ML13276A064

ANP-3248NP, Revision 1, Areva RAI Responses for Browns Ferry Atrium 10XM Fuel Transition, Enclosure 3

ML13276A064
Person / Time
Site:	Browns Ferry
Issue date:	09/30/2013
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
ANP-3248NP, Rev 1
Download: ML13276A064 (79)

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Enclosure 3ANP-3248NP, AREVA RAI Responses for Browns Ferry ATRIUM-10 XM Fuel Transition

-Non Proprietary

'

ANP-3248NP Revision1 AREVA RAI Responses forBrowns Ferry ATRIUM IOXMFuel Transition September 2013AAREVAAREVA NP Inc.

AREVA NP Inc.ANP-3248NP Revision 1AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition AREVA NP Inc.ANP-3248NP Revision 1AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition Copyright

© 2013AREVA NP Inc.All Right ReservedAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page iNature of ChangesItem Page Description and Justification

1. AllThe page numbering was incorrect in the Table of Contents andSection 1.0. Minor changes to References 1 and 31. Updated SNPBRAI-17 consistent with Reference

1. No additional changes weremade.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 10XMFuel Transition ANP-3248NP Revision 1Page iiContents1 .0 In tro d u c tio n ..................................................................................................................

1-12.0 NRC Questions and AREVA Response

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2-13 .0 R e fe re n c e s ...................................................................................................................

3-1TablesTable SNPB RAI 8-1Table SNPB RAI 11-1State Point'Comparisons at Rated Power ..........................................

2-31Sensitivity of Pellet Conductivity onCOTRANSA2/XCOBRA-T

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2-39Table SNPB RAI 17-1 [.................................................................................

2 -5 0Table SNPB RAI 17-2 [.................................................................................................

2 -5 1Table SNPB RAI 18-1 Impact of Thermal Conductivity Degradation on BrownsFerry ATRIUM 1OXM LOCA Analysis Results ....................................

2-55Table SNPB RAI 26-1 Browns Ferry Unit 2 Cycle 19 Overpressurization Biasesa n d R e s u lts ........................................................................................

2 -6 3Table SRXB RAI 2-1 Contribution of Total Predicted Rods in BT by NuclearF u e l T y p e ...........................................................................................

2 -6 6This document contains a total of 76 pages.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page iiiFiguresFigure SNPB RAI 1-1Figure SNPB RAI 1-2Figure SNPB RAI 1-3Figure SNPB RAI 3-1Figure SNPB RAI 3-2Figure SNPB RAI 3-3Figure SNPB RAI 4-1Figure SNPB RAI 4-2Figure SNPB RAI 4-3Figure SNPB RAI 4-4Figure SNPB RAI 4-5Figure SNPB RAI 5-1Figure SNPB RAI 5-2Figure SNPB RAI 5-3Figure SNPB RAI 6-1Figure SNPB RAI 6-2Figure SNPB RAI 6-3Figure SNPB RAI 11-1Figure SNPB RAI 12-1Figure SNPB RAI 17-1Figure SNPB RAI 17-2Lattice Reactivity Comparison at Same Enrichment

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2-4Isotopic Depletion Variation, BLEU-CGU

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2-5Fissile Isotope Variation, BLEU-CGU

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2-5Browns Ferry Equilibrium Cycle RODEX4 MaximumC orrosio n R esults ..........................................................................

2-10Liftoff Measurement Data on AREVA ATRIUM-10 Fuelat the Brow ns Ferry U nits ..............................................................

2-11Zircaloy-2 Stress Relieved Cladding Oxide, Historical Liftoff M easurem ent D ata ..............................................................

2-13ATRIUM 10A Standard FUELGUARD

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2-15ATRIUM 10A Improved FUELGUARD Bottom View ......................

2-16ATRIUM 10A Improved FUELGUARD Side View ..........................

2-16ATRIUM 1OXM Improved FUELGUARD Bottom View ...................

2-17ATRIUM 1OXM Improved FUELGUARD Side View .......................

2-18ATRIUM-10 Rod Bow MCPR Penalty ............................................

2-20MCPR Penalty Model vs. Test Data ...............................................

2-21ATRIUM 1OXM and ATRIUM-10 95/95 % gap closure ...................

2-23Channel Fluence Gradient Distribution for BrownsFerry U nit 2 C ycle 19 .....................................................................

2-26Browns Ferry Unit 2 Cycle 19 Reference LoadingP a tte rn ...........................................................................................

2 -2 7Browns Ferry Unit 2 Cycle 19 FLCPR for Assemblies Exceeding Database Bounds .........................................................

2-28Fuel Thermal Conductivity Relative to No BurnupDegradation as a Function of Temperature andE x p o s u re .......................................................................................

2 -4 0Allowable Transient Overpower Ratio versus RodN odal Exposure

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2-44[] ............................................................................................

2 -5 1[I ............................................................................................

2 -5 2AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 1-11.0 Introduction Tennessee Valley Authority (TVA) submitted a License Amendment Request (LAR) to changethe Browns Ferry Technical Specifications in support of reload fuel transition to ATRIUMTM1OXM *. In response to the LAR, the US Nuclear Regulatory Commission (NRC) has issued aninitial set of questions, in the form of Request for Additional Information (RAI), Reference 1.Based on the information provided in this report, TVA will prepare a formal response to the NRCRAIs.* ATRIUM is a trademark of AREVA NP.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-12.0 NRC Questions and AREVA ResponseThe NRC questions (i.e., RAIs) listed below are according to Reference 1:SNPB RAI-1 ANP-3159P, Section 1.0It has been stated in your LAR submittals that TVA intends to continue use of blended lowenriched uranium (BLEU) for the manufacture of fuel pellets for the ATRIUM 1OXM fuel design.(a) Apart from the difference in density of the BLEU fuel from commercial grade fuel, listother differences in the BLEU fuel such as isotopic composition, physical properties, andneutronics characteristics from the commercial grade fuel.AREVA Response:

The primary difference between BLEU and commercial grade uranium (CGU) is theconcentration of the uranium isotopes of U234 and U236. BLEU material has a higherconcentration of these isotopes when compared to the maximum allowed values for enrichedCGU defined by ASTM C966-10.

Chemically, there is no difference between BLEU and CGU.Within the fuel manufacturing

process, the U234 and U236 isotopes are inseparable from itsoriginal BLEU feed stock.Both CGU and BLEU material is subject to the same maximum U235 enrichment of 4.95%. Thefollowing table provides a CGU versus BLEU comparison of U23' and U236 concentrations.

TheCGU allowable values are from the ASTM C966-10 specification with the U234 equivalent weightpercent based upon the maximum allowable U235 enrichment.

The BLEU concentrations arealso based upon feed material at the maximum allowed enrichment.

Enriched CGU from ASTM C966-10Typical BLEUmax allowable Equivalent Isotope concentration Weight %

• Weight %U234 1. 1OE+04 pg1gU235 0.0546 wt% U 0.09 wt% UU236250 pg/gU 0.025 wt% U 1.60 wt% U* For material enriched to 4.95% U235.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-2The BLEU feed material may be used directly in 4.95 wt% rods or down blended for lower rodenrichments.

The commercial grade natural uranium used in the blending process contains0.0057 wt% U234 and no U236.The small changes in isotopic impurities of the BLEU fuel do not significantly affect the physicalproperties of the fuel. The physical properties for U02 and U02-Gd2O3 fuel are identified in theRODEX4 topical report (Reference

2) with details provided in RODEX4 theory manual (detailed in Reference 4 of BAW-10247PA).

Isotopes of uranium (e.g., U234, U235, U236 and U238) have the same electronic structure.

Theyalso occupy the same space. Consequently, the substitution of a U234 or U2316 for a U238 (or U235)atom in the lattice does not constitute a point defect and does not change the local electronic configuration.

The fuel thermal conductivity is therefore independent of the U234 and U236content as it is also independent of the amount of U235.As discussed in the response to (b) below, the impact on the fuel isotopic composition duringdepletion is small. Since the fissile isotopic inventory does not significantly deviate from normalexpected variations the corresponding changes in fission products are insignificant.

Becausethe changes are very small, these differences will result in a change in fuel thermal conductivity that can be neglected.

For the same reasons the thermal conductivity is not affected by the presence of U234 and U236,other thermal mechanical properties are also not affected.

This includes thermal expansion, heat capacity,

enthalpy, Young's modulus, Poisson's ratio, creep, melting temperature andemissivity.

The fuel density is slightly less by an insignificant amount. As a result, the physicalproperties used in the RODEX4 models are applied to BLEU fuel without change.The primary difference in neutronic characteristics of BLEU relative to CGU fuel is decreased reactivity due to the higher concentration of U236.The U236 isotope has neutron poisoning impact. For the BLELI assemblies that have a combination of BLEU and CGU rods the neteffect is a reduction of reactivity approximately equivalent to a 0. 3% reduction in U235enrichment.

In other words, the enrichment of a BLEU assembly would need to beapproximately that much higher to provide the same amount of energy production.

However,AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-3since both assemblies are limited to the same maximum enrichment value the impact is usuallyseen in a larger required batch fraction for a BLEU versus an equivalent CGU reload.This reduction in reactivity is illustrated in Figure SNPB RAI 1-1 in which ATRIUM IOXM latticesare compared.

In this comparison, the lattices are identical except the additional U234 and U236was removed for the CGU lattice (i.e., both lattices have exactly the same U235 enrichment andgadolinia distributions).

(b) If the isotopic content of the BLEU fuel is different from that of the commercial gradefuel, what is the impact on the buildup of various uranium isotopes during the depletion of the fuel?AREVA Response:

The CASMO-4/MICROBURN-B2 code system explicitly models the U234 and U236 with cross-section data for a range of temperatures and voids. The behavior of these uranium isotopesunder irradiation is well understood.

The lattice depletion (CA SMO-4) and 3D core simulator (MICROBURN-B2) codes track these isotopes to account for the off-spec concentrations.

The impact on the usage of the BLEU material is accounted for by explicitly including the U234and U236 isotopic concentrations in the fuel design and licensing process.

Design changes toaddress the presence of the higher concentration of the U236 include increasing the reload batchsize, modifying lattice enrichments and gadolinium loading.Figure SNPB RAI 1-2 illustrates the difference in the isotopic buildup and depletion for the samecomparison lattices used in the previous reactivity comparison.

At beginning of life the primarydifferences are in the U216 and U231 isotopes.

The U238 difference is simply a compensating reduction at BOL due to the inclusion of the U234 and U236 isotopes.

Focusing on the significant fissile isotopes, Figure SNPB RAI 1-3 shows a slight increase in fissile inventory with exposurefor the BLEU lattice.

This is due to the neutron poisoning effect due to the presence of the U236and the conversion of U234 to U235. These changes in fissile isotope quantities do notsignificantly differ from the variations normally experienced in a reload due to changes in fuelenrichment and gadolinia

loadings, or operation.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-4The BLEU material has been used in 10 reloads for the Browns Ferry units with the first reloadin Unit 2 Cycle 14.1.21.1CCCM1.00.90.80.7010 20 30 40GWd/MTU50 60 70Figure SNPB RAI 1-1Lattice Reactivity Comparison at SameEnrichment AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-5151050-5EU'I-CI----------10-150 10 20 30 40 50 60 70GWd/MTU--U-234----U-235--!!-U-236

-aU-238Figure SNPB RAI 1-2 Isotopic Depletion Variation, BLEU-CGU4-Cam0.50.40.30.20.10-0.1-0.2-0.3-0.4-0.5~ w u- .,amp0 10 20 30GWd/MTU40 50 60 70--U-235--m-- PU-239-,4- PU-241Figure SNPB RAI 1-3 Fissile Isotope Variation, BLEU-CGUAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-6SNPB RAI-2 ANP-3159P Section 3.2.3 of ANP-3159P, page 3-5 indicates that "LHGR (linear heat generation rate)margins are provided along with uncertainties due to channel bow for input to the statistical analysis."

Provide details of how the channel bow uncertainties are incorporated in to thestatistical analysis.

AREVA Response:

The uncertainty in the calculated channel bow leads to an associated uncertainty in the fuel rodpower level. This uncertainty in power is taken into account as part of the RODEX4 statistical application methodology.

A series of steps are carried out to assess the effect of channel bowand its associated model uncertainty on the fuel rod thermal-mechanical behavior by accounting for channel bow in the generation of the fuel rod power histories.

F]AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-7fIThe above description is consistent with the methodology described in Reference

2. Additional information is contained in the third round of RAI responses to this RODEX4 topical report.The RODEX4 results presented in ANP-3159P for Browns Ferry Unit 2 Cycle 19 include theadjustments as described above to account for power uncertainties from channel bow.SNPB RAI-3 ANP-3159P Section 3.2.7 of ANP-3159P indicates that a program is in progress to monitor crud buildup andoxidation as water chemistry changes are implemented.

(a) Provide details of this program.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-8AREVA Response:

TVA instituted a healthy fuel examination program in the mid 2000s, as part of establishing abaseline for fuel performance at Browns Ferry Nuclear.

The fuel inspections supporting thisprogram are performed by the fuel vendor. The scope typically involves the inspection of oneassembly from each batch following completion of an operating cycle. The highest exposureonce burnt bundle is held out for inspection; a high exposure twice burnt bundle, and a highexposure thrice burnt bundle, are also inspected.

The inspection typically includes a peripheral examination of the bundle with the channel removed, to assess general performance andensure no abnormal physical distortion is present.

A limited number of fuel rods are removed(typically six), washed to remove the loose crud, and measurements of liftoff (oxide plustenacious crud) are taken on each removed fuel rod, along with profilometry measurements andeddy current testing for flaws. The scope of the inspections in the Browns Ferry Nuclear healthyfuel inspection program exceeds the guidance for post irradiation surveillance discussed insection 4.2 of the NUGEG-0800 Standard Review Plan. The scope of the program has beenexpanded with the introduction of On Line Noble Chemistry (OLNC).TVA implemented OLNC at Browns Ferry Unit 3 beginning in Cycle 15. At the time TVAinstituted the OLNC injection there was no operating experience of AREVA fuel with OLNC, andsubsequently, TVA and AREVA initiated a long term program to validate that OLNC does notresult in a decrease in fuel reliability for AREVA fuel. The initial step in this program was toestablish a baseline by performing exams on fuel from Browns Ferry Unit 3 Cycle 14 for once,twice and thrice burnt assemblies.

A detailed visual examination of each fuel assembly andchannel was performed.

Detailed examinations were performed for six fuel rods from each fuelassembly, consisting of.:* Washing* Eddy current testing for flaws" Profilometry

" Liftoff thickness measurements

" Visual examinations Additionally, crud samples were obtained for two rods on the once burnt assembly at twolocations-one sample with a brush and one with a blade, for a total of eight samples.

Thesesamples were then analyzed by AREVA to characterize the crud deposits for Browns FerryUnit 3 prior to the initiation of the OLNC program.

Specifically:

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-9" Bulk chemical analysis of selected fuel deposit particulate samples.

These sampleswere analyzed for elemental and metallic oxide distribution and possibly for thedetermination of crystal grain size.* Fuel deposit flakes were selected for evaluation of the OD and ID features of thedeposits, including porosity and density, the distribution and size of boiling chimneys, and the elemental and metal oxide distribution within any given flake.AREVA used the data obtained, as well as Browns Ferry Unit 3 water chemistry data, tobenchmark the AREVA crud and corrosion risk assessment tool for Browns Ferry Unit 3 specificconditions and performed a detailed risk assessment for Cycle 15.TVA and AREVA performed a similar post-irradiation exam following Cycle 15-once, twice andthrice burnt assemblies with crud scrapes obtained for rods from the once burnt assembly.

AREVA is in the process of characterizing the crud scrapes and will update the detailed riskassessment.

Exam and crud analyses are planned following Browns Ferry Unit 3 Cycles 16 and 17 tocomplete this program.Unit 3 is the only Browns Ferry Unit currently using OLNC. Water chemistry assessments willbe performed to determine the applicability of the Unit 3 results to the other Browns Ferry Unitswhen OLNC is implemented for those Units.(b) Provide details as to how the guidance on treatment of corrosion, crud, andhydrogen content per NUREG-0800 Standard Review Plan (SRP), Section 4.2 issatisfied, andAREVA Response:

Fuel rod corrosion is addressed by following the approved RODEX4 methodology described inReference

2. Details on the statistical

results, handling of SER restrictions on the methodology, and the accounting for crud are provided below. As described in the response to (a) above, thewater chemistry changes are being carefully implemented to ensure the fuel is not adversely affected.

Following completion of the program, any changes to crud conditions will be taken intoaccount in the fuel rod analyses, as required.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-10The RODEX4 fuel rod analysis methodology makes use of a statistical approach that involvessampling fuel rods in the reload batch. Input uncertainties for power level, model parameters, and manufacturing design parameters are randomly varied based on known distributions of theinputs. The results are treated to demonstrate that a large fraction of the rods [] U02 rods from the equilibrium cycle batch are shown in Figure SNPB RAI 3-1 below.Figure SNPB RAI 3-1 Browns Ferry Equilibrium Cycle RODEX4Maximum Corrosion ResultsAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-11The maximum calculated corrosion is f]An SER restriction imposed on RODEX4 requires that the calculations account for an expected, design basis crud thickness and it may be based on plant-specific history.

As part of the RAIresponses to the RODEX4 topical report, it is stated that the existing corrosion model includes adesign basis level of crud. That is, the liftoff measurement data used to benchmark theRODEX4 corrosion model include normal, low levels of crud for the plants represented by themeasurement data. If the plant-specific measurements indicate abnormal crud levels, theanalyses for that plant must take into account a design basis crud thickness that can be derivedfrom the plant-specific data.In the case of the Browns Ferry plants, plant-specific liftoff data are available from the fuelsurveillance program described in the response to (a) above. Figure SNPB RAI 3-2 showsrecent eddy-current liftoff data acquired at the Browns Ferry plants on AREVA A TRIUM-IC fuel.The data are identified in the legend as, for example, "BFE2, EOC16" (Browns Ferry Unit 2 atthe end of cycle 16).Figure SNPB RAI 3-2 Liftoff Measurement Data on AREVA ATRIUM-10 Fuelat the Browns Ferry UnitsAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-12The liftoff data represent the [1. The liftoff measurement technique includes any tenacious crud in addition to oxide.Fuel assemblies with the highest end-of-cycle exposures were selected for measurement at theBrowns Ferry units with the intent to obtain liftoff values that conservatively represent the reloadbatches.Also shown in the figure are the data used for the RODEX4 corrosion model benchmark andmodel uncertainty evaluation.

The data are identified in the legend with reactor codes and dateof examination (e.g., All, 2001) and are from AREVA BWR l0x10 fuel. The Browns Ferry liftoffdata fall within the database range and follow the trend of the model benchmark data.The Browns Ferry liftoff measurements are interpreted to exhibit [ slightly] after three 2-year cycles (approximately 49000 hours) bounds the corrosion and crudconditions at the Browns Ferry plants.As part of the RODEX4 methodology, the approved limit for corrosion is [ ]. During thefirst reload application of RODEX4, the limit was challenged by the NRC because of a concernabout the effect of spallation on the cladding integrity.

Spallation can produce localized surfacediscontinuities in the cladding and may also result in the formation of hydride lenses that couldcause premature failure.

To avoid the issue of spallation, the limit was reduced to [ I.The [ ] limit was established from a review of historical liftoff measurement data onAREVA BWR fuel. Figure SNPB RAI 3-3 displays the data.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-13Figure SNPB RAI 3-3 Zircaloy-2 Stress Relieved Cladding Oxide, Historical Liftoff Measurement DataThe maximum measured value from the data was []The maximum calculated oxide value off]. If higher liftoffdata are encountered as part of the water chemistry program at the Browns Ferry units, the crudinputs to the RODEX4 analyses will be adjusted as already required by the SER restriction onthe RODEX4 methodology to ensure the crud levels are properly taken into account.Currently, the RODEX4 code does not include an approved hydrogen pickup model nor is therean approved hydrogen concentration limit. Consistent with a preceding Technical Specification AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-14revision that was updated and approved to include the RODEX4 methodology (Reference 3in ANP-3159P),

the application of the RODEX4 methodology with the fJ which is a conservative method to ensureadequate cladding strength and ductility are maintained, thus satisfying the guidance inSRP 4.2.(c) In ANP- 3159P Section 3.2.7, it is stated, in part, that as a result of concerns thatwere raised on the effect of non-uniform corrosion, such as spallation, and localized hydride formations on the ductility limit on cladding, a regulatory commitment wasmade to reduce the limit oxide limit to the value in Reference 3 that is listed forANP-3159P.

Provide details of how this regulatory commitment to reduce the oxidelimit to the value specified in Reference 3 of ANP-3159P is implemented at BFNUnits 1, 2, and 3.AREVA Response:

TVA to provide response.

SNPB RAI-4 ANP-3082P Thermal hydraulic compatibility and characterization analyses have been performed and theresults are summarized in ANP-3082P.

The transition cores for the Browns Ferry units consistof ATRIUM 10 with both Standard FUELGUARD (SFG) and Improved FUELGUARD (IFG) lowertie plates. Provide a detailed description of the differences between SFG and IFG with respectto their geometry (preferably using drawings),

contribution to the pressure drop and contribution to thermal margin performance by the improvement in design of the FUELGUARDs.

AREVA Response:

The ATRIUM IOA with Standard FUELGUARD consists of 36 blades (34 without drain holesand 2 with drain holes), and 8 grid rods. Blades are assembled in slots and grid rods arethereafter inserted and brazed together.

See Figure SNPB RAI 4-1 hereafter.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-15exPLFR BUSHINWATER CHANNEBUSH I NGGRID?XCURVED BLADE WITH-' \-CURVED BLADE W/ODRAIN HOLES DRAIN HOLESFigure SNPB RAI 4-1 ATRIUM IOA Standard FUELGUARD The Improved FUELGUARD is similar to the Standard FUELGUARD; however there are 34 halfinterstitial strips that run parallel to the grid rods, both directly below and between (also below)them. These are utilized to increase filter efficiency.

See Figure SNPB RAI 4-2 and FigureSNPB RAI 4-3.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-16$aCUVE BLADE DRAIN HOLESSIXURVED BLADE WflHDRAIN HOLESFigure SNPB RAI 4-2 ATRIUM IOA Improved FUELGUARD Bottom ViewINTERSTITIAL Figure SNPB RAI 4-3 ATRIUM IOA Improved FUELGUARD Side ViewNote: The difference in orientation is based on third vs. first angle projection.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-17For clarity, Figure SNPB RAI 4-4 and Figure SNPB RAI 4-5 are included below to illustrate theATRIUM IOXM Improved FUELGUARD depiction, the difference being the PLFR Bushingquality and locations.

CURVED BLADEWITH DRAINHOLESINTERSTITIAL STRIP-MIX 1-OHWATER CHANNELBUSHINGROD'\-N"LFR BUSHING-URVED BLADEWITHOUT DRAINHOLESFigure SNPB RAI 4-4 ATRIUM IOXM Improved FUELGUARD Bottom ViewAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-18Figure SNPB RAI 4-5 ATRIUM IOXM Improved FUELGUARD Side ViewNote: All Figures were taken from modified (for clarity) manufacturing drawings.

The impact on pressure drop between the SFG and IFG LTP is best shown in the full coreevaluations shown in Tables 3.9 and 3. 10 of ANP-3082P.

The core pressure drop at ratedconditions for a full core of ATRIUM- 10 with SFG is [ ] psid and a full core with IFG isI ] psid. The core pressure drop at off-rated conditions for a full core of ATRIUM-10 withSFG is [ ] psid and a full core with IFG is [ ] psid. These differences between the twoLTPs are not significant on the core pressure drop.The impact on thermal margin performance between the SFG and IFG LTP is apparent whenthe two tie plates are resident in the same core loading since. pressure drop, core flow splits,and leakage rates affect the critical power. In Tables 3.9 and 3.10 of ANP-3082P, the resultsprovided for "Transition Core Loading 2"presents the critical power performance for the samecore conditions applied to a SFG LTP assembly and IFG LTP assembly.

The results at ratedconditions show for the A TRIUM- 10 assembly f]AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-19SNPB RAI-5 ANP-3082P Section 3.4. ANP-3150P, Section 3.3.5It is stated in the above section that the rod closure due to rod bow was assessed for impact onthermal margins.(a) Describe how the CPR penalty was determined as function of exposure, AREVA Response:

AREVA's BWR rod bow CPR penalty is based on rod closure and was derived using openliterature data. Based on this data, it was concluded that thermal margins were not substantially reduced for closures up to 30%.AREVA's model application for A TRIUM-IC type fuel was presented in an informational submittal to the NRC, EMF-95-52(P),

Reference 22.The discussion on gap closure behavior versus fuel exposure, and the attendant effects on CPRthermal margins, is given in 5(c).(b) Provide an assessment of how the thermal margin calculations are affected by the rodbow at various exposures, AREVA Response:

The thermal margin versus rod bow (% closure) for the A TRIUM-IC fuel design is presented below.The discussion on gap closure behavior versus fuel exposure, and the attendant effects on CPRthermal margins, is given in 5(c).AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-20Figure SNPB RAI 5-1 ATRIUM-10 Rod Bow MCPR PenaltyTo assure that this model is conservative, AREVA ran a CHF test on an ATRIUM-10 bundle inwhich two rods were welded together.

The results of that test are shown in Figure SNPBRAI 5-2. As is seen from the plot, AREVA over predicts the penalty by a factor of 2.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-21Figure SNPB RAI 5-2 MCPR Penalty Model vs. Test Data(c) Justify your prediction that less rod bow for ATRIUM 1OXM than for ATRIUM 10 byshowing typical analysis/calculations, and provide how the rod bow behavior is impactedby fuel burnup.AREVA Response:

AREVA uses the NRC approved correlation described in topical report XN-NF-75-32(P)(A)

Supplement I (Reference 19). The correlation was developed[

request of the NRC as discussed in Reference

19. [] at theIAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1 OXMFuel Transition ANP-3248NP Revision 1Page 2-22IIBased on the correlation above the [J primary factors impacting rod bow are fIThe ATRIUM IOXM [] Figure SNPB RAI 5-3 belowcompares the predicted rod bow for both the ATRIUM-10 and the ATRIUM 1OXMf]AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-23Figure SNPB RAI 5-3 ATRIUM IOXM and A TRIUM-IO 95/95 % gapclosureSNPB RAI-6 ANP-10307PA, Section 2.2. Channel BowIt has been stated in Section 2.2.1 the channel growth correlation used to determine the channelbow magnitude is a continuous function of fast fluence from the beginning to the end of life. Themodel coefficients were computed using databases consisting of channel length measurements acquired by AREVA from European Boiling Water Reactors (BWRs) and from Pressurized Water Reactor guide tube data.(a) Provide supporting details to demonstrate that the above mentioned channel bowdatabase is applicable to Browns Ferry units' operating conditions.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-24AREVA Response:

The calculated fast fluence gradients for the core design provided in Reference 3 have beencompared to the upper and lower bounds of the channel bow database.

As shown in the tablebelow and Figure SNPB RAI 6-1, the Browns Ferry Unit 2 Cycle 19 reference core designremains bounded by the upper but not the lower bound of the channel bow database.

A total of four bundles were found to have channel fast fluence gradients slightly below thelower bounds of the channel bow database.

These bundles are identified in the following table.As shown in Figure SNPB RAI 6-2, all assemblies that were identified to exceed the bounds ofthe channel fast fluence database are located in low power locations near the core periphery (i.e., located one row in from the outside of the core). This is an expected result due to theincreased core leakage at the periphery which can result in increased fluence gradients.

Due tothe low power in these locations, the affected assemblies have significant margin to the coreAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-25limiting CPR throughout cycle operation.

This is illustrated in Figure SNPB RAI 6-3 in which theFLCPR (Fraction of Limiting Critical Power Ratio) is plotted for each of the affected assemblies along with the core limiting MFLCPR. Consequently, the potential impact on SLMCPR isinsignificant due to the margin exhibited by the assemblies expected to exceed the bounds ofthe channel fast fluence database.

However, as noted in the response to part (b) below, it isproposed that future core designs will be subject to an equivalent license condition.

This qualitative assessment for Browns Ferry Unit 2 Cycle 19 was quantitatively confirmed byapplying the proposed license condition.

In this case, an augmented uncertainty was applied tothe four bundles identified as exceeding the lower bound of the database.

LMCPR calculations using SAFLIM3D with the augmented channel bow uncertainty for theaffected assemblies were performed on a consistent basis as analysis using the standardchannel bow uncertainty.

The table below provides a comparison of the SLMCPR percent ofboiling transition rods between the calculated SLMCPR results provided in Table 2 ofReference 32 and Table 4.2 of Reference 5 (listed below as "without proposed licensecondition')

and the application of the license condition.

Comparison of the results show that theimpact of applying the augmented channel bow uncertainty to the assemblies of interest is notsignificant.

Percent of rods in boiling transition Loop SLMCPR SLMCPR results without SLMCPR results withConfiguration proposed license condition proposed license condition 1.04 0.0834 0.0834TLO1.06 0.0417 0.04031.05 0.0921 0.0849SLO1.08 0.0331 0.0316Small differences in the results are expected even though the assemblies are in low power, non-limiting locations.

The power distribution in the affected assemblies and the neighboring AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-26assemblies are affected the most by the increased channel bow uncertainty.

The effect on thepower distribution for assemblies further away is greatly diminished

[]Figure SNPB RAI 6-1 Channel Fluence Gradient Distribution forBrowns Ferry Unit 2 Cycle 19AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-27Figure SNPB RAI 6-2 Browns Ferry Unit 2 Cycle 19 Reference Loading PatternAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 10XMFuel Transition ANP-3248NP Revision 1Page 2-28c.LL1.00.90.80.70.60.50.40.30.2*~ ~ ~ ~ ~~~o =N4 2 maa** ~l u*ij0510Cycle Exposure (GWd/MTU) 1520-Core Limiting

• FBD258A FBD262 x FBD266 .FBD270Figure SNPB RAI 6-3 Browns Ferry Unit 2 Cycle 19 FLCPR forAssemblies Exceeding Database Bounds(b) During review of Brunswick Steam Electric Plant (BSEP) ATRIUM 10XM fuel transition LAR, the Nuclear Regulatory Commission (NRC) staff determined that the predictive model for channel bow was validated against empirical data that was not bounding ofBSEP's expected performance.

To resolve this issue, the licensee for BSEP agreed toincrease the channel bow uncertainty in the SLMCPR calculation for the most severelydeflected fuel channels.

In view of the excessive channel bow that occurred at BSEP alicense condition was proposed for BSEP Units 1 and 2 in connection with the use ofAREVA channel bow model outside the range of the channel bow measurement database from which its uncertainty was quantified

(

Reference:

Letter, BSEP 13-0002,from Michael J. Annacone (Duke Energy) to NRC, "Supplement to License Amendment Request for Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, CORE OPERATING LIMITS REPORT (COLR), and Revision to Technical Specification 2.1.1.2 Minimum Critical Power Ratio Safety Limit," Duke Energy, January22, 2013.) Confirm whether a similar license condition is required for the BFN Units 1, 2,and 3.AREVA Response:

TVA to provide response.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-29SNPB RAI-7 ANP-3082P.

Section 3.5Discuss the impact on bypass exit subcooling while transitioning between transition corecombination of AREVA fuel and GE14 to a full core ATRIUM 1OXM fuel design at BFN units.AREVA ResDonse:

Table 3.9 of ANP-3082P presents the rated conditions bypass flow for various core loadings, either full core of GEl 4, A TRIUM-IO, and ATRIUM IOXM, and various representative transition loadings of these fuel types. The AREVA design criterion for the bypass is based onmaintaining bypass flow fractions (refer to Section 4.1.5 of Reference 18). Fuel designs areconsidered to be hydraulically compatible when the bypass flow characteristics of the reload fuelassemblies do not differ significantly from the existing fuel in order to provide adequate flow inthe bypass region. If needed to achieve similar bypass flow fractions between fuel designs, thebypass flow hole of the lower tie plate is modified.

Inherently, if bypass flow is maintained,

[] when transitioning cores.The largest difference in core bypass flow fraction between any of the full cores or multipletransition core loadings for GE14 and AREVA fuel is [ ] of rated core flow.The actual transition scenario for Browns Ferry is represented by "Transition Loading 3". Thecore bypass flow fraction between the transition loading and full core ATRIUM IOXM is [] of rated core flow. The insignificant impact on the core bypass flow fractions resultsin[ ]SNPB RAI-8 ANP-3082P.

Section 3.2Table 3.4 of ANP-3082 provides input conditions for thermal hydraulic compatibility analysis fortwo of the statepoints 100 percent power/1 00 percent flow and 62 percent power/37.3 percentflow.(a) Provide the basis for the thermal margin analysis performed at 62 percent power/37.3 percent flow statepoint andAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-30AREVA Response:

Results presented for 62%P/ 37.3%F in Table 3.10 of ANP-3082(P) provide comparisons of thecalculated critical power ratio performance of the GE14, A TRIUM-10, and ATRIUM IOXM fuelassemblies for both full core and transition core loadings.

The addition of this off-rated statepoint is to demonstrate compatibility is maintained for both rated and off-rated conditions.

[(b) Justify why the analysis was not done at 100 percent power/1 05 percent flow asindicated in Figure 1.1, BFN Power Flow Map -100 percent original licensed thermalpower (OLTP) of ANP-3167(P),

BFN Unit 2 Reload Safety Analysis.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-31AREVA Response:

Hydraulic compatibility analyses are performed to demonstrate operation within the power/flow map. The report provides a demonstration of hydraulic compatibility at rated and off-rated conditions.

Performing calculations at increased core flow and different off-rated statepoints willnot change this conclusion.

As an example, Table SNPB RAI 8-1 provides results originally performed for the initialtransition to A TRIUM-10 fuel for the Browns Ferry extended power uprate. The key parameters of interest in the table are the relative differences between the statepoints.

As seen in the table,relative differences between GEl4 and A TRIUM-IO fuel are comparable between thestatepoints; therefore, the 100%P/100%F statepoint is adequate in demonstrating compatibility at rated conditions, the 100%P/105%F statepoint is not needed to further demonstrate compatibility.

Table SNPB RAI 8-1 State Point Comparisons at Rated PowerILAssembly Flow(klbm/hr)

Statepoint GE14 ATRIUM-10

% Difference Percent AssemblyBypass FlowStatepoint GE14 ATRIUM-10 Difference ISNPB RAI-9 ANP-3150P.

Section 3.3.1. Table 3.1Provide details of the procedure, assumptions, methodology and results for the "stressevaluations" that were performed to confirm the design margin and to establish a baseline foradding accident loads for the determination of loading limits on fuel assembly components.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-32AREVA Response:

As discussed in ANP-3150P Section 3.3.1 (Reference 24), [ ] thefuel assembly structural components do not receive significant loads during normal and AO0conditions.

fJ No analyses areperformed to confirm design margin under normal operating and AO0 conditions fITo ensure the structural integrity off I Section III of the ASME Boiler andPressure Vessel code (Reference

25) is used to establish acceptable design limits. To evaluatethe stresses under normal operating conditions,

[J The maximum normal operation

[] for BFE2-19 is then compared againstthe limit to ensure that adequate margin is maintained.

To evaluate the stress under AOO and accident conditions,

[IAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-33IIFor the [] the normal operating stresses

[] The designmargin is confirmed by comparing the resulting stress to the design limit as defined by SectionIll of the ASME Boiler and Pressure Vessel code (Reference 25).RAI-1O provides additional details on how [ ] atBFE2-19 were calculated.

Additional information on the stress evaluation results andcomparison to the load limits can be found in Table 3. 1, Section 3.4.4 of Reference 24.SNPB RAI-10 ANP-3150P.

Section 3.4.4(a) Describe AREVA ATRIUM-10 and ATRIUM 1OXM fuel assemblies' dynamic structural response to combined seismic/loss-of-coolant accident (LOCA) loadings.

Providedetails of the model used for assembly with and without a fuel channel, acceleration used in the calculations, uncertainty allowances in the calculations, and results withmargin to established limits.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-34AREVA Response:

A plant specific analysis was not performed because it was not required.

A change in fuelassembly design may not necessitate a full reanalysis if it can be shown that the fuel design isdynamically similar to the fuel assembly design in the Reactor Pressure Vessel (RPV) seismicanalysis of record (ANF-89-98(P)(A),

Reference 18, Section 3.2.7). A comparison betweenfuels []There is an existing seismic analysis of record which uses a reference GNF legacy fuel type.The fuel acceleration from this analysis for an Operating-Basis Earthquake (OBE) is [ Iand for Safe-Shutdown Earthquake (SSE) this value is [ ] The SSEacceleration can be applied to the A TRIUM-IC and ATRIUM IOXM designs iffI.The first reload of the ATRIUM-IC, documented in EMF-2971(P)

Revision 1 (Reference 27)concluded that the [J Thus, the existing reactor seismicanalysis of record was not reanalyzed for the ATRIUM- 10 and remained applicable for BrownsFerry Unit 3 reload (Reference

27) and follow on reloads (References 28 and 29) referenced inBFN UFSAR Amendment 23.Regarding the ATRIUM IOXM, the structural response to combined seismic/LOCA loadings[J The current reload of the ATRIUM IOXM is supplied with thesame 100/75-mil Advanced Fuel Channel (AFC) as the ATRIUM-IC design, with a fuelassembly

[ J. This results in a [] accepted designs at BrownsAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-35Ferry. Thus, the [the channeled fuel assembly will experience an SSE acceleration off] and.The table below compares the dynamic properties of the GNF analyzed fuel of record, theATRIUM-IO and ATRIUM IOXM.There are no specific criteria with respect to comparing dynamic properties for BWR fuel.AREVA defines this threshold as f] of the analyzed design. This threshold for dynamic compatibility has beenused in both PWR and BWR evaluations and is documented in the RAI responses and SER ofEMF-92-116(P)(A),

Revision 0 (Reference 23).Since the ATRIUM IOXM is found to be dynamically similar to the analyzed fuel of record themargin to results is found by comparing the fuel channel acceleration limit off Ito thef J. This provides aAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-36The fuel assembly analysis design criteria are established in ANF-89-98(P)(A)

(Reference 18).The [ ] is used as the design input to a staticfinite element analysis of the A TRIUM- 10 fuel assembly components (load chain, fuel rods,water channel, tie plates, and spacer grids) that demonstrates acceptance to ASME mechanical design criteria in a seismic event. The analysis also confirms that the [] as documented in the current and historical topical reports, (References 26, 30, and 31). Therefore, if the fj For added conservatism theFEA static analysis model assumesI.The ATRIUM- 10 fuel assembly component analysis t1. Theallowable stress or load limits for the ATRIUM IOXM were updated to new limits based ontesting of ATRIUM IOXM components.

This information is tabulated in ANF-3150P, Table 3-1,Criteria Section 3.4.4 (Reference 24).Any uncertainty presented in the analyses is accommodated by a large degree of conservatism given a margin greater than a J between the maximum imposed acceleration andthe allowable acceleration.

(b) Provide details of the testing done to obtain the dynamic characteristics of the fuelassembly and spacer grids under varying conditions of stiffness, natural frequencies anddamping values with and without the fuel channel.

Provide details of the evaluation ofBFN ATRIUM 10/ATRIUM 1OXM fuel assembly structural response to externally appliedforces (seismic and LOCA) and show how the acceptance criteria in NUREG-0800, Chapter 4.2, Appendix A,Section IV are satisfied.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-37AREVA Response:

While a full test campaign was conducted for the ATRIUM IOXM, channeled and un-channeled, no dynamic testing conducted on the ATRIUM IOXM was used to support the transition to theATRIUM IOXM at Browns Ferry in regards to the structural response of the fuel, see (a).Testing was utilized to determine the ATRIUM IOXM fuel assembly component allowable loading for the spacer grid and tie plates. The dynamic properties, e.g., fuel channel stiffness and mass, of the fuel assembly were calculated and verified through testing for both achanneled and un-channeled assembly and were used for[]. The full array of testingconducted for the ATRIUM I OXM design is discussed in Section 4.0 of Reference 24.Structural response to externally applied forces is discussed in part (a). Fuel assemblyacceptance criteria per the Standard Review Plan are listed in Reference 18 and Reference 24using the same numbered criteria sections; Table 7.3 in Reference 18 and in Table 3-1, 3-2 and3-3 in Reference

24. Acceptance criteria of Section IV, Appendix A of the SRP Chapter 4.2 areaddressed in Table 3-1, Criteria Section 3.4.4 of Reference 24.SNPB RAI-1I ANP-3170P Section 3.1The core average gap conductance used in COTRANSA2 system calculations and the hotchannel gap conductance used in XCOBRA-T hot channel calculations are obtained fromRODEX2 calculations.

The sensitivity to conductivity and gap conductance for Anticipated Operational Occurrence (AOO) analyses is in the opposite directions for the core and the hotchannel.

This means that putting more energy into the coolant (higher thermalconductivity/higher gap conductance) is nonconservative for the system calculation butconservative for the hot channel calculations.

Provide, with quantitative

examples, how thesecompeting effects between the core and hot channel calculations are balanced to minimize theoverall impact of thermal conductivity degradation.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-38AREVA Response:

As background information, sensitivity analyses demonstrating the trends for gap conductance were previously presented in Section 3.2 of Reference 4.Both gap conductance and pellet conductivity are components of the fuel rod thermal timeconstant.

For thermal conductivity degradation (TCD), pellet conductivity is degraded resulting in an increase in thermal resistance and an increase in the rod thermal time constant.

A lower(degraded) conductivity in the system model (COTRANSA2) results in an increased lag in thefluid response to the changing neutron power. For a limiting pressurization event, the lowerconductivity results in an increase in reactor power due to the lag in void formation that wouldotherwise mitigate the power rise. This increase in transient reactor power results in a largerreduction of thermal margin during the event; therefore, the lack of modeling TCD in the systemis non-conservative.

However, an increase in the rod thermal time constant causes a hold up ofheat in the fuel pellet and results in a lag in the change of heat flux at the cladding/fluid interface.

In the hot channel calculations (XCOBRA-T),

this effect is seen as a reduction in heatflux response during the event which leads to a smaller reduction of thermal margin; therefore, the lack of modeling TCD in the hot channel is conservative.

To provide quantitative examples of these competing effects related solely to pellet conductivity, transient analyses were performed for the FWCF event for BFN Unit 2 Cycle 19 at 100% Power/ 105% Flow at end-of-cycle (EOC). The impact of TCD was assessed[

] The impact of TCD, based on RODEX4 ATRIUM IOXM studies,was obtained from Figure SNPB RAI 11-1. [J The results are provided in TableSNPB RAI 11-1. In the table, base refers to unaltered pellet conductivity.

As seen in theresults, the trends are consistent with the previous paragraph.

As noted, the reduction in pelletconductivity for both the system and hot channel tend to minimize the overall impact of thermalconductivity degradation; for this specific

example, the thermal margins slightly decreased whendegraded conductivity was utilized.

This overall trend is consistent with the discussion inAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-39Reference 8 Section 3. 1; the impact of thermal conductivity degradation is small relative to theconservatism in the CO TRA NSA 2/XCOBRA-T methodology.

Table SNPB RAI 11-1 Sensitivity of Pellet Conductivity onCOTRANSA2/XCOBRA-T COTRANSA2 XCOBRA-T Change inACPR ACPR fromPellet Conductivity Pellet Conductivity Original CaseAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-40Figure SNPB RAI 11-1 Fuel Thermal Conductivity Relative to NoBurnup Degradation as a Function of Temperature and ExposureSNPB RAI-12 ANP-3145P, ANP-2637 Section 3.0Nuclear core design analyses establish operating margins for minimum critical power ratio(MCPR), maximum average planar LHGR (MAPLHGR),

and LHGR. Two exposure dependent LHGR limits are established for each fuel design; one a steady state operating fuel design limit(FDL) and the other for the protection against the power transient (PAPT) limit.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-41(a) Provide the details of the FDL and PAPT limits as a function of exposure and show thatsufficient margin exists between the steady state and transient LHGR limits,AREVA Response:

The RODEX4 methodology used for the ATRIUMTM IOXM fuel differs significantly from theprevious RODEX2 methodology.

One of the areas of difference is in its treatment of transients, for example no direct equivalent to the previous Protection Against Power Transient (PAPT)limit exists for fast transient protection.

Transient protection with RODEX4 is obtained as a multi-step process.

First, the fuel designlimit (FDL) itself is established to ensure that the thermal-mechanical criteria are met usingpower histories from expected operation but generated assuming an anticipated operational occurrence (AO0) occurs during the life of the assembly.

Specifically, transient responses forthe quasi steady-state events off] The RODEX4licensing topical report (Reference

2) details the approved methodology used to calculate theFDL including transient evaluations.

The transient evaluation approach is also described insome detail in the response to part (b) of this question.

The results of the cycle-specific licensing analyses are also used to establish power and/or flowdependent multipliers (or set-downs) to the LHGR limits where needed. These power and flowdependent multipliers are identified as LHGRFACp and LHGRFACf, respectively.

Theapplication of these multipliers ensures that the thermal mechanical criteria are met throughout the operating domain for both steady-state operation and potential licensing transients.

Therequired LHGR limits and associated set-downs are provided in the cycle specific ReloadAnalysis report provided for each operating cycle. This is included in Tables 8.4 (steady-state AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-42LHGR limits),

8.5 (LHGRFACp),

and 8.6 (LHGRFACf) for the reference cycle provided for theLAR (Reference 5).(b) Show that the transient LHGR design limit satisfies the strain and fuel overheating design criteria.

AREVA Response:

In evaluating AQOs, the events are divided into two basic categories

-slow transients and fasttransients.

Both sets of analyses are performed using the RODEX4 thermal-mechanical methodology.

As a result of analyzing various events, sets of LHGR multipliers are established that limit the transients, as necessary, such that the analyses satisfy the transient criteria forcladding strain (1.0% strain limit) and fuel overheating (fuel melt limit). A summary of theanalysis methodology used in the Browns Ferry transition analysis is provided below.A slow transient is defined as one that can be analyzed using a steady-state solution.

The slowor steady-state transient LHGR design limit for RODEX4 is expressed in terms of an allowable overpower ratio for a transient.

This ratio is defined as the maximum tolerable rod nodal powercalculated during a transient divided by the steady-state power level just prior to the transient.

This ratio is determined such that the transient design criteria are satisfied.

Examples of slowtransients would be a CRWE (control rod withdrawal error), flow runup, or LOFWH (loss offeedwater heater).Fast transients, by nature, occur over a very short period of time with potentially high neutronflux levels. These fast transients are typically caused by a pressurization event and areevaluated separately.

The RODEX4 topical report (Reference

2) describes the code and associated methodology.

Applications examples in the topical report demonstrate the analysis of fast transients.

Slowtransients, which are also evaluated as part of the reload licensing

process, are treated similarly to obtain an allowable transient overpower ratio. The ratio is determined to be able to deriveflow-dependent multipliers on the LHGR limit to protect the cladding transient strain and fueltemperature criteria from anticipated transients occurring during operation.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-43Following the method described for evaluating transients in the RODEX4 topical report, a[IAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-44Figure SNPB RAI 12-1 Allowable Transient Overpower Ratio versusRod Nodal ExposureIn the plot, the minimum ratio off]A cycle-specific analysis was also performed for Unit 2 Cycle 19 using the cycle neutronics design and cycle power history data. The results displayed above bound the Cycle 19 results.The minimum ratio of[] is generally sufficient to support operation.

For fast transients, the RODEX4 input of exposure dependent power history for the limiting fuelrod includes a description of fuel operation at the applicable FDL LHGR immediately prior to thefast transient.

To this power history, a conservative prediction of the transient power excursion from a COTRANSA2 calculation is added to the RODEX4 calculation.

The analysis isperformed at various exposure increments.

A series of[ IAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-45f J For eachfast transient considered, the FDL fraction is decreased (set-down) as necessary to satisfy thestrain and temperature criteria.

The analysis performed per this methodology is used toestablish power-dependent limits on the maximum LHGR that can be allowed for the fuel designand operating cycle of interest.

As indicated in part (a) above, the combination of the slow transient allowable overpower ratioand the fast transient methodology are used to develop the LHGRFACp and LHGRFACfmultipliers required for cycle operation to ensure the strain and temperature criteria are satisfied during a transient event. Application of the FDL and the power/flow dependent multipliers protects the thermal-mechanical criteria during both normal expected operation or during apostulated AOO.(c) Confirm that the fuel Thermal Conductivity Degradation (TCD) with exposure has beentaken into account in generating/adjusting the LHGR limits.AREVA Response:

The FDL and LHGR multipliers are evaluated using RODEX4 following the approvedmethodology.

RODEX4 takes into account thermal conductivity degradation.

The FDL andLHGR multipliers are therefore generated and adjusted while taking into account thermalconductivity degradation effects.(d) Section 2.0 of ANP-3145P suggests that for a complete description of fresh reloadassemblies, see Reference 6 which is listed as ANP-3144(P)

Revision 0, Nuclear FuelDesign Report BFE-19 LAR ATRIUM 1OXM, August 2012. This report is not available tothe NRC staff. Please submit a copy of this report to the NRC or provide a completedescription in your response.

AREVA Response:

TVA to provide the requested report.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-46SNPB RAI-13 ANP-3148P.

Section 3.3Describe how the hot excess reactivity and shutdown margin are maintained per Technical Specification values during the transition cycles and during the equilibrium cycle of operation ofthe Browns Ferry units.AREVA Response:

TVA to provide response.

SNPB RAI-14 ANP-3140, Section 3In BWR terminology, assembly critical power is defined as the minimum assembly power thatresults in onset of Boiling Transition (BT) (dryout) at any location in the assembly.

Theacceptance criterion related to BT in BWRs is that the limiting value of Minimum Critical PowerRatio (MCPR) such that at least 99.9 percent of the fuel rods in the core are not expected toexperience BT during normal operation and AOOs. The SLMCPR is determined such that atleast 99.9 percent of the fuel rods in the core are expected to avoid BT if the MCPR is greaterthan or equal to the SLMCPR. The SLMCPR methodology uses a critical power correlation tocalculate CPR for a fuel assembly based on thermal hydraulic conditions and power distribution in the assembly.

(a) Section 3.3.1 of ANP-3140P describes how an additive constant (that accounts for localeffects such as spacing and geometry) is determined based on predictions of the criticalpower correlation and comparisons to test data. Figure 3-2 of ANP-3140P identifies thetest bundle where majority of the rods were dryout was observed.

Explain how the testresults (with majority rods in dryout) are used in the accurate determination of CPRcorrelation additive constants?

AREVA Response:

Figure 3-2 has all test results overlaid (a total of ] tests). It is observed that across all tests,many positions have had one or more indications of dryout, and some other positions have hadno indications of dryout. If one examines a single test, [ J are peaked and it iscommon to observe only a single rod indicate dryout. As described in section 3.3. 1, anexperimental additive constant is determined for each of the rods as the value that causes thecritical power correlation to exactly predict the critical power for that measurement.

In the caseof the rod indicating dryout, this experimental additive constant is a best estimate.

In the caseAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-47that the rod was not observed to dryout in that measurement, the experimental additive constantis conservative (it is assumed in the computation of the experimental additive constant thatdryout was observed; higher power is required to observe dryout on the rod).(b) Figure 3-5 of ANP-3140P lists additive constants comparison of original ACE/ATRIUM 1OXM and revised ACE/ATRIUM 1OXM. Section 3.3.4 indicates that the observedchanges in additive constants are generally small. However, a random check by staff onthe changes in additive constants has shown that the changes vary from 2 percent to 80percent in magnitude.

Please explain why there are larger variations contrary to what isstated in Section 3.3.4.AREVA Response:

Because additive constants have both positive and negative values, percent differences can bequite exaggerated when the base value of the additive constant is near zero. If a difference inadditive constant is calculated for each rod position (revised value -original value), thedifference in additive constant ranges from f 1, with an average magnitude ofchange oft ].[] Thus, theconclusion is that the observed changes in additive constant are generally small.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-48SNPB RAI-15 ANP-3140, Section 5.0Provide a detailed summary of the K-factor method that is described in Section 5 of ANP-3140.

If the methodology appears in an NRC-accepted topical report, please refer to the topical reportand identify the appropriate sections of the topical report that discuss the K-factor method.AREVA Response:

An imposed requirement on the critical power correlation is that the critical power is monotonic with K-factor.

This means that as the K-factor is increased, the critical power decreases.

Thisis achieved fIf the K-factors of two rods are compared at each axial level, the limiting rod and another rodthat may be close to limiting, it may be found that the limiting rod has the highest K-factor of thetwo rods in several axial nodes, but not all nodes. For the sake of this example, let the other rodhave the highest K-factor at the remaining axial nodes. [J Thus, a conservative critical power is calculated fAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-49IISNPB RAI-16 ANP-3152, ANP-3167P While reviewing ANP-3167P, Browns Ferry Unit 2 Cycle 19 Reload Analysis, the NRC staffnoticed that the Reference 7 listed in Section 9.0 of ANP-3167P, ANP-3153(p)

Browns FerryUnits 1, 2 and 3 LOCA-ECCS (emergency core cooling system) Analysis MAPLHGR Limit forATRIUM 1OXM Fuel, is NOT included in your LAR package.

Submit this reference to the NRCfor staffs review and evaluation of LOCA-ECCS analysis with the MAPLHGR limits for theATRIUM 1OXM fuel design.AREVA Response:

TVA to provide the requested report.SNPB RAI-17 ANP-3152P Section 2.0Provide detailed summary results of the analyses that are referred to in the second and thirdsentences of page 2-2.AREVA Response:

IIAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-50I] The results support the statement

[ITable SNPB RAI 17-1 [IAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM IOXMFuel Transition ANP-3248NP Revision 1Page 2-51Table SNPB RAI 17-2 [IFigure SNPB RAI 17-1 [IAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-52Figure SNPB RAI 17-2 [ISNPB RAI-18 ANP-3152P, ANP-3170P RODEX4 computer code was used to assess the potential impact of exposure (burnup)degradation of U02 thermal conductivity (TCD) on the fuel parameters and this assessment wasused to make adjustments to input for RODEX2 TCD with exposure that was used in the LOCAanalysis.

(a) Provide details of the adjustments made to RODEX2 input for LOCA analysis.

AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-53AREVA Response:

The original Browns Ferry LOCA analyses were performed without adjustments for TCD(References 6 and 7). The impact of TCD was addressed separately in AREVA's corrective action program and summarized in Reference 8, Section 3.2 and Table 3-1. In summary, thelimiting PCT was not impacted since it occurred at beginning of life (BOL) where TCD is notpresent.The approach taken to address TCD at the time of the analyses was an interim process.

Theimpact of TCD was determined by running RODEX4 fuel rod depletions and then repeating them with an input option selected to turn off TCD. Ratio corrections for key parameters weredetermined from the RODEX4 runs and then applied to the RODEX2 output that is used inHUXY. Specifically, the RODEX2 output parameters modified by the ratios were:Specifically for Browns Ferry, the ATRIUM IOXM LOCA analyses of Reference 8 were repeatedwith the new analysis process for evaluating the effects of TCD for comparison.

The newAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-54results are provided in Table 18-1. As shown in the table, the results re-affirm the conclusions of Reference 8 that show no impact on the limiting PCT.(b) Provide a discussion of impact of gap conductance between the fuel pellet and thecladding on ECCS performance analysis results resulting from fuel thermal conductivity degradation with exposure and irradiation effects on the ceramic fuel.AREVA Response:

As noted in (a),[IAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-55Table SNPB RAI 18-1 Impact of Thermal Conductivity Degradation on Browns Ferry ATRIUM IOXM LOCA Analysis ResultsTemperature Temperature (OF) Difference (OF)AveragePlanarExposure(GWd/MTU)

PCTWithoutTCDPCT WithTCDIncreasein PCT Margin todue to LimitingTCD PCTSNPB RAI-19 ANP-3152Figure 6.19 of ANP-3152P illustrates the variation of cladding temperature with time for thelimiting recirculation line break. The cladding temperature experiences the first peak at around190 seconds in to the transient.

Please explain the cause of this intermediate temperature peak.AREVA Response:

The event consists of a small break where ADS is a key factor in depressurizing the vesselduring the blowdown period. As shown in Table 6.2 and Figures 6.1 and 6.3 of Reference 6, theADS valves depressurize the vessel at approximately 190 seconds.

This occurs before the endof blowdown/time of rated core spray. This depressurization causes a lower plenum flashthrough the core which cools the fuel as seen in Figures 6.16 and 6.17 of Reference 6.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-56Following the flash, the node of interest returns to a quality of 1, the heatup then continues untilPCT is reached.SNPB RAI-20 ANP-3152.

Section 8.0Explain the basis for applying a factor of 0.85 multiplier to the two-loop operation (TLO)MAPLHGR limit for the single-loop operation (SLO) Single failure-Battery (DC) power board A(SF-BATTIBA)

LOCA analysis.

AREVA Response:

During SLO the pump in one recirculation loop is not operating.

A break may occur in eitherloop, but the results for a break in the inactive loop would be similar to those from a TLO break.A break in the active loop during SLO will result in an earlier loss of core heat transfer relative toa similar break occurring during TLO. This occurs because there will be an immediate loss ofjet pump drive flow resulting in a large drop in core cooling.

Therefore, fuel rod temperatures will increase faster in an SLO LOCA relative to a TLO LOCA. Also, the early loss of core heattransfer will result in higher stored energy in the fuel rods at the start of heatup. Applying anSLO multiplier to the MAPLHGR limits can reduce the increased severity of an SLO LOCA.f] To achieve this goal, a SLOmultiplier of 0.85 was chosen. This resulted in a SLO PCTof 17470F, compared to a TLO PCTof 1909 OF. Since SLO is at reduce power and flow, operation with a reduced MAPLHGR limit isnot overly restrictive.

SNPB RAI-21 ANP-3152Explain the impact on LOCA (EGOS performance) analysis and Title 10 of Code of FederalRegulations, Section 50.46 acceptance criteria for BFN Units 1, 2, and 3 coastdown operation with final feedwater temperature reduction (FFTR) as well as operation with feedwater heatersout-of-service (FHOOS).AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-57AREVA Response:

Coastdown with FFWTR or FHOOS does not create a limiting condition for the LOCA analyses.

I] Therefore, the combination of coastdown with FFTR orFHOOS is not a limiting condition for 10 CFR 50.46 acceptance criteria.

SNPB RAI-22 ANP-3167P.

Section 4.2BFN Unit 2 Cycle 19 SLMCPR calculated from Reference 12 (ANP-10307P) resulted in a valueof 1.04 for TLO and a value of 1.05 for SLO and listed in Reference 8 (Document No. 51-9191258-001).

These SLMCPR values are conservatively increased to 1.06 for TLO and 1.08for SLO. Provide basis for the adoption of these new values.AREVA Response:

TVA to provide response.

SNPB RAI-23 ANP-3167P.

Section 4.3Provide a summary of the analysis and results for BFN units operation with FFTR and FHOOSthat complies with the licensing requirements for the Option III (Oscillation Power Range Monitor(OPRM)) stability solution.

Also provide details and results for the analysis that supports BFNoperation with backup stability protection

regions, if the Option III OPRM system is declaredinoperable.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-58AREVA Response:

BFN is licensed for the FFTR and FHOOS modes of operation, with a reduction in finalfeedwater temperature of 55 degrees F at the current licensed power of 3458 MW. The use ofreduced feedwater temperature was included as an assumption in the engineering evaluation performed in support of the power uprate program (which increased the licensed power levelfrom 3293 MW to 3458 MW). This evaluation was included as Enclosure 5 to Technical Specification change TS-384, submitted on October 1, 1997 (ADAMS Accession:

9710070314).

NRC approval of this LAR was provided in Technical Specification Amendment 254 for Unit 2,and Amendment 214 for Unit 3 (ADAMS Accession:

ML042670045).

For Unit 1, TVA made a similar request to uprate the power level to 3458 MW This requestwas made as Supplement 1 to Technical Specification change TS-431, and was submitted onSeptember 22, 2006 (ADAMS Accession:

ML062680459).

The use of FFTR and FHOOS wasalso considered in this uprate request, and the use of reduced feedwater temperature wasspecifically discussed within the associated NRC SER approving this uprate request (ADAMSAccession:

ML063350404).

The approval of the uprate was provided in Unit I Technical Specification Amendment 269.The OPRM setpoint calculation and the methodology applied is described in Section 4.3 ofReference

5. This calculation was performed in accordance with the requirements of the NRCapproved licensing topical report NEDO-32465-A, (Reference 9), with the addition of thecalculation of a plant specific DIVOM (Delta over Initial CPR Versus Oscillation Magnitude).

AREVA plant specific DIVOM calculations follow BWROG guideline GE-NE-0000-0028-9714-RO, (Reference 10), and are performed in accordance with NRC approved licensing topicalreport BAW-10255(P)(A)

Revision 2 (Reference 11). As noted in both the BWROG guideline and the AREVA licensing topical report (Reference 11), DIVOM is not sensitive to changes infeedwater temperature.

The OPRM setpoint results provided in Table 4.3 of Reference 5 include two columns ofOLMCPR values. The column labeled OLMCPR(SS) presents results to protect the fuel from apostulated oscillation occurring during steady-state operation at reduced power and flowconditions.

The column labeled OLMCPR(2PT) presents results to protect the fuel from anAREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-59oscillation postulated to occur following a two recirculation pump trip transient.

[] Therefore, the results provided in Table 4.3 of Reference 5 boundboth nominal and reduced feedwater temperature conditions.

f]AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-60The Backup Stability Protection (BSP) calculation is performed on a cycle specific basis toconfirm or define scram and controlled entry regions on the power flow map. These regions aredefined by decay ratios that indicate a higher potential for a stability event to occur. The BSPand the associated methodology applied is described in section 4.3 of Reference

5. The resultsprovided in Table 4.4 of Reference 5 provide two separate scram and controlled entry regions,one to be applied during normal feedwater temperature operation and one to be applied whenoperating with reduced feedwater temperature (i.e., FHOOS or FFTR).The BSP analysis was performed following the guidance provided by the BWROG in 0G02-0119-260, (Reference 12). Calculations were performed using the NRC approved STAIF code,EMF-CC-074(P)(A)

Volume 4 Revision 0 (Reference 13). The required STAIF acceptance criteria

]was met.Specifically, for Browns Ferry Unit 2 Cycle 19 the following table provides the maximumcalculated decay ratios for the scram and controlled entry regions defined in Table 4.4 ofReference 5.SNPB RAI-24 ANP-3167P, Section 5.1.3For feedwater controller failure (FWCF) event scenario, Figure 5.4 (Percent Rated Versus Time)indicates a sharp spike in relative rated power to about 375 percent and a simultaneous sharpreduction in relative steam flow at about 15 seconds in to the event. Please describe the causeof this behavior during the FWCF event.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-61AREVA Response:

Due to a failure of the feedwater control system to maximum demand, there is a continual rise invessel water level that eventually reaches the high water level trip setpoint.

The high waterlevel trip causes the turbine stop valves to close in order to prevent damage to the turbine fromexcessive liquid inventory in the steam lines. Valve closure creates a compression wavetraveling back to the core, causing void collapse and subsequent rapid power excursion.

In thisexample, the valve closure occurs just prior to 15 seconds.

The closure of the turbine stopvalves also initiates a reactor scram and a recirculation pump trip. The reactor scram results ina loss of steam production and the sharp decrease in steam flow occurs.SNPB RAI-25 ANP-3167P, Section 6.1Section 6.1 reports that ATRIUM 1OXM LOCA analysis for BFN Units PCT is 1903 OF and thepeak local metal water reaction is 1.16 percent.

However, ANP-3152P, Browns Ferry Units 1, 2and 3 LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel, Table 6.1 reports that the PCTis 1909 OF and the maximum local cladding oxidation is 1.20 percent.

Clarify the discrepancy between the information on LOCA results from these two documents.

AREVA Response:

[] The limiting PCT of 1903 9F, as presented in Section 6.1 of Reference 5, is obtainedfrom the MAPLHGR report (Reference 7).SNPB RAI-26 ANP-3172P.

Section 7.1Table 7.1 lists the ASME Overpressurization analysis results for maximum vessel pressure forlower-plenum and the maximum dome pressure.

Please specify whether the pressure resultsAREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 2-62include any other analysis/measurement uncertainties in addition to the 7 pounds per squareinch increase that binds a bias in the void-quality correlations as indicated.

AREVA Response:

The pressure results provided in Table 7. 1 of Reference 5 do not include any otheranalysis/measurement uncertainties in addition to the 7-psi bias increase due to the void-quality correlation.

The bias assessment is discussed in detail in Sections 3.2 and 4.2 of Reference 14.However, there are additional issues related to Doppler-effects and exposure-dependent thermal conductivity degradation.

The impact of these effects on the representative Unit 2Cycle 19 core design is presented in Table SNPB RAI 26-1. The evaluation of these additional issues did not challenge the pressure limits.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-63Table SNPB RAI 26-1 Browns Ferry Unit 2 Cycle 19Overpressurization Biases and Results.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-64RAIs by Reactor Systems Branch (SRXB)SRXB RAI-1In page 2 of Enclosure to the LAR, it was stated,"The current ACE correlation for XM fuel has an identified deficiency.

The deficiency involves a nonconservatism in the axial averaging process used to determine the Kfactor, which is an input to the correlation

...TVA is including a BFN specific ACEsupplement in Attachments 27 and 28."Did the deficiency in the current ACE correlation identified above impact BFN Unit 2 Cycle 19SLMCPR values? If it did, was the updated methodology as provided in the generic and/orBFN-specific ACE supplements used for BFN Unit 2 Cycle 19 SLMCPR calculation?

Explain ifupdated ACE correlation was not used.AREVA Response:

The deficiency in the current ACE correlation did have an impact, though relatively insignificant, on the representative Browns Ferry Unit 2 Cycle 19 SLMCPR values. The calculations supporting the representative Cycle 19 design implemented the updated ACE correlation provided in the References 15 and 16.The following table presents a comparison of the percentage of rods in boiling transition calculated for the lowest supportable and submitted two-loop operation (TLO) and single-loop operation (SLO) SLMCPRs.

As shown in each instance, the calculated rods in boiling transition is equal to or worse when implementing the updated ACE correlation.

Percent of Rods in Boiling Transition ACE/ATRIUM IOXM NRC-Approved Loop SLMCPR Critical Power Correlation ACE/ATRIUM IOXMConfiguration with Nodal K-Factor Model Critical Power Correlation 1.04 0.0834 0.0820TLO1.06 0.0417 0.03601.05 0.0921 0.0719SLO_______ 1.08 0. 0331 0. 0331AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1 OXM Revision 1Fuel Transition Page 2-65SRXB RAI-2Since Unit 2 Cycle 19 core is expected to include both ATRIUM-10 and ATRIUM-10 XM fueldesigns, which fuel design is more limiting from the standpoint of SLMCPR, and why?AREVA Response:

The SLMCPR is defined as the minimum value of the critical power ratio ensuring less than0. 1% of the fuel rods are expected to experience boiling transition during normal operation, oran abnormal operational occurrence.

A single value for SLMCPR is calculated for all fuel in thecore using the methodology described in Reference

17. One value is calculated for TLO andone is calculated for SLO.The SLMCPR analysis is performed at each cycle exposure with a power distribution thatconservatively represents expected operating states that could both exist at the operating limitMCPR (OLMCPR) and produce a MCPR equal to the SLMCPR during an anticipated operational occurrence.

The limiting fuel design is dependent on what the power distribution isat the cycle exposure being analyzed.

As an example, the representative Browns Ferry Unit 2 Cycle 19 limiting exposure is different for TLO and SLO. The percentages of the total number of fuel rods predicted to experience boiling transition in the overall Monte Carlo statistical evaluation associated with each nuclearfuel type are presented in Table SRXB RAI 2-1.AREVA NP Inc.

AREVA RAI Responses forBrowns Ferry ATRIUM 1OXMFuel Transition ANP-3248NP Revision 1Page 2-66Table SRXB RAI 2-1 Contribution of Total Predicted Rodsin BT by Nuclear Fuel TypeSRXB RAI-3Regarding the calculations performed for Unit 2 Cycle 19 reload safety analysis, as provided inANP-3167(P),

Revision 0, AREVA NP Inc., November 2012 (Attachment 12), confirm that mostrecently approved methodologies were used, including RODEX4, which accounts for thedegradation of thermal conductivity with increasing fuel burnup using upper limit on calculated clad oxide thickness, and the updated methodology for ACE correlation that addresses anonconservatism in the axial averaging process used to determine the K factor, which is aninput to the correlation (as discussed in page 2 of Enclosure to the LAR). If the most recentlyupdated and approved methodologies were not used for BFN Unit 2 Cycle 19 reload safetyanalyses, then please provide justification.

AREVA Response:

The most recently updated and approved methodologies, including RODEX4 and the updatedmethodology for the ACE correlations, were used for the representative Browns Ferry Unit 2Cycle 19 transition analysis.

The methodology reports used in the analysis are provided in thereference section of the reload report, Reference 5.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 3-13.0 References

1. Letter, F. E. Saba (NRC) to J. W. Shea(TVA),

"Browns Ferry Nuclear Plant, Units 1, 2,and 3, Request for Additional Information for Technical Specification Change TS-478Regarding Addition of Analytical Methodologies to TS 5.6.5 for Browns Ferry NuclearPlant Units 1, 2, and 3, and Revision of TS 2.1.1.2 for BFN Unit 2 (TAC NumbersMF0877, MF0878 and MF0879),"

USNRC, August 30, 2013. (38-9211081-001)

2. BAW-1 0247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology forBoiling Water Reactors, AREVA NP Inc., February 2008.3. ANP-3145(P)

Revision 0, Browns Ferry Unit 2 Cycle 19 LAR Fuel Cycle Design, AREVANP, August 2012.4. ANP-2860P Revision 2, Browns Ferry Unit 1-Summary of Responses to Request forAdditional Information, AREVA NP Inc., October 2009.5. ANP-3167(P)

Revision 0, Browns Ferry Unit 2 Cycle 19 Reload Analysis, AREVA NP,November 2012.6. ANP-3152(P)

Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Break SpectrumAnalysis for A TRIUMTM IOXM Fuel, AREVA NP, October 2012.7. ANP-3153(P)

Revision 0, Browns Ferry Units 1, 2, and 3 LOCA-ECCS AnalysisMAPLHGR Limit for A TRIUMTM IOXM Fuel, AREVA NP, October 2012.8. ANP-3170(P)

Revision 0, Evaluation of Fuel Conductivity Degradation for ATRIUMIOXM Fuel for Brown Ferry Units 1, 2 and 3, AREVA NP, November 2012.9. NEDO-32465-A, Licensing Topical Report, "Reactor Stability Detect and SuppressSolutions Licensing Basis Methodology for Reload Applications,"

GE Nuclear Energy,August 1996.10. GE-NE-0000-0028-9714-RO, "Plant-Specific Regional Mode DIVOM Procedure Guideline",

July 14, 2004.11. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.12. Letter, Alan Chung (GE) BWR Owner's Group Detect and Suppress II Committee, OG02-0119-260, "Backup Stability Protection (BSP) for Inoperable Option III Solution,"

July 17, 2002.13. EMF-CC-074(P)(A)

Volume 4 Revision 0, BWR Stability Analysis

-Assessment ofSTAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.14. ANP-2860(P)

Revision 2 Supplement 1P Revision 0, Browns Ferry Unit 1 -Summary ofResponses to Request for Additional Information Extension for ATRIUM I OXM, AREVANP, November 2012.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1Fuel Transition Page 3-215. ANP-1 0298PA Revision 0 Supplement 1 P Revision 0, Improved K-factor Model forACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, December 2011.16. ANP-3140(P)

Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model forACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, August 2012.17. ANP-1 0307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling WaterReactors, AREVA NP, June 2011.18. ANF-89-98(P)(A)

Revision 1 and Supplement 1, Generic Mechanical Design Criteria forBWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.19. XN-NF-75-32(P)(A)

Supplements 1 through 4, Computational Procedure for Evaluating Fuel Rod Bowing, Exxon Nuclear Company, October 1983. (Base document notapproved.)

20. XN-NF-82-06(P)(A)

Supplement 1 Revision 2, Qualification of Exxon Nuclear Fuel forExtended Bumup, Supplement 1, "Extended Burnup Qualification of ENC 9x9 BWRFuel", May 1988.21. "Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal MarginCalculations for Light Water Reactors (Revision 1)", NRC Report dated February 16,1977.22. EMF-95-52(P),

Fuel Design Evaluation for Siemens Power Corporation A TRIUM-1OBWR Reload Fuel, Siemens Power Corporation, December 1998.23. EMF-92-116(P)(A),

Revision 0, Generic Mechanical Design Criteria for PWR FuelDesigns, February 1999.24. ANP-3150P Revision 0, "Mechanical Design Report for Browns Ferry ATRIUM IOXMFuel Assemblies",

AREVA Inc., October 2012.25. ASME Boiler and Pressure Vessel Code,Section III, Division 1, American Society ofMechanical Engineers.

26. EMF-93-177(P)(A),

Revision 1, Mechanical Design for BWR Fuel Channels, August2005.27. EMF-2971 (P), Revision 1, Mechanical and Thermal-Hydraulic Design Report for BrownsFerry Unit 3 Batches BFC-1 and BFC-1A ATRIUM-1O Fuel Assemblies, January 2004.28. EMF-3114(P),

Revision 0, Mechanical and Design Report for Browns Ferry Unit 2 BatchBFE2-14 ATRIUM-10 Fuel Assemblies, September 2004.29. ANP-2537(P),

Revision 0, Mechanical Design Report for Browns Ferry Unit 2 ReloadBFE2-15 ATRIUM- 10 Fuel Assemblies, May 2006.30. XN-81-51 (P)(A), LOCA-Seismic Structural Response of an Exxon Nuclear CompanyBWR Jet Pump Fuel Assembly, Exxon Nuclear Company, May 1986.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1 OXM Revision 1Fuel Transition Page 3-331. XN-NF-84-97(P)(A),

LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet PumpFuel Assembly, Exxon Nuclear Company, August 1986.32. 51-9191258-001, Browns Ferry Unit 2 Cycle 19 MCPR Safety Limit Analysis WithSAFLIM3D Methodology, AREVA NP, October 2012.33. EMF-85-74(P),

Revision 0, Supplement I (P)(A) and Supplement 2(P)(A),

RODEX2A(BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.AREVA NP Inc.

Enclosure 4License Condition Related to Treatment of Channel Bow Uncertainty Amendment NumberXXX (Unit 1)License Condition Implementation DateThe fuel channel bow standard deviation component Upon implementation ofof the channel bow model uncertainty used by Amendment No.ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.11) todetermine the Safety Limit Minimum Critical Power Ratioshall be increased by the ratio of channel fluence gradientto the nearest channel fluence gradient bound of thechannel measurement

database, when applied to channelswith fluence gradients outside the bounds of themeasurement database from which the model uncertainty is determined.

Amendment License Condition NumberImplementation DateXXX (Unit 2)The fuel channel bow standard deviation component Upon implementation ofof the channel bow model uncertainty used by Amendment No.ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.11) todetermine the Safety Limit Minimum Critical Power Ratioshall be increased by the ratio of channel fluence gradientto the nearest channel fluence gradient bound of thechannel measurement

database, when applied to channelswith fluence gradients outside the bounds of themeasurement database from which the model uncertainty is determined.

Amendment License Condition NumberImplementation DateXXX (Unit 3)The fuel channel bow standard deviation component Upon implementation ofof the channel bow model uncertainty used by Amendment No.ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.11) todetermine the Safety Limit Minimum Critical Power Ratioshall be increased by the ratio of channel fluence gradientto the nearest channel fluence gradient bound of thechannel measurement

database, when applied to channelswith fluence gradients outside the bounds of themeasurement database from which the model uncertainty is determined.