fuel assanelies. An assm otion of menta m credible deformation (i.e.,

fully collapsed gries) may be ende unless other asswetions are justified.

Control rod insertacility is a third criterion that must be satisfied for the LOCAs that require insertion to assure suscriticality. Loads free the aest severe LOCA that requires control rod insertion must be comeined with the SSE loads, and control rod insertability must be demonstrated for that connined load. For a PWR, if connined loads on the grids remain below Pg79g as donned aeove, then sign 1Neant deforestion of the fuel assemely would not occur and control FM insertion would not be interfered with by lateral displacement of the guide tunes. If connined loads on the grids exceed Pg79g, then M tional analysis is needed to show that deformation is not severe enough to prevent control rod insertion.

For a Mt. several conditions must be est to demonstrate control rod insertanility:

(a) connined loads on the channel box aust remain below the alloweele value defined aeove for components other than grids because a saali amount of channel deformation could interfere with control alade insertion, and (b) vertical liftoff forces must not unseat the lower tieplate from the fuel support piece Decause the resulting loss of lateral fuel mundle positioning could also interfere with control blade insertion.

2. Safe shutdown Earthouane Two critaria apoly for the SSE: (a) fuel rod fragmentation sust not occur as a result of the seisste loads, and (b) control rod insertanility sust se assured.

The first criterion is satisfied my the criteria in Paragraon

• The second criterion sust ce satisfied for SSE toads alone if no snalysis for casoined loads is required by Psragraan 1.

prooosed Rev. 2 3 raft ; 4.2 44

E. REFERENCE _S

1. R. L. Gruse, " Review of LW Fuel Systee Macnanical Response with Receemeneations for het Acceptance Criteria," Idene National Engineering Laserstory, 8tlREG/CR-1018, Septaneer 1979.

2. R. L. Gree, " Pressurized hter Reacter Lateral Core Rossense Routine, FAMEC (Fuel Assaeoly Machenical Response Code)," Idene National EngineeMag Lateratory, NUREG/CR-1019, Septasse 1979.

3. R. L. Gree, " Technical Evaluation of PWt Fuel Spacer Grid Response Lead Sensitivity Studies," Idene National Engineering Lateratory, MUREG/CR-1020, Septmoser 1979.

4.2.A5 PMOo M Rev. 2 3rstt 1

_ _ _ . . . _ _ _ _ _ . _ _ _ _ _ _ _ _ . . . . . .