ML20138A124

Topical Rept Evaluation of BAW-10153P, Extended Burnup Evaluation. Rept Acceptable for Ref in License Applications

ML20138A124
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Issue date:	09/30/1985
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NUDOCS 8512110571
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.i SAFETY EVALUATION REPORT OF A

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l BABC0CK & WILCOX

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! LICENSING TOPICAL REPORT

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, " EXTENDED BURNUP EVALUATION," BAW-10153P, SEPTEMBER 1982 i

i Prep.ared By l; Core Performance Branch and Accident Evaluation Branch

!k Division of Systems Integration

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Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Connission f

B512110571 851203 PDR TOPRP EMVB C

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! TABLE OF CONTENTS

1.0 INTRODUCTION

j 2.0 FUEL SYSTEM DAMAGE j a Design Stress i 4

b Cladding Design Strain i c Strain Fatigue d Fretting Wear

!- e Oxidation and Crud Buildup f Rod Bowing Axial Growth Rod Pressure Assembly Liftoff Control Material Leaching 3.0 FUEL R0D FAILURE e a Hydriding i; b Cladding Collapse I c Overheating of Cladding i d Overheating of Fuel Pellets (e Pellet / Cladding Interaction (f Cladding Rupture (g Fuel Rod Mechanical Fracturing 4.0 FUEL C00 LABILITY (a Fragmentation of Embrittled Cladding (b Violent Expulsion of Fuel (c Cladding Ballooning (d Fuel Assembly Structural Damage from External Forces t 5.0 NUCLEAR DESIGN

. 6.0 RADIOLOGICAL CONSIDERATIONS OF POSTULATED ACCIDENTS WITH EXTENDED BURNUP OPERATION

7.0 CONCLUSION

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8.0 REFERENCES

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1.0 INTRODUCTION

I Economics and prudent utilization of resources has led utilities to seek more f'

efficient use of current generation light water reactors (LWRs). Improved fuel utilization is one of the avenues being pursued for greater efficiency. One of the greater improvements in fuel utilization is to increase the fuel discharge j exposure which is currently at batch average burnups of approximately 28 mwd /kgM for BWRs and approximately 33 mwd /kgM for PWRs to batch average burnups of approximately 40 mwd /kgM and 50 mwd /kgM or above, respectively. The higher l

I discharge exposures result in lower fuel fabrication costs per cycle and a

[ better utilization of the plutonium produced in-reactor. The longer residence I time in reactor also reduces spent fuel storage needs.

I f In response to this trend for extended burnup fuel operation, the Nuclear Regu-

{ latory Commission (NRC) requested (Ref.1) each fuel vendor to prepare and sub-

! mit a topical report for review and approval that covers extended burnup

,I experience, methods and test data to provide a generic basis for operation at f extended burnups.

, Babcock & Wilcox (B&W) submitted such a report (Ref. 2) requesting generic li-censing approval of B&W-designed Mark B fuel assemblies for batch average burnups up to a proprietary value specified in their report (Ref. 2). This review also covers tha criteria and analysis methods used to show that the Mark B fuel design is capable of obtaining the extended burnup level requested. B&W hasalsoprovidedresponses(Refs.3,4,5)toNRCquestionsconcerningthis submittal.

f B&W has indicated (Ref. 2) that they plan to extend the burnup levels of their fuel designs to higher levels than requested in this submittal (Ref. 2), the

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, level of which is contingent on the feedback from demonstration and test assem-

, bly programs. This may be accomplished by an addendum to this report or by submittal of a new report.

, This technical review and evaluation has been perfonned by Pacific Northwest

! Laboratory (PNL) under contract (FIN B2533) with the United States NRC. The review has been based on References 2 through 5 and Section 4.2 of the Standard l

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i f Review Plan (SRP) (Ref. 6) and covers the fuel assembly, fuel rods, and burn- )

I able poison rods but does not include the control element assembly for extended l i burnup operation.

ibis report follows the intent of Section 4.2 of the SRP, where appropriate for j a generic review, to insure that all licer. sing requirements of the fuel system are reviewed with respect to extended burnup operation. The objective of Sec-

! tion 4.2 and this review are to provide assurance that as a result of extended

! burnup operation (a) the system is not damaged as a result of normal operation f

and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number

[ of fuel rod failures is not underestimated for postulated accidents, and (d)

,[ coolability is always maintained. "Not damaged" is defined as meaning that l fuel rods do not fail, that fuel system dimensions remain within operational tolerances, and that functional capabilities are not reduced below those as-sumed in the safety analysis. This objective implements General Design Criterion (GDC) 10 of 10 CFR Part 50, Appendix A (" General Design Criteria for Nuclear

, Power Plants") 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, there-7 fore, been breached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (" Reactor Site Criteria") for postulated accidents.

"Coolability", which is sometimes tenned "coolable geometry", means, in general, that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channels to pennit removal of residual heat after a severe accident. The general requirements to maintain control rod insertability and

! core coolability appear repeatedly in the General Design Criteria (e.g., GDC t

i 27and35). Specific coolability requirements for the loss-of-coolant accidents are given in 10 CFR Part 50.46 (" Acceptance Criteria for Emergency Core Cooling i

Systems for Light Water Nuclear Power Reactors").

i In order to meet the above stated objectives and follow the fonnat of Section 4.2, this review covers the following three categories: (1) Fuel System Damage Mechanisms, which are most applicable to normal operation and anticipated oper-ational occurrences; (2) Fuel Rod Failure Mechanisms, which apply to normal l

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! operation, anticipated operational occurrences, and postulated accidents; and

[ (3) Fuel Coolability, which applies to postulated accidents, t

In addition to Section 4.2 of the SRP, we have reviewed those aspects of Sec-l tion 4.3, " Nuclear Design" which are affected by high burnup and these are dis-

'I cussed in general terms in Section 5.0. A brief discussion of radiological consequences of operation with high burnup fuel is given in Section 6.0 and our ,

! conclusions and Regulatory Position are given in Section 7.0.

I f As noted earlier, this review is intended to provide generic approval of the criteria and methods used by B&W for licensing of their fuel designs to the i extended burnup level requested (Ref. 2). In addition, this review is also l intended to provide generic approval of their Mark B fuel design to the burnup level requested. This approval does not mean that the Mark B fuel design is approved for extended burnups in a specific reactor because there are reactor

specific analyses that need to be addressed, e.g., seismic and LOCA loads analyses and internal rod pressure analyses. This can be accomplished by a reactor specific analysis or by referencing a previous generic analysis that bounds the

! reactor specific case in question, i

The criteria sections in this review address limiting values for fuel damage that are acceptable under the three major categories of failure mechanisms listed above and in the SRP. The purpose of this review is to determine if the B&W criteria are applicable to extended burnup operation of their fuel. These criteria along with certain definitions for fuel failure constitute the SAFDLs required by GDC 10.

s The evaluation sections review the methods that B&W uses to demonstrate that

the design criteria have been met for extended burnup operation and thus are reviewed with respect to their applicability to the proposed range of extended burnup operation. These methods and data may include operating experience, prototype testing and analytical techniques. This section also reviews the

$ generic analyses performed by B&W for the Mark B fuel design at the extended burnup level requested (Ref. 2).

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} 2.0 FUEL SYSTEM DAMAGE t

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The design criteria in this section should not be exceeded during nomal opera- i

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tion including anticipated operational occurrences (A00s). The evaluation por- l

! tion of each damage mechanism demonstrates that the design criteria are not I exceeded during normal operation and A00s for the Mark B design at the extended j burnup level requested.

l (a)DesignStress

't Bases / Criteria - The B&W design criteria for fuel assembly components and fuel j rod cladding stresses are presented in Table 2.2.3-1 and Section 2.3.1.1 of Reference 2, respectively.

The stress limits for the fuel assembly structural components in Table 2.2.3-1 have been reviewed and found to meet Section III of the ASME Boiler and Pres-sure Vessel Code (Ref. 7). These limits are consistent with those in Section 4.2 of the SRP and are therefore found acceptable for extended burnup application.

The design criterion for fuel rod stresses is that all stresses not exceed the j minimum unirradiated yield strength of the Zircaloy-4 cladding. This criterion is judged to be conservative for extended burnup applications because Zircaloy-4 yield strength has been shown to increase with increased irradiation l

exposure and thus is substantially higher at extended burnups than the unirradiated value. Consequently, this criterion is approved for extended I

burnup application.

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Evaluation - B&W has used the TACO-2 code (Ref. 8) and Section III of the ASME code to show that Mark B fuel rod and assembly components, respectively, for a generic reactor, meet the above criteria with adequate margin at the extended

, burnup level requested. The TACO-2 code takes into account those parameters

important for determining cladding stresses and strains at extended burnups, such as pellet thermal expansion and swelling, cladding creep and fuel rod /

coolant system pressure differences. As a result the NRC has previously ap-proved (Ref. 9) the use of this code for licensing applications without a specific burnup restriction. As a result of this review of B&W extended 4

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'[ burnup methods, we find TACO-2 to be acceptable to the burnup limit requested

<{ by B&W in Reference 2. Consequently, these generic analyses are found acceptable j for the Mark B fuel design up to the burnup level requested.

! 1 (b) Cladding Design Strain l

l Bases / Criteria - The B&W design criterion for fuel rod cladding strain is that j the maximum hoop strain shall not exceed 1%. This criterion is intended to

-j preclude excessive cladding deformation from anticipated operational tran-

[ sients. This is the same criterion that is used in Section 4.2 of the SRP. l

~h The material property that could have a significant impact on the cladding strain criterion at extended burnup levels is cladding ductility. The strain

] criterion could be impacted if cladding ductility were decreased, as a result

< of extended burnup operations, to a level that would allow cladding fatiure

without the 1% cladding strain criterion being exceeded in the B&W analyses.

i, From examination of irradiated Zircaloy cladding ductility data (Refs. 10,11),

it has been concluded that ductility decreases with increasing fluence at low f

.j burnup levels, i.e., < 12 mwd /kgM, but asyntotically approaches a small fluence i dependence beyond these low burnups. Consequently, cladding ductility shows little change for the increased burnup levels projected. In addition, B&W has ,

irradiated experimental and lead test assemblies with average burnups up to 50 mwd /kgM with no adverse effects in cladding ductility (Ref. 2).

From the above, we can conclude that the strain limits proposed by B&W are ap-plicable for extended burnup application to the burnup requested.

ll Evaluation - The NRC-approved B&W fuel performance code, TACO-2 (Ref. 8), has j been used to show that the Mark B fuel design generically meets the above cri-terion with adequate margin. As noted in the Design Stress section, this code takes into account those parameters important for determining cladding stresses and strains at extended burnups and is approved to the burnup level requested by B&W in Reference 2. Consequently, the generic analyses for the Mark B design are found acceptable up to the burnup level requested.

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(c) Strain Fatigue Bases / Criteria - The B&W design criterion for fuel rod cladding fatigue is that the cumulative fatigue usage factor be less than 0.9 when a minimum safety fac-f ter of 2 on the stress amplitude or a minimum safety factor of 20 on the number j i of cycles, whichever is the more conservative, is imposed. This criterion is l

' somewhat more conservative than that described in Section 4 '; of the SRP be-cause the B&W criterion requires a fatigue usage factor te be less than 0.9 l while the SRP criterion requires the fatigue usage factor to be less than 1.0. l As noted for cladding strain, the material property that could have a signifi-

'~ cant effect on the strain fatigue criterion is cladding ductility. As dis-1 cussed in the above section on design strain, extended burnup operation has shown little observable effects on cladding ductility. From this it is con-I cluded that extended burnup operation does not reduce the applicability of the strain fatigue limits, which are therefore found acceptable for use in extended i burnup applications. l

Evaluation - The B&W methodology for determining strain fatigue is based

,[ on the procedures outlined in the ASME Boiler and Pressure Vessel Code (Ref. 7) lf l

along with the O'Donnell and Langer design curve (Ref. 12) for fatigue usage.

This methodology is the same as that suggested in Section 4.2 of the SRP. This methodology also takes into account daily load follow operation and the addi-

, tional fatigue load cycles that may result from extended burnup operation.

Therefore, the above methodology is found to model operational and material i

behavior parameters important for determining strain fatigue at extended burnups and thus are acceptable for extended burnup application. '

(d) Fretting Wear Bases / Criteria - Fretting wear is a concern for fuel and burnable poison rods,

, and the fuel assembly guide tubes. Fretting, or wear, may occur on the fuel and/or burnable rod cladding surfaces in contact with the spacer grids if there is a reduction in grid spacing loads in combination with small amplitude,

! flow-induced, vibratory forces. Guide tube wear may result when there is flow l

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h induced motion between the control rod ends and the inner wall of the guide tube.

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While the Standard Review Plan (SRP), Section 4.2 (Ref. 6), does not provide i numerical bounding-value acceptance criteria for fretting wear, it does stipu-i late that the allowable fretting wear should be stated in the safety analysis report and that the stress / strain and fatigue limits should presume the exis-tence of this wear.

Evaluation - B&W has indicated (Refs. 2, 3, and 4) that they do not believe

. fretting wear is significant for their fuel designs because of the lack of observable fretting wear from both in-reactor and out-of-reactor tests on fuel

[ assemblies.

I j In order to support the fact that B&W does not reduce cladding thickness due to

! fretting, B&W has indicated that examination of B&W Inconel grid assemb' lies

[ with burnups up to a proprietary number have shown no significant wear and the i examination of Zircaloy grid assemblies after one cycle of irradiation has

. shown no grid wear. As further support B&W has indicated that the effects of small amounts of fretting wear on the cladding are insignificant particularly f when compared to the large margin of conservatism used in these analyses such

} as minimum unirradiated yield stress. We concur with this assessment; however, f further examinations should be performed on the Zircaloy grid assemblies up to the extended burnup levels requested in order to confinn the lack of fretting wear on Mark B assemblies.

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The effects of cladding wear and thus cladding thinning on the LOCA analysis has been addressed by B&W (Refs. 3,4) and found to be less limiting than cur-i rent analysis methods. We concur with this assessment.

1 l B&W analyses of unworn guide tubes have shown very large margins which B&W has l'

indicated (Ref. 3) more than bounds the effects of small amounts of fretting wear at extended burnups. In order to support this claim B&W has provided wear measurement data from cold worked guide tubes that show essentially no wear for up to two cycles of irradiation. B&W has further indicated that control rods will not be parked in the same position for four cycles of irradiation.

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? Out-of-reactor flow tests of annealed guide tubes have shown no increased wear I

over cold worked tubes. However, since annealed tubes are to be used for fu-ture Mark B assemblies at extended burnups, B&W intends to make in-reactor wear measurements on lead test assemblies with annealed guide tubes. From this it ,

is concluded that guide tube wear does not appear to be a problem for the Mark B assemblies; however, because there is a change in guide tube material for

. Mark B assemblies confirmation data should be provided from lead test and/or experimental assemblies near the extended burnups requested.

(e) Oxidation and Crud Buildup j Bases / Criteria - Section 4.2 of the Standard Review Plan identifies cladding

oxidation and crud buildup as potential fuel system damage mechanisms. General

! mechanical properties of the cladding are not significantly impacted by thin t

oxides or crud buildup. The major means of controlling fuel damage due to cladding oxidation and crud is through water chemistry controls, materials used in the primary system, and fuel surveillance programs that are all reactor spe-

cific. Because these controls are already included in the specific reactor design, a design limit on cladding oxidation and crud is judged to be redundant and thus is not necessary.

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This does not, however, eliminate the need to include the effects of cladding oxidation and crud in safety analyses such as for LOCA and mechanica! analyses.

This will be discussed in further detail in the evaluation presented below.

l Evaluation - B&W has indicated (Refs. 3,4) that the effects of cladding oxida-I tion and crud are not explicitly included in their thermal analyses but that

[ they are implicitly considered in their steady-state fuel performance code, l TACO-2(Ref.8). However, they have indicated that the ef b t of cladding thinning due to cladding oxidation is explicitly accounted for in their mechan-ical analyses. \

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There is a normal amount of oxidation expected during the irradiation of fuel rods that is dependent on time in reactor. Therefore, extended burnups will result in thicker oxide layers that provide an extra thermal barrier affecting thermal analyses and cladding thinning that can affect the mechanical analyses.

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P B&W has responded (Refs. 3,4) that the most limiting LOCA analysis is early in life when thermal stored energy is highest and oxidation is insignificant. We concur with this assessment. For other themally dependent analyses, e.g.,

[ fuel melting and internal rod pressures at extended burnup, our consultants {

have confi med by independent analyses that there are other conservatisms such  !

I as peak linear heat generation rate and code conservatisms that are more than l an order of magnitude larger than the themal effect of the oxide layer )

': (approximately 25*C) at extended burnups. Consequently, these conservatisms  ;

more than compensate for the fact that cladding oxidation and crud are not l i explicitly accounted for in these themal analyses.

B&W has indicated that a reduction in cladding thickness (the value of which is proprietary) due to cladding oxidation is accounted for in their mechanical L

analyses. Close examination of this value indicates that it is close to the ij mean of the oxide thickness data at the burnup level requested but that this 3

is nearly a factor of 2 below the upper bounds of these data. In response to a question B&W has indicated that an upper bound value of cladding thinning due to oxidation is not necessary because of large conservative margins that exist in these analyses such as minimum unirradiated yield stress. As a check i on the effects of upper bound oxidation on the stress analysis the reduction

! in cladding thickness used by B&W was doubled and the change in cladding j stresses was found to be less than 4%. This is judged to be insignificant in comparison to the conservatisms that exist in this analysis.

As stated above, B&W has further indicated that cladding thickness is not reduced due to fretting or oxidation in the LOCA cladding rupture analysis [see Section 3.0(f) p of this report] because their analyses have shown that beginning-of-life condi-

I tions are the most limiting. In addition they have indicated that including iI cladding thinning effects in the LOCA analysis of extended burnup fuel results in lower cladding rupture strains which are less conservative than those obtained from the current methods.

f From the above, it is concluded that the effects of cladding oxidation on ther-mal analyses are insignificant up to the extended burnup level requested and )

thus need not be incorporated in these analyses. It is further concluded that the B&W method of reducing cladding thickness due to oxidation is adequate up to the extended burnup level requested.

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t (f)RodBowing

,g Bases / Criteria - Fuel and burnable poison rod bowing are phenomena that alter the design pitch dimension between adjacent rods. Bowing affects local nucle-i ar power peaking and the local heat transfer to the coolant. Rather than plac-

[ ing design limits on the amount of bowing that is pemitted, the effects of

.* bowing are included in the safety analysis. This is consistent with the Stan-dard Review Plan and we have approved (Ref.13) this for current burnup levels, j and this is also judged to be acceptable for extended burnups. The methods used for predicting the degree of rod bowing at extended burnups are evaluated

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I Evaluation - The methods used to account for the effect of fuel rod bowing in i

[ B&W assemblies have been addressed in Reference 14 which has been approved (Ref. 13) for current burnup levels. B&W has indicated that they intend to f

I apply these methods and models to extended burnup assemblies. In order to sup-port this they have presented (Refs. 2,3) extended burnup rod bow data that

. l agree with the approved model. These data also show that rod bow may saturate at extended burnups; however, these data are judged to be limited with large un-

! certainties and thus cannot be used to justify rod bow saturation at this time.

It should be noted that extended burnup assemblies will in general not be lim-iting in terms of DNBR because the power peaking in extended burnup assemblies il is dropping faster than the power penalties applied to account for rod bow.

However, although it is concluded that the licensing methodology approved (Ref. 13) previously for current burnups is also applicable for extended burnup assemblies, this does not eliminate the need to confirm whether or not specific licensing applications for extended burnup operation are DNBR limited.

[ (g) Axial Growth i

Bases / Criteria - The core components requiring axial-dimensional evaluation are l the control rods, burnable poison rods, fuel rods, and fuel assemblies. The l control rods are not included in this extended burnup review. The growth of l the second two is mainly governed by (a) the behavior of poison and fuel

! pellets, and their interaction with the Zircaloy-4 cladding, and (b) the irradiation-and-stress (due to rod pressures being less than coolant pressure)

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induced growth of the Zircaloy-4 cladding. The growth of the latter is a func-

tion of both the compressive creep and the irradiation-induced growth of the

! Zircaloy-4 guide tubes. For the Zircaloy cladding and fuel assembly guide

, tubes, the critical tolerances that require controlling are (a) the clearance between the fuel rod and the fuel assembly upper end fitting, and (b) clearance _

between the fuel assemblies and the core internals. Failure to adequately de-I sign for the former may result in fuel rod bowing, and for the latter may re-

[ sult in collapse of the assembly holddown springs.

B&W has indicated that the design basis for both of these clearances is that no

axial interference is allowed. This design basis is found to be acceptable for

! extended burnup application.

! Evaluation - B&W has stated (Ref. 2) that Mark B assemblies intended for ex-tended burnup applications will have annealed guide tubes which exhibit lower axial irradiation growth rates than the cold worked and stress relieved tubes

'! currently being used. B&W is currently irradiating Mark B lead test assemblies I

to confinn the lower irradiation growth rates of the annealed tubes.

l t B&W has axial irradiation growth data for their fuel rods up to rod average

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j burnups of approximately 40 mwd /kgM an'd has indicated that lead test assembly data will extend the rod average burnups to 60 mwd /kgM.

g From the above, it is concluded that rod and assembly axial growth data need to be obtained up to the extended burnup level requested in this submittal (Ref.

j [ 2). In the meantime, this issue should be addressed in design change specific j submittals for Mark B assemblies with annealed guide tubes.

f (h)RodPressure Bases / Criteria - Rod internal pressure is a driving force for, rather than a direct mechanism of, fuel system damage, that could contribute to the loss of dimensional stability and cladding integrity. The Standard Review Plan presents an acceptance criterion that is sufficient to preclude fuel damage in this regard and has been widely used by the industry; it states that rod inter-nal gas pressure should remain below the nominal system pressure during normal l

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operation unless otherwise justified. B&W has endorsed (Ref. 2) this criterion

[ and it is judged to be conservative with respect to extended burnup operation, j Therefore, this criterion is acceptable for extended burnup application.

Evaluation - The models and methods used by B&W to evaluate whether their de-l signs meet the above criterion are examined in this section. The models used aredescribedintheTACO-2codedocumentation(Ref.8)whichhasbeenpreviously approved by the NRC. The review of this code paid particular attention to those j parameters important to internal rod pressure predictions, i.e., the thermal and fission gas release models. Therefore, the TACO-2 code is approved for use in the evaluation of rod internal pressures of extended burnup fuel.

f An important parameter of the methodology used in the internal rod pressure l evaluations is the power history used as input to the TACO-2 code. Power his-i tory is very important because fission gas release and thus internal rod pres-f sures are strongly dependent on the fuel thermal history. In response to a question on the power histories used, B&W has indicated (Ref. 5) that the power histories input to the TACO-2 code for calculating internal rod pressures are based on a generic power history envelope with appropriate conservatisms for nuclear uncertainty and potential operational transients which bound all ex-

[ pected rod power histories. The predicted power history of each fuel rod de-termined from the fuel cycle design process, with the uncertainty and transient allowances added, is compared to the generic envelope. If any individual rod power level exceeds the envelope or if the predicted fuel rod burnup is higher than that which is justified by the generic analysis, then that individual rod history is evaluated with respect to the above criterion. This methodology for evaluating internal rod pressures is judged to be bounding and thus acceptable

, for extended burnup application.

(i)AssemblyLiftoff Bases / Criteria - The Standard Review Plan calls for the fuel assembly holddown capability (wet weight and spring forces) to exceed worst-case hydraulic loads for normal operation, which includes anticipated operational occurrences. B&W 12

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I i endorses this design basis. This is found acceptable for application to extended burnup assemblies.

I Evaluation - The above design basis is addressed by B&W by a bounding calcula-l tion using worst case LOCA loads. The B&W methods used for calculating hydrau-l lic loads during a LOCA are addressed in Reference 15. The fuel assembly liftoff forces are a function of plant hydraulic forces, spring forces, and

assembly dimensional changes. Extended burnup irradiation will result in addi-E tional holddown spring relaxation and assembly length increases which will have

'I opposing effects on assembly holddown forces.

4 B&W has indicated (Ref. 2) that for their Mark B extended burnup assemblies j they will use annealed guioe tubes which will result in less axial assembly j growth. In order to compensate for this in assembly holddown forces they have i redesigned their holddown springs to provide more holddown force at extended burnups. The effects of these design changes on assembly growth and holddown l

l spring relaxation will be verified (Refs. 2,3) by B&W in # e examination of lead I test assemblies, currently under irradiation. From this it appears that B&W has

[ employed adequate design changes in their Mark B assemblies to maintain assem-lj bly holddown capability at extended burnups; however, this needs to be con-l firmed by lead test assembly data up to the burnup level requested in the B&W submittal (Ref. 2). In the meantime, this issue should be addressed in design specific submittals-for Mark B assemblies with the annealed guide tubes and redesigned holddown springs.

) (j) Control Material Leaching I

'l Bases / Criteria - The Standard Review Plan and General Design Criteria require l

that reactivity control be maintained. Rod reactivity can sometimes be lost by leaching of certain poison materials if the c1 4 ding of control-bearing materi-s al has been breached.

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Evaluation - The control rods are not within the scope of this review and will t not be discussed further. Reactivity loss from burnable poison rods at extend-ed burnup levels is found to be insignificant because nearly all of the reac-

!' tivity controlling boron-10 is burned out at these burnup levels.

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h l f Consequently, reactivity loss due to leaching of burnable poison rods at ex-tended burnups is not judged to be significant. No further evaluations are needed in this area, with respect to extended burnup.

3.0 FUEL R00 FAILURE 3 j In the following paragraphs, fuel rod failure thresholds and analysis methods

! for the failure mechanisms listed in the Standard Review Plan are reviewed.

j When the failure thresholds are applied to normal operation including antici-pated operational occurrences, they are used as limits (and hence SAFDLs) since j fuel failure under those conditions should not occur according to the tradi-l tional conservative interpretation of General Design Criterion 10. When these thresholds are used for postulated accidents, fuel. failures are pennitted, but they must be accounted for in the dose calculations required by 10 CFR 100.

The basis or reason for establishing these failure thresholds is thus estab-i lished by GDC 10 and Part 100 and only the threshold values and the analysis methods used to assure that they are met are reviewed below.

1 (a)Hydriding f

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Bases / Criteria - Internal hydriding as a cladding failure mechanism is preclud-ed by controlling the level of hydrogen impurities in the fuel pellets during fabrication and is an early-in-life failure mechanism. The moisture level of B&W fuel pellets is controlled by drying the pellets in the cladding and taking a statistical sample to ensure that the moisture level is below a particular (proprietary) level. This level is not stated in this submittal but previous B&W plant reviews have indicated that this value is below the value recommended

,l in the SRP. Consequently, this is found to be acceptable.

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. Evaluation - B&W has presented cladding hydrogen content measurements from fuel rods with burnups up to approximately 50 mwd /kgM. This hydrogen pickup is a function of irradiation time and follows the oxide thickness data closely be-cause it is introduced by external cladding oxidation, see Section 2.0(e).

Examination of these data shows that considerable margin exists between the amount of hydrogen that will result in cladding embrittlement and the 14

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f p hydrogen levels anticipated at the extended burnup requested. In addition, the ll Mark B fuel design has not shown any evidence of primary hydride failures in the past. From this, it is concluded that there is reasonable evidence that primary hydriding will not be a likely failure mechanism at the extended burnup requested.

!I (b)CladdingCollapse

.j Bases / Criteria - If axial gaps in the fuel pellet column were to occur due to l

[ densification, the cladding would have the potential of collapsing into a gap (i.e., flattening). Because of the large local strains that would result from

j j collapse, the cladding is assumed to fail. It is a B&W design basis that clad-k ding collapse is precluded during the fuel rod and burnable poison rod design r lifetime. This design basis is the same as that in the Standard Review Plan and is also found to be conservative and thus acceptable for extended burnup applications.

Evaluation - The longer in-reactor residence times associated with extended-l[ burnup fuel will increase the amount of creep of an unsupported fuel cladding.

4 Extensive postirradiation examinations by B&W and other fuel vendors have not shown any evidence of cladding collapse or large local ovalities in current fuel designs. This is primarily the result of their use of prepressurized rods and stable fuel in current generation designs.

B&W has indicated (Ref. 2) that the Mark B fuel rod design will not experience cladding collapse at burnur even greater than the extended burnup level requested in this submittal using the approved analysis methods described in References 8 and 16. These analysis methods and are very conservative for the stable fuel designs currently being employed by B&W because they assume an axial gap has fomed in the fuel column and the tube is unsupported. This methodology accounts for the longer in-reactor residence times associated with extended burnup fuel and is thus found acceptable for extended burnup application. In addition, B&W has indicated that the Mark B fuel design is capable of meeting the above criterion for cladding collapse at the extended burnup level requested in this submittal.

15 a.- .- - .. . _ - . . -

t 1

i j (c) Overheating of Cladding

! Bases / Criteria - The design limit for the prevention of fuel failures due to j l overheating is that there will be at least 95% probability at a 95% confidence

level that departure from nucleate boiling (DNB) will not occur on a fuel rod l having the minimum DNBR during nonnal operation and anticipated operational occurrences. This design limit is consistent with the thennal margin criterion of SRP Section 4.2 and is found to be acceptable for extended burnup application.

t Evaluation - As stated in SRP Section 4.2, adequate cooling is assumed to exist l when the thermal margin criterion to limit the departure from nucleate boiling

'li (DNB) or boiling transition in the core is satisfied.

i

.l (d) Overheating of Fuel Pellets 1

q Bases / Criteria - As a second method of avoiding cladding failure due to over-j heating, B&W has indicated that they preclude fuel centerline melting for nor-i mal operation and anticipated operational occurrences. This design limit is

!I the same as given in Section 4.2 of the SRP and is found to be acceptable for i extended burnup application.

i

,; Evaluation - The B&W evaluation of the fuel centerline melt limit is performed with the TACO-2 (Ref. 8) fuel performance code. Although TACO-2 has been i previously approved we have reviewed the code with respect to extended burnup and conclude that it is acceptable up to the burnup level requested in the B&W extended burnup topical report (Reference 2).

i

! In evaluating the above criterion B&W uses a fuel melting temperature relation-ship that is burnup dependent (Ref. 2) This relationship has been ust.d exten-sively in the nuclear industry and has been reviewed with respect to extended burnup application and found to be acceptable.

(e) Pellet / Cladding Interaction i Bases / Criteria - As indicated in SRP Section 4.2, there are no generally appli-cable criteria for pellet / cladding interaction (PCI) failure. However, two 16

__ c -

4 acceptance criteria of limited application are presented in the SRP for PCI:

'l (1) less than 1% transient-induced cladding strain, and (2) no centerline fuel

$ melting. Both of these limits are used by B&W and have been found acceptable

! for extended burnup application, see Sections 2.0(b) and 3.0(d).

[ Evaluation - As noted earlier, B&W utilizes the TACO-2 code to show that their i fuel meets both the cladding strain and fuel melt criteria and this has been I

found acceptable for extended burnup applications, f In addition, B&W has stated that they have adopted several cladding design cri-teria to help mitigate the effects of PCI. In this respect the cladding is

specified so it 1) maintains sufficient ductility over its irradiated life, 2)

I has a high hoop creep rate to accommodate PCI strains, and 3) has a superior ID surface finish to minimize stress corrosion cracking (SCC) initiation sites.

B&W has also indicated that as a result of these criteria the B&W fuel design

(

has shown excellent performance with respect to PCI for burnups up to 40 mwd /kgM, I

and that additional fuel examinations are to be performed on 5 cycle fuel. From this, it is concluded that B&W fuel has been designed to help mitigate the effects of PCI and thus is acceptable for extended burnup application.

I (f)CladdingRupture i

Bases / Criteria - There are no specific design limits associated with cladding rupture other than the 10 CFR 50 Appendix K requirement that the incidence of rupture not be underestimated. The rupture model is an integral portion of the ECCS evaluation model which is discussed in Sections 4.0(a) and (c).

Evaluation - The cladding deformation ar,J rupture models used by B&W in their LOCA-ECCS analysis are directly coupled to their models for cladding ballooning and flow blockage. B&W has perfonned a generic ECCS evaluation of extended burnup fuel performance up to the requested burnup level and concluded that

! peak cladding temperatures, cladding rupture and balloning are most limiting I at beginning of life. The TACO-2 code, as discussed 3.0(d), is approved for

{ providing input to the ECCS analyses up to the requested burnup levels.

B&W has indicated that in order to insure that the generic TACO-2 ECCS analyses l are bounding, B&W will perform TACO-2 calcualtions for each reload application i}

! 17

5 .

l of extended burnup fuel. If fuel temperatures and internal rod pressures for

'I any specific reload application are not bounded by the generic TACO-2 analyses, l then B&W must perform ECCS analyses to insure that beginning of life conditions i are still limiting for this analysis. Further discussions of the impact of extended burnup operation on the LOCA-ECCS analysis are provided in Section 4.0(c).

I (g) Fuel Rod Mechanical Fracturing i

1

, Bases / Criteria - The tenn " mechanical fracture" refers to a cladding defect that is caused by an externally applied force such as a load derived from-4 core-plate motion or a hydraulic load. These loads are bounded by the loads of 4

(' a safe-shutdown earthquake (SSE) and LOCA, and the mechanical fracturing analy-sis is usually done as a part of the SSE-LOCA loads analysis [see Section L 5.0(d)ofthisSER].  !

Evaluation - The discussion of the SSE-LOCA loading analysis is given in Sec-tion 5.0(d) of this SER.

i

, 4.0 FUEL C00 LABILITY t

For accidents in which severe fuel damage might occur, core coolability must be maintained as required by several General Design Criteria (e.g., GDC 27 and 35).

In the following paragraphs, limits and methods to assure that coolability is maintained are reviewed for the severe damage mechanisms listed in the Standard i Review Plan.

i i I

l (a)FracmentationofEmbrittledCladding Bases / Criteria - The most severe occurrence of cladding oxidation and possible j fragmentation during a Condition III and IV accident is a result of a signifi-cant degree of cladding oxidation during a LOCA. In order to reduce the ef-facts of cladding oxidation for a LOCA, B&W uses an acceptance criterion of 2200*F on peak cladding temperature and 17% on maximum cladding oxidation as l prescribed by 10 CFR 50.46.

Evaluation - As noted in Section 3.0(f), B&W has perfonned a generic ECCS analysis to demonstrate that beginning of life conditions are most limiting i and will confirm this on a reload specific basis. Therefore, fragmentation i, of embrittled cladding is not an extended burnup issue for B&W fuel.

18

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I I .

i I (b) Violent Expulsion of Fuel

! l Bases / Criteria - In a severe reactivity initiated accident (RIA) such as a con-trol rod ejection accident, large and rapid deposition of energy in the fuel  !

could result in melting, fragmentation, and dispersal of fuel. The mechanical t action associated with fuel dispersal might be sufficient to destroy fuel clad-

ding and the rod-bundle geometry and to provide significant pressure pulses in the primary system. To limit the effects of an RIA event, Regulatory Guide j 1.77 recommends that the radially-averaged energy deposition at the hottest I axial location be restricted to less than 280 cal /g.

There is evidence (Ref.17) that the radial average enthalpy limit of 280 cal /g

) U02 may not maintain core coolability particularly for irradiated fuel rods.

Reference 17 further indicates that "neither severe fuel rod damage nor loss of normal geometry is expected at radial average peak fuel enthalpies below about

~

240 cal /g U02." The reference also indicates that while irradiated fue1 rods I failed at significantly lower enthalpies than unirradiated rods, there was lit-tle sensitivity, to the degree of burnup to approximately 33 mwd /kgM. However, it was noted that the number of previously irradiated rods, i.e., burnups I

greater than 4 mwd /kgM, were very small with only four rods with burnups great-

[ er than 10 mwd /kgM and two above 30 mwd /kgM. The two rods with burnups above 30 mwd /kgM had relatively benign failures with radial average fuel enthalpies of approximately 150 cal /g 002.

I I

The assumptions made in analysis of the CRA ejection accident and the methods used are very conservative, however, and could be refined to yield peak enthalpies substantially lower than presently calculated. There is therefore l little incentive' to reduce the 280 cal /g limit to a value like 240 cal /gm when h calculated values are less than 240 cal /gm and could be reduced further by eliminating the conservatisms in the calculations.

1 The B&W safety criteria for the control rod ejection accident are:

1. A CRA ejection accident shall not cause further violation of the RCS integrity.

l 19 i

_ . _ _ _ _ . _ _ _ _ _ _ . . . _ _ v r'

b ,

I

I

)

b The maximum fuel enthalpy for the hottest fuel rod shall not exceed 280 2.

1 cal /g.

i I ThesecriteriaareconsistentwiththecriteriaspecifiedinSection4.2oflhe SRP and thus are found to be acceptable for extended burnup applications.

l f Evaluation - The B&W analysis of a control rod ejection accident for a generic Mark B assembly at extended burnups has shown a peak fuel enthalpy less than half of the above criterion. Since this provides substantial margin to the 4

limit, the Mark B fuel design is found to be acceptable for generic applica-tions at the extended burnups requested in this submittal.

{

(c) Cladding Ballooning .

,g Bases / Criteria - Zircaloy cladding will balloon (swell) under certain combina-

}

i tions of temperature, heating rate, and stress during a LOCA. There are no

, specific design limits associated with cladding ballooning other than the 10 l CFR 50 Appendix K requirement that the degree of swelling not be underestimated.

Evaluation - As noted in Section 3.0(f) B&W has performed a generic ECCS analysis to demonstrate that beginning of life conditions are most limiting and will confirm this on a reload specific basis. Therefore, cladding ballooning is not an extended burnup issue for B&W fuel. However, a brief discussion of the effects of extended burnup on the LOCA analysis is provided for information

{ purposes.

The TACO-2 code has been approved by the NRC for providing initial input, e.g.,

stored energy and internal rod pressures, for the LOCA analysis. As noted ear-lier this code accounts for those steady-state fuel performance parameters that are burnup dependent.

With respect to cladding rupture and ballooning during the LOCA, there is evi-I dence that cladding oxidation at extended burnup levels and LOCA temperatures may result in reduced cladding strains (Ref. 18) from those traditionally pre-dicted for LOCA. These data are not conclusive, however, because the tests were not perfonned with an oxidizing atmosphere nor under irradiation conditions.

20

[ .

ll a

f Irrespective of whether these data are applicable to a LOCA, reduced cladding

!! strains would result in less flow blockage and thus the current analysis meth-ods would be more conservative with respect to this criterion. In addition, d the high cladding temperatures associated with the LOCA analysis will anneal any irradiation damage effects on cladding properties.

l (d) Fuel Assembly Structural Damage From External Forces Bases / Criteria - Earthquakes and postulated pipe breaks in the reactor coolant f

system would result in external forces on the fuel assembly. SRP Section 4.2 i! and associated Appendix A state that fuel system coolability should be main-tained and that damage should not be so severe as 'to prevent control rod inser-

!s tion when required during these low probability accidents. B&W endorses these

, criteria and they are found to be acceptable for extended burnup application.

Evaluation - The generic models i.:d analysis methods used by B&W to evaluate lL the combined safe shutdown earthquake (SSE) and LOCA loads have been described in Reference 19 and approved by the NRC (Ref. 20) up to current burnup levels.

I j;I B&W has indicated that the fuel rod-to-upper end fitting clearances, grid re-

!{ laxation, and holddown spring relaxation, all dependent on burnup, have been

!I examined for their effects on the SSE-LOCA loads. TN fuel rod-to-upper end

fitting clearances have been addressed in Section T. 0(g) and these clearances are used as input to this analysis. B&W has ind'cated that a limiting case of i a fully relaxed intermediate grid is used in toe vertical seismic and LOCA ana-l lyses. This decreases the fuel assemblies' natural frequencies but has an in-

,; significant effect on the spacer grid impact loads. The irradiation-induced lI relaxation of the holddown springs decreases the spring force on the fuel as-5' sembly but B&W has indicated that the spring rate is not affected and that this l has an insignificant effect on the dynamic fuel assembly response.

i' i

The material property that could have an impact on these analyses at extended l burnup levels is material ductility. These analyses could be impacted if clad-21

- ,,- . ~ _ - _ - . ,

h e .

P ding or assembly ductility were decreased, as a result of extended burnup oper-ation, to a level that would allow cladding or assembly failure not accounted for in the analysis. As noted in Section 2.0(b), the decrease in material duc-

!, tility is expected to be small for the increased exposure and burnup levels I requested and no adverse effects have been observed for assembly average

, burnups up to 50 mwd /kgM. From the above evaluation, it is c teluded that the above analysis methods are acceptable for extended burnup app.ication, f 5.0 NUCLEAR DESIGN i

Typical extended fuel burnup and increased fuel cycle length core designs uti-f lize higher fuel enrichments, low leakage patterns, burnable poison rods and/or axial blankets. Higher fuel enrichment is required to reduce the number of

(

feed assemblies and offset the reactivity loss resulting from the higher fis-g sion product inventory. The core neutron economy is improved by reducing the radial leakage using low leakage loading patterns in which the high burnup fuel is located on the core periphery. Axial blankets are used to flatten the axial burnup distribution and improve fuel utilization. The increased power peaking resulting from the larger reactivity differences between the fresh and high

, burnup fuel and the use of low leakage loading patterns is generally controlled j l using burnable poison rods.

4 These features affect the physics characteristics of high burnup core designs.

The increased fuel depletion in high burnup cores results in an increase in the plutonium fission fraction and the fisson product inventory, the higher pluto-nium fission fraction in turn hardens the neutron spectrum and increases the neutron production per unit energy. The increased fission product inventory l and use of burnable absorbers tends to increase absorption and also harden the

, neutron spectrum.

t While the increased fuel burnup does affect the core physics characteristics,

! the changes are relatively small and the physics parameters are determined us-i ing standard calculational methods and procedures. The high burnup neutronic effects enter through the microscopic cross sections and fuel assembly lattice group constants. The present calculations of these parameters account for sub-stantial levels of plutonium, fission products and burnable absorbers, and these methods are expected to adequately treat the neutronics changes associat-I 22

f I .

I t

ed with extended fuel burnup. The depletion methods used to track the plutoni-

{ um and fission product isotopics and various normalization procedures are also

[ expected to be equally valid for high burnup fuel configurations.

The high burnup fuel physics characteristics and core configuration affect the core nuclear safety parameters. The major effect is to increase the power in l

the low burnup and/or centrally located fuel assemblies and to decrease the power in the high burnup and/or peripherally located fuel assemblies. The re-f sulting increase in the number and power of the peak powered rods is typically i controlled by use of burnable poison rods.

I i The increased fission product inventory and use of burnable absorbers increases thermal and epithermal absorption and hardens the core neutron spectrum. These factors combine to reduce the boron and control rod worth, prompt neutron life-

[

time and Doppler coefficient. The moderator temperature coefficient may in-crease or decrease depending on the particular high burnup design, and is also controlled using burnable absorbers as in present core designs. The delayed neutron fraction is also reduced as a result of the increased plutonium fission j fraction.

In addition to improving the neutron e'conomy, the low leakage patterns reduce the pressure vessel damage fluence by shifting the power toward the center of a the core and away from the vessel. This fluence reduction is partially offset, t

however, by the harder neutron spectrum and increased neutron production (per MeV) of the high burnup fuel.

L The calculation of the high burnup core safety parameters is carried out using

{ the same core and lattice methods and procedures used for present core designs.

l' The changes in the core safety parameters resulting from the higher fuel burnup j',

designs tend to be relatively small as a result of the low relative importance of the high burnup fuel and the tendency for the increase in plutonium fission ,

rates and fission product inventory to saturate. These calculated safety par-l' ameters provide the core neutronics input to the required plant transient and f:

'; accident analysis.

! As the above discussion indicates, the effect of high burnup on the physics design is expected to result in relatively small changes in the predicted e

23

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I l*

characteristics of the core, and also relatively small extensions in range of I

+ the methods used to calculate the characteristics. Because high burnup fuel I is not subject to limiting duty and because of its low relative importance in determining the core characteristics, we conclude that present methods are adequate for high burnup designs. To provide added assurance that these lIi methods are adequate, we recomend that Babcock & Wilcox pay special attention to comparisons of predicted and measured physics parameters (particularly power distributions) which are monitored during the reactor cycle. A systematic pattern of deviation between predictions and measurements would provide an I indication of potential problems. We intend to take an active role in following

these comparisons.

6.0 RADIOLOGICAL CONSIDERATIONS OF POSTULATED ACCIDENTS WITH EXTENDED BURNUP OPERATION 4

To ensure that accidents involving the movement of fuel do not constitute an offsite health and safety issue, design events are assessed. Analyses of fuel

, handling accidents assume release of the entire volatile radionuclide fuel as-3 sembly gap and plenum inventory under nominally 23 feet of water after the as-

! sembly has cooled substantially (usually at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for BWR assemblies, 72 or 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for PWR assemblies). For assemblies with burnup up to 38,000 mwd /t batch average at discharge, Regulatory Guide 1.25 assumptions are used.

These stipulate an inventory of ten percent of the total fuel assembly iodines l and noble gases (with the exception of 30 percent for 85 Kr)inthegapand j plenum volumes released upon clad perforation. An iodine decontamination factor (DF)of100("EvaluationofFissionProductReleaseandTransportfora Fuel Handling Accident," G. Burley, USAEC, Revised Oct. 5,1971) is assumed for I 23 feet of water cover, and appropriate airborne radionuclide filtration / mixing, o if any, is applied in the analysis before release to the atmosphere. The de-contamination factor is based, in part, on an analysis of work presented in WCAP-7518-L, " Radiological Consequences of a Fuel Handling Accident," M. J.

! Bell et al June 1970, NES Proprietary Class 2." For fuel handling accident

,; offsite radiological consequence evaluations involving fuel assemblies with burnup greater than 38,000 mwd /t batch average at discharge (extended burne assemblies), the analysis is presently performed using Regulatory Guide 1.25 assumptions, but with modified gap and plenum fractional volatile radionuclide inventories. The fractional inventories range from a few percent (less than 24 l.

, ,L the R. G. 1.25 ten percent recommendation) to as much as 40-50 percent for

{ certain high burnups/radionuclide combinations. The gap and plenum fractional inventories for the highest power assembly are computed as a function of at least burnup, and at most time, temperature, and burnup using the GAPCON-THERMAL-2

,i computer code in conjunction with the ANS 5.4 fission gas release standard (model) proposed by the American Nuclear Society in " Radioactive Gas Release from LWR

{

{ Fuel", C. E. Beyer, draft NUREG CR-2715, April 1982. In generating these i estimated fractional inventories, the conservative assumption of fuel assembly

{ operation at a constant maximum-allowed peak linear heat generation rate (LHGR) for PWR's or MAPLHGR for BWR's is made. This assumption appears to be conservative

{

within a factor of 2-3 for gap and plenum volatile inventories.

In addition to the conservative assumption regarding fuel assembly power opera-tion noted above, there are two other significant sources of conservatism in the staff's analysis. The iodine decontamination factor (DF) assigned to the pool is taken to be a factor of 100. It can be inferred from the report upon

! which this factor is based (WCAP-7518-L) that this value is probably conserva-tive by about a factor of three. Finally, plateout of volatile iodine released l from the fuel into the gap and fuel rod plenum has been entirely neglected.

f Although not well quantified, a tentative estimate suggests that about 10 per-lh cent or less of the iodine released into the gap will remain volatile at the

,[ fairly low temperatures after the fuel has been allowed to cool for about a day or more.

Because of the significance of these conservatisms, the staff intends to study l t and quantify them in more detail and to use the results of such evaluations to appropriately revise the staff's Standard Review Plan (SRP), NUREG-0800. In 1 the interim, the staff concludes that consideration of all three factors to-gether noted above may permit a significant reduction of estimated thyroid dos-es compared to existing analyses. Adequate justification by licensees on a

, case-by-case basis, or by vendors on a generic basis, are likely to provide

, sufficient bases for departing from SRP criteria until such time as detailed changes can be made. A reduction by a factor of two is likely to be appropri-ate and conservative. Consequently, with regard to evaluation of thyroid dosec

. for fuel-handling accidents involving extended-burnup fuel (> 38,000 mwd / tonne),

1 and pending SRP revision, it is likely that justification can be provided for lower estimates of thyroid doses from fuel handling accidents by a factor of two in departures from SRP review criteria.

i 25 i

i __ _ _ _ _ . _ ._ _ _ . _ _ . _ _ . _ _ . - - _ _ . _ . . _ . _ _ _ . _ . .

I .

L -

(

7.0 CONCLUSION

S l-

{ The review of Babcock & Wilcox's submittal as described in BAW-10153P and re-

! sponses to NRC questions in References 3 through 5, for application of their ,

design criteria, analysis methods and generic approval of their Mark B fuel

! design to extended burnups has been completed. As a result of our review, we

conclude that the criteria, analysis methods and generic Mark B fuel design are

! applicable to licensing at the extended burnup level requested in this submittal '

I t (Ref. 2) with the provision that confirmatory data for specific SAFDLs be i obtained up to the burnup level requested in this submittal (Ref. 2). The

following confirmatory data have been identified in this review as being needed

l (1) fuel rod and annealed guide tube wear data from Mark B assemblies with ,

j Zircaloy grids, (2) rod and assembly growth data from Mark B assemblies with

] annealed guide tubes, and (3) assembly holddown spring relaxation data.

We consider the above items as confirmatory because we expect that the data will demonstrate acceptable performance to the requested burnup levels. Design

(

} specific applications of this high burnup technology therefore can be initiated I before the confirmatory data have been collected, but the confirmation must be I provided from lead test assemblies before the high burnup levels are reached by significant numbers of fuel assemblies. Plant specific applications for high burnup should address provision of this confirmation before the high burnup levels are reached.

6 l

With the above provisions, we have concluded that the B&W criteria, analysis methods and generic Mark B fuel design, as described in the extended burnup l topical report and response to questions, References 2 through 5, for extended l burnup application are adequate such that 1) fuel damage is not expected to lI occur as a result of normal operation and anticipated operational occurrences,

! 2) fuel damage during postulated accidents would not be severe enough to pre-vent control rod insertion when it is required, and 3) core coolability will always be maintained even after postulated accidents.

This conclusion is based on two primary factors:

1) B&W has provided sufficient evidence that the design criteria will allow for safe operation of the Mark B design fuel at the proposed extended burnup level, and 26

9 -

t -

.i

2) The B&W analysis methods used to assure that these criteria are met have f*

j been based on adequate extended burnup operating experience and prototype testing.

We recommend that licensees with B&W fuel and B&W should monitor fuel cycles l with extended burnup fuel and should inform the staff of any significant deviation between prediction and measurement of various physics parameters.

, Additionally, the staff concludes that in its offsite radiological consequence

! evaluations, because of the assumption of fuel assembly operation at constant maximum-allowed peak LHGR, assumed DF of 100 for the pool, and neglect of

! plateout credit for volatile gap radioiodines, an additional reduction of estimated offsite thyroid doses by a factor of two is appropriate and conservative.

li 4

l-e d

i

?

I

. r f

27 w -

-w ,w y , y ,-,w -, >>-c---, w- p -e e-w-r y p

i l -

k

8.0 REFERENCES

?

1. Letter, L. S. Rubenstein (NRC) to J. H. Taylor (B&W), dated June 2,1981.

f

2. " Extended Burnup Evaluation", BAW 10153P, Babcock & Wilcox, Lynchburg, I Virginia, September 1982.

t

3. Letter, J. H. Taylor (B&W) to D. Moran (NRC), " Request Number 2 for Additional Infonnation - BAW-10153P, ' Extended Burnup Evaluation,' January, 13, 1984," dated March 13, 1984.

4. Letter, J. H. Taylor (B&W) to C. O. Thomas (NRC), dated September 5,1984.

5. Letter, J. H. Taylor (B&W) to C. O. Thomas (NRC), " Extended Burnup Topical Report (BAW-10153)," dated February 15, 1985.

6. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear l

Power Plans--LWR Edition", NUREG/0800, Section 4.2, System Design, Rev. 2, July 1981.

f

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

i

8. Y. H. Hsii, et al., " TAC 02 - Fuel Pin Performance Analysis", BAW-10141PA, Babcock & Wilcox, Lynchburg, Virginia, June 1983.

9. Letter, C. O. Thomas (NRC) to J. H. Taylor (BRW), " Acceptance for Referenc-ing of Licensing Topical Report BAW-10141(P)", dated April 13, 1983.

10. C. J. Baroch, "Effect of Irradiation at 130, 650, and 775*F on Tensile Properties of Zircaloy-4 at 70, 650, and 775*F," Properties of Reactor Structural alloys After Neutron or Particle Irradiation, ASTM stp 570,

p. 129, 1974.

11. M. Shimada, et al., " Ductility Loss of Ion-Irradiated Zircaloy-2 in Iodine Environment," Effects of Radiation on Materials: Tenth Conference, ASTM STP 725, p. 233, 1981.

28

~

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f l 12. W. J. O'Donnell and B. F. Langer, " Fatigue Design Basis for Zircaloy l Components," Nuc. Sci. Eng., Vol. 20, p. 1, 1964.

13. Letter, C. O. Thomas (NRC) to J.' H. Taylor (B&W), " Acceptance for Referenc-I ing of Licensing Topical Report BAW 10147P", dated February 15, 1983.

i I

14. " Fuel Rod Bowing in Babcock & Wilcox Fuel Designs", BAW-10147P, Babcock &

Wilcox, Lynchburg, Virginia, April 1981.

15. F. R. Burke, " Reactor Internals Design Analysis for Normal, Upset and l

g Faulted Conditions", BAW-10060, Babcock & Wilcox, Virginia, June 1977.

16. " Program to Determine Inreactor Performance of B&W Fuels - Cladding Creep Collapse", BAW-10084, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, October 1980.

17. P. E. MacDonald, et al., " Assessment of Light Water Reactor Fuel Damage During a Reactivity Initiated Accident," Nuclear Safety, Vol. 21 No. 5,

p. 582, September 1980.

l

18. P. Hoffman, " Influence of Iodine 'on the Strain and Rupture Behavior of I

Zircaloy-4 Cladding Tubes at High Temperatures," Zirconium in the Nuclear Industry, ASTM STP 681, p. 409, American Society for Testing and Materials, 1979.

I

19. S. J. Shah, R. E. Lied, " Mark C Fuel Assembly LOCA-Seismic Analysis",

l BAW-10133(P) Revision 1, Babcock &Wilcox,Lynchburg, Virginia,May

! 1979.

l. 20. Letter, C. O. Thomas (NRC) to J. H. Taylor (B&W), " Acceptance for Referenc-ing of Licensing Topical Report BAW-10133(P) Revision 1", dated October 13, 1982.

l.

29

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