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)

Text

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

-Non Proprietary

'

ANP-3248NP Revision1 AREVA RAI Responses for Browns Ferry ATRIUM IOXM Fuel Transition September 2013 A AREVA AREVA NP Inc.

AREVA NP Inc.ANP-3248NP Revision 1 AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition AREVA NP Inc.ANP-3248NP Revision 1 AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition Copyright

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page i Nature of Changes Item Page Description and Justification

1. All The page numbering was incorrect in the Table of Contents and Section 1.0. Minor changes to References 1 and 31. Updated SNPB RAI-17 consistent with Reference

1. No additional changes were made.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 10XM Fuel Transition ANP-3248NP Revision 1 Page ii Contents 1 .0 In tro d u c tio n ..................................................................................................................

1-1 2.0 NRC Questions and AREVA Response ........................................................................

2-1 3 .0 R e fe re n c e s ...................................................................................................................

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

2-31 Sensitivity of Pellet Conductivity on COTRANSA2/XCOBRA-T

..................................................................

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

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

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

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

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

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page iii Figures Figure SNPB RAI 1-1 Figure SNPB RAI 1-2 Figure SNPB RAI 1-3 Figure SNPB RAI 3-1 Figure SNPB RAI 3-2 Figure SNPB RAI 3-3 Figure SNPB RAI 4-1 Figure SNPB RAI 4-2 Figure SNPB RAI 4-3 Figure SNPB RAI 4-4 Figure SNPB RAI 4-5 Figure SNPB RAI 5-1 Figure SNPB RAI 5-2 Figure SNPB RAI 5-3 Figure SNPB RAI 6-1 Figure SNPB RAI 6-2 Figure SNPB RAI 6-3 Figure SNPB RAI 11-1 Figure SNPB RAI 12-1 Figure SNPB RAI 17-1 Figure SNPB RAI 17-2 Lattice Reactivity Comparison at Same Enrichment

.........................

2-4 Isotopic Depletion Variation, BLEU-CGU .........................................

2-5 Fissile Isotope Variation, BLEU-CGU ..............................................

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

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

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

2-13 ATRIUM 10A Standard FUELGUARD

...........................................

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

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

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

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

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

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

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

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

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

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

2-28 Fuel Thermal Conductivity Relative to No Burnup Degradation as a Function of Temperature and E x p o s u re .......................................................................................

2 -4 0 Allowable Transient Overpower Ratio versus Rod N odal Exposure .............................................................................

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

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

2 -5 2 AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 1-1 1.0 Introduction Tennessee Valley Authority (TVA) submitted a License Amendment Request (LAR) to change the Browns Ferry Technical Specifications in support of reload fuel transition to ATRIUM T M 1OXM *. In response to the LAR, the US Nuclear Regulatory Commission (NRC) has issued an initial 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 NRC RAIs.* ATRIUM is a trademark of AREVA NP.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-1 2.0 NRC Questions and AREVA Response The NRC questions (i.e., RAIs) listed below are according to Reference 1: SNPB RAI-1 ANP-3159P, Section 1.0 It has been stated in your LAR submittals that TVA intends to continue use of blended low enriched 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, list other differences in the BLEU fuel such as isotopic composition, physical properties, and neutronics characteristics from the commercial grade fuel.AREVA Response: The primary difference between BLEU and commercial grade uranium (CGU) is the concentration of the uranium isotopes of U 2 3 4 and U236. BLEU material has a higher concentration of these isotopes when compared to the maximum allowed values for enriched CGU defined by ASTM C966-10. Chemically, there is no difference between BLEU and CGU.Within the fuel manufacturing process, the U 2 3 4 and U 2 3 6 isotopes are inseparable from its original BLEU feed stock.Both CGU and BLEU material is subject to the same maximum U 2 3 5 enrichment of 4.95%. The following table provides a CGU versus BLEU comparison of U 2 3' and U 2 3 6 concentrations.

The CGU allowable values are from the ASTM C966-10 specification with the U 2 3 4 equivalent weight percent based upon the maximum allowable U 2 3 5 enrichment.

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

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

• Weight %U 2 3 4 1. 1OE+04 pg1gU 2 3 5 0.0546 wt% U 0.09 wt% U U 2 3 6 250 pg/gU 0.025 wt% U 1.60 wt% U* For material enriched to 4.95% U 2 3 5.AREVA NP Inc.

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

The commercial grade natural uranium used in the blending process contains 0.0057 wt% U 2 3 4 and no U 2 3 6.The small changes in isotopic impurities of the BLEU fuel do not significantly affect the physical properties of the fuel. The physical properties for U0 2 and U0 2-Gd 2 O 3 fuel are identified in the RODEX4 topical report (Reference

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

Isotopes of uranium (e.g., U 2 3 4 , U 2 3 5 , U 2 3 6 and U 2 3 8) have the same electronic structure.

They also occupy the same space. Consequently, the substitution of a U 2 3 4 or U 2 3 1 6 for a U 2 3 8 (or U 2 3 5)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 U 2 3 4 and U 2 3 6 content as it is also independent of the amount of U 2 3 5.As discussed in the response to (b) below, the impact on the fuel isotopic composition during depletion is small. Since the fissile isotopic inventory does not significantly deviate from normal expected variations the corresponding changes in fission products are insignificant.

Because the 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 U 2 3 4 and U 2 3 6 , other thermal mechanical properties are also not affected.

This includes thermal expansion, heat capacity, enthalpy, Young's modulus, Poisson's ratio, creep, melting temperature and emissivity.

The fuel density is slightly less by an insignificant amount. As a result, the physical properties 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 U 2 3 6.The U 2 3 6 isotope has neutron poisoning impact. For the BLELI assemblies that have a combination of BLEU and CGU rods the net effect is a reduction of reactivity approximately equivalent to a 0. 3% reduction in U 2 3 5 enrichment.

In other words, the enrichment of a BLEU assembly would need to be approximately 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 1 Fuel Transition Page 2-3 since both assemblies are limited to the same maximum enrichment value the impact is usually seen 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 lattices are compared.

In this comparison, the lattices are identical except the additional U 2 3 4 and U 2 3 6 was removed for the CGU lattice (i.e., both lattices have exactly the same U 2 3 5 enrichment and gadolinia distributions).(b) If the isotopic content of the BLEU fuel is different from that of the commercial grade fuel, 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 U 2 3 4 and U 2 3 6 with cross-section data for a range of temperatures and voids. The behavior of these uranium isotopes under 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 U 2 3 4 and U 2 3 6 isotopic concentrations in the fuel design and licensing process. Design changes to address the presence of the higher concentration of the U 2 3 6 include increasing the reload batch size, modifying lattice enrichments and gadolinium loading.Figure SNPB RAI 1-2 illustrates the difference in the isotopic buildup and depletion for the same comparison lattices used in the previous reactivity comparison.

At beginning of life the primary differences are in the U 2 1 6 and U 2 3 1 isotopes.

The U 2 3 8 difference is simply a compensating reduction at BOL due to the inclusion of the U 2 3 4 and U 2 3 6 isotopes.

Focusing on the significant fissile isotopes, Figure SNPB RAI 1-3 shows a slight increase in fissile inventory with exposure for the BLEU lattice. This is due to the neutron poisoning effect due to the presence of the U 2 3 6 and the conversion of U 2 3 4 to U235. These changes in fissile isotope quantities do not significantly differ from the variations normally experienced in a reload due to changes in fuel enrichment and gadolinia loadings, or operation.

AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-4 The BLEU material has been used in 10 reloads for the Browns Ferry units with the first reload in Unit 2 Cycle 14.1.2 1.1 C C C M 1.0 0.9 0.8 0.7 0 10 20 30 40 GWd/MTU 50 60 70 Figure SNPB RAI 1-1 Lattice Reactivity Comparison at Same Enrichment AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-5 15 10 5 0-5 EU'I-C I----------10-15 0 10 20 30 40 50 60 70 GWd/MTU--U-234----U-235--!!-U-236 -aU-238 Figure SNPB RAI 1-2 Isotopic Depletion Variation, BLEU-CGU 4-C am 0.5 0.4 0.3 0.2 0.1 0-0.1-0.2-0.3-0.4-0.5~ w u- .,amp 0 10 20 30 GWd/MTU 40 50 60 70--U-235--m-- PU-239-,4- PU-241 Figure SNPB RAI 1-3 Fissile Isotope Variation, BLEU-CGU AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-6 SNPB 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 the statistical analysis.AREVA Response: The uncertainty in the calculated channel bow leads to an associated uncertainty in the fuel rod power 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 bow and 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 for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-7 f I The 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 the adjustments 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 and oxidation 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 1 Fuel Transition Page 2-8 AREVA Response: TVA instituted a healthy fuel examination program in the mid 2000s, as part of establishing a baseline for fuel performance at Browns Ferry Nuclear. The fuel inspections supporting this program are performed by the fuel vendor. The scope typically involves the inspection of one assembly from each batch following completion of an operating cycle. The highest exposure once burnt bundle is held out for inspection; a high exposure twice burnt bundle, and a high exposure thrice burnt bundle, are also inspected.

The inspection typically includes a peripheral examination of the bundle with the channel removed, to assess general performance and ensure 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 plus tenacious crud) are taken on each removed fuel rod, along with profilometry measurements and eddy current testing for flaws. The scope of the inspections in the Browns Ferry Nuclear healthy fuel inspection program exceeds the guidance for post irradiation surveillance discussed in section 4.2 of the NUGEG-0800 Standard Review Plan. The scope of the program has been expanded with the introduction of On Line Noble Chemistry (OLNC).TVA implemented OLNC at Browns Ferry Unit 3 beginning in Cycle 15. At the time TVA instituted the OLNC injection there was no operating experience of AREVA fuel with OLNC, and subsequently, TVA and AREVA initiated a long term program to validate that OLNC does not result in a decrease in fuel reliability for AREVA fuel. The initial step in this program was to establish 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 and channel was performed.

Detailed examinations were performed for six fuel rods from each fuel assembly, 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 two locations-one sample with a brush and one with a blade, for a total of eight samples. These samples were then analyzed by AREVA to characterize the crud deposits for Browns Ferry Unit 3 prior to the initiation of the OLNC program. Specifically:

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-9" Bulk chemical analysis of selected fuel deposit particulate samples. These samples were analyzed for elemental and metallic oxide distribution and possibly for the determination of crystal grain size.* Fuel deposit flakes were selected for evaluation of the OD and ID features of the deposits, 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, to benchmark the AREVA crud and corrosion risk assessment tool for Browns Ferry Unit 3 specific conditions and performed a detailed risk assessment for Cycle 15.TVA and AREVA performed a similar post-irradiation exam following Cycle 15-once, twice and thrice 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 risk assessment.

Exam and crud analyses are planned following Browns Ferry Unit 3 Cycles 16 and 17 to complete this program.Unit 3 is the only Browns Ferry Unit currently using OLNC. Water chemistry assessments will be performed to determine the applicability of the Unit 3 results to the other Browns Ferry Units when OLNC is implemented for those Units.(b) Provide details as to how the guidance on treatment of corrosion, crud, and hydrogen content per NUREG-0800 Standard Review Plan (SRP), Section 4.2 is satisfied, and AREVA Response: Fuel rod corrosion is addressed by following the approved RODEX4 methodology described in Reference

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, the water 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 into account in the fuel rod analyses, as required.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-10 The RODEX4 fuel rod analysis methodology makes use of a statistical approach that involves sampling 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 the inputs. The results are treated to demonstrate that a large fraction of the rods [] U0 2 rods from the equilibrium cycle batch are shown in Figure SNPB RAI 3-1 below.Figure SNPB RAI 3-1 Browns Ferry Equilibrium Cycle RODEX4 Maximum Corrosion Results AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-11 The 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 RAI responses to the RODEX4 topical report, it is stated that the existing corrosion model includes a design basis level of crud. That is, the liftoff measurement data used to benchmark the RODEX4 corrosion model include normal, low levels of crud for the plants represented by the measurement data. If the plant-specific measurements indicate abnormal crud levels, the analyses for that plant must take into account a design basis crud thickness that can be derived from the plant-specific data.In the case of the Browns Ferry plants, plant-specific liftoff data are available from the fuel surveillance program described in the response to (a) above. Figure SNPB RAI 3-2 shows recent 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 at the end of cycle 16).Figure SNPB RAI 3-2 Liftoff Measurement Data on AREVA ATRIUM-10 Fuel at the Browns Ferry Units AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-12 The 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 the Browns Ferry units with the intent to obtain liftoff values that conservatively represent the reload batches.Also shown in the figure are the data used for the RODEX4 corrosion model benchmark and model uncertainty evaluation.

The data are identified in the legend with reactor codes and date of examination (e.g., All, 2001) and are from AREVA BWR l0x10 fuel. The Browns Ferry liftoff data 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 crud conditions at the Browns Ferry plants.As part of the RODEX4 methodology, the approved limit for corrosion is [ ]. During the first reload application of RODEX4, the limit was challenged by the NRC because of a concern about the effect of spallation on the cladding integrity.

Spallation can produce localized surface discontinuities in the cladding and may also result in the formation of hydride lenses that could cause 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 on AREVA BWR fuel. Figure SNPB RAI 3-3 displays the data.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-13 Figure SNPB RAI 3-3 Zircaloy-2 Stress Relieved Cladding Oxide, Historical Liftoff Measurement Data The maximum measured value from the data was []The maximum calculated oxide value off]. If higher liftoff data are encountered as part of the water chemistry program at the Browns Ferry units, the crud inputs to the RODEX4 analyses will be adjusted as already required by the SER restriction on the 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 there an approved hydrogen concentration limit. Consistent with a preceding Technical Specification AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-14 revision that was updated and approved to include the RODEX4 methodology (Reference 3 in ANP-3159P), the application of the RODEX4 methodology with the f J which is a conservative method to ensure adequate cladding strength and ductility are maintained, thus satisfying the guidance in SRP 4.2.(c) In ANP- 3159P Section 3.2.7, it is stated, in part, that as a result of concerns that were raised on the effect of non-uniform corrosion, such as spallation, and localized hydride formations on the ductility limit on cladding, a regulatory commitment was made to reduce the limit oxide limit to the value in Reference 3 that is listed for ANP-3159P.

Provide details of how this regulatory commitment to reduce the oxide limit to the value specified in Reference 3 of ANP-3159P is implemented at BFN Units 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 the results are summarized in ANP-3082P.

The transition cores for the Browns Ferry units consist of ATRIUM 10 with both Standard FUELGUARD (SFG) and Improved FUELGUARD (IFG) lower tie plates. Provide a detailed description of the differences between SFG and IFG with respect to 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 holes and 2 with drain holes), and 8 grid rods. Blades are assembled in slots and grid rods are thereafter inserted and brazed together.

See Figure SNPB RAI 4-1 hereafter.

AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-15 ex PLFR BUSHIN WATER CHANNE BUSH I NG GRID?X CURVED BLADE WITH-' \-CURVED BLADE W/O DRAIN HOLES DRAIN HOLES Figure SNPB RAI 4-1 ATRIUM IOA Standard FUELGUARD The Improved FUELGUARD is similar to the Standard FUELGUARD; however there are 34 half interstitial 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 Figure SNPB RAI 4-3.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-16$a CUVE BLADE DRAIN HOLES SIX URVED BLADE WflH DRAIN HOLES Figure SNPB RAI 4-2 ATRIUM IOA Improved FUELGUARD Bottom View INTERSTITIAL Figure SNPB RAI 4-3 ATRIUM IOA Improved FUELGUARD Side View Note: The difference in orientation is based on third vs. first angle projection.

AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-17 For clarity, Figure SNPB RAI 4-4 and Figure SNPB RAI 4-5 are included below to illustrate the ATRIUM IOXM Improved FUELGUARD depiction, the difference being the PLFR Bushing quality and locations.

CURVED BLADE WITH DRAIN HOLES INTERSTITIAL STRIP-MIX 1-OH WATER CHANNEL BUSHING ROD'\-N"LFR BUSHING-URVED BLADE WITHOUT DRAIN HOLES Figure SNPB RAI 4-4 ATRIUM IOXM Improved FUELGUARD Bottom View AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-18 Figure SNPB RAI 4-5 ATRIUM IOXM Improved FUELGUARD Side View Note: 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 core evaluations shown in Tables 3.9 and 3. 10 of ANP-3082P.

The core pressure drop at rated conditions for a full core of ATRIUM- 10 with SFG is [ ] psid and a full core with IFG is I ] psid. The core pressure drop at off-rated conditions for a full core of ATRIUM-10 with SFG is [ ] psid and a full core with IFG is [ ] psid. These differences between the two LTPs are not significant on the core pressure drop.The impact on thermal margin performance between the SFG and IFG LTP is apparent when the 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 results provided for "Transition Core Loading 2"presents the critical power performance for the same core conditions applied to a SFG LTP assembly and IFG LTP assembly.

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-19 SNPB RAI-5 ANP-3082P Section 3.4. ANP-3150P, Section 3.3.5 It is stated in the above section that the rod closure due to rod bow was assessed for impact on thermal 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 open literature 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 CPR thermal margins, is given in 5(c).(b) Provide an assessment of how the thermal margin calculations are affected by the rod bow 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 CPR thermal margins, is given in 5(c).AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-20 Figure SNPB RAI 5-1 ATRIUM-10 Rod Bow MCPR Penalty To assure that this model is conservative, AREVA ran a CHF test on an ATRIUM-10 bundle in which two rods were welded together.

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-21 Figure 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 by showing typical analysis/calculations, and provide how the rod bow behavior is impacted by 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 the I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1 OXM Fuel Transition ANP-3248NP Revision 1 Page 2-22 I I Based on the correlation above the [J primary factors impacting rod bow are f I The ATRIUM IOXM [] Figure SNPB RAI 5-3 below compares the predicted rod bow for both the ATRIUM-10 and the ATRIUM 1OXMf]AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-23 Figure SNPB RAI 5-3 ATRIUM IOXM and A TRIUM-IO 95/95 % gap closure SNPB RAI-6 ANP-10307PA, Section 2.2. Channel Bow It has been stated in Section 2.2.1 the channel growth correlation used to determine the channel bow magnitude is a continuous function of fast fluence from the beginning to the end of life. The model 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 bow database is applicable to Browns Ferry units' operating conditions.

AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-24 AREVA Response: The calculated fast fluence gradients for the core design provided in Reference 3 have been compared to the upper and lower bounds of the channel bow database.

As shown in the table below and Figure SNPB RAI 6-1, the Browns Ferry Unit 2 Cycle 19 reference core design remains 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 the lower 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 of the 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 the increased core leakage at the periphery which can result in increased fluence gradients.

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

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

This is illustrated in Figure SNPB RAI 6-3 in which the FLCPR (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 is insignificant due to the margin exhibited by the assemblies expected to exceed the bounds of the channel fast fluence database.

However, as noted in the response to part (b) below, it is proposed 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 by applying the proposed license condition.

In this case, an augmented uncertainty was applied to the four bundles identified as exceeding the lower bound of the database.L MCPR calculations using SAFLIM3D with the augmented channel bow uncertainty for the affected assemblies were performed on a consistent basis as analysis using the standard channel bow uncertainty.

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

and the application of the license condition.

Comparison of the results show that the impact of applying the augmented channel bow uncertainty to the assemblies of interest is not significant.

Percent of rods in boiling transition Loop SLMCPR SLMCPR results without SLMCPR results with Configuration proposed license condition proposed license condition 1.04 0.0834 0.0834 TLO 1.06 0.0417 0.0403 1.05 0.0921 0.0849 SLO 1.08 0.0331 0.0316 Small 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 for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-26 assemblies are affected the most by the increased channel bow uncertainty.

The effect on the power distribution for assemblies further away is greatly diminished

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-27 Figure SNPB RAI 6-2 Browns Ferry Unit 2 Cycle 19 Reference Loading Pattern AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 10XM Fuel Transition ANP-3248NP Revision 1 Page 2-28 c.LL 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2*~ ~ ~ ~ ~~~o =N4 2 maa** ~l u*ij 0 5 10 Cycle Exposure (GWd/MTU)15 20-Core Limiting

• FBD258 A FBD262 x FBD266 .FBD270 Figure SNPB RAI 6-3 Browns Ferry Unit 2 Cycle 19 FLCPR for Assemblies 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 of BSEP's expected performance.

To resolve this issue, the licensee for BSEP agreed to increase the channel bow uncertainty in the SLMCPR calculation for the most severely deflected fuel channels.

In view of the excessive channel bow that occurred at BSEP a license condition was proposed for BSEP Units 1 and 2 in connection with the use of AREVA 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, January 22, 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 1 Fuel Transition Page 2-29 SNPB RAI-7 ANP-3082P.

Section 3.5 Discuss the impact on bypass exit subcooling while transitioning between transition core combination 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 on maintaining bypass flow fractions (refer to Section 4.1.5 of Reference 18). Fuel designs are considered to be hydraulically compatible when the bypass flow characteristics of the reload fuel assemblies do not differ significantly from the existing fuel in order to provide adequate flow in the bypass region. If needed to achieve similar bypass flow fractions between fuel designs, the bypass 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 multiple transition 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". The core 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 results in[ ]SNPB RAI-8 ANP-3082P.

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-30 AREVA Response: Results presented for 62%P/ 37.3%F in Table 3.10 of ANP-3082(P) provide comparisons of the calculated critical power ratio performance of the GE14, A TRIUM-10, and ATRIUM IOXM fuel assemblies 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 as indicated in Figure 1.1, BFN Power Flow Map -100 percent original licensed thermal power (OLTP) of ANP-3167(P), BFN Unit 2 Reload Safety Analysis.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-31 AREVA 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 will not change this conclusion.

As an example, Table SNPB RAI 8-1 provides results originally performed for the initial transition 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 the statepoints; 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 Power I L Assembly Flow (klbm/hr)Statepoint GE14 ATRIUM-10

% Difference Percent Assembly Bypass Flow Statepoint GE14 ATRIUM-10 Difference I SNPB RAI-9 ANP-3150P.

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

AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-32 AREVA Response: As discussed in ANP-3150P Section 3.3.1 (Reference 24), [ ] the fuel assembly structural components do not receive significant loads during normal and AO0 conditions.

f J No analyses are performed to confirm design margin under normal operating and AO0 conditions f I To ensure the structural integrity off I Section III of the ASME Boiler and Pressure Vessel code (Reference

25) is used to establish acceptable design limits. To evaluate the stresses under normal operating conditions, [J The maximum normal operation

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

To evaluate the stress under AOO and accident conditions, [I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-33 I I For the [] the normal operating stresses [] The design margin is confirmed by comparing the resulting stress to the design limit as defined by Section Ill of the ASME Boiler and Pressure Vessel code (Reference 25).RAI-1O provides additional details on how [ ] at BFE2-19 were calculated.

Additional information on the stress evaluation results and comparison 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.

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

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-34 AREVA Response: A plant specific analysis was not performed because it was not required.

A change in fuel assembly design may not necessitate a full reanalysis if it can be shown that the fuel design is dynamically similar to the fuel assembly design in the Reactor Pressure Vessel (RPV) seismic analysis of record (ANF-89-98(P)(A), Reference 18, Section 3.2.7). A comparison between fuels []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 [ I and for Safe-Shutdown Earthquake (SSE) this value is [ ] The SSE acceleration can be applied to the A TRIUM-IC and ATRIUM IOXM designs iff I.The first reload of the ATRIUM-IC, documented in EMF-2971(P)

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

27) and follow on reloads (References 28 and 29) referenced in BFN 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 the same 100/75-mil Advanced Fuel Channel (AFC) as the ATRIUM-IC design, with a fuel assembly [ J. This results in a [] accepted designs at Browns AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-35 Ferry. 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, the ATRIUM-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 been used in both PWR and BWR evaluations and is documented in the RAI responses and SER of EMF-92-116(P)(A), Revision 0 (Reference 23).Since the ATRIUM IOXM is found to be dynamically similar to the analyzed fuel of record the margin to results is found by comparing the fuel channel acceleration limit off Ito the f J. This provides a AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-36 The fuel assembly analysis design criteria are established in ANF-89-98(P)(A) (Reference 18).The [ ] is used as the design input to a static finite 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 f j For added conservatism the FEA static analysis model assumes I.The ATRIUM- 10 fuel assembly component analysis t 1. The allowable stress or load limits for the ATRIUM IOXM were updated to new limits based on testing 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 and the allowable acceleration.(b) Provide details of the testing done to obtain the dynamic characteristics of the fuel assembly and spacer grids under varying conditions of stiffness, natural frequencies and damping values with and without the fuel channel. Provide details of the evaluation of BFN ATRIUM 10/ATRIUM 1OXM fuel assembly structural response to externally applied forces (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 1 Fuel Transition Page 2-37 AREVA 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 the ATRIUM 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 a channeled and un-channeled assembly and were used for[]. The full array of testing conducted 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 assembly acceptance criteria per the Standard Review Plan are listed in Reference 18 and Reference 24 using the same numbered criteria sections; Table 7.3 in Reference 18 and in Table 3-1, 3-2 and 3-3 in Reference

24. Acceptance criteria of Section IV, Appendix A of the SRP Chapter 4.2 are addressed in Table 3-1, Criteria Section 3.4.4 of Reference 24.SNPB RAI-1I ANP-3170P Section 3.1 The core average gap conductance used in COTRANSA2 system calculations and the hot channel gap conductance used in XCOBRA-T hot channel calculations are obtained from RODEX2 calculations.

The sensitivity to conductivity and gap conductance for Anticipated Operational Occurrence (AOO) analyses is in the opposite directions for the core and the hot channel. This means that putting more energy into the coolant (higher thermal conductivity/higher gap conductance) is nonconservative for the system calculation but conservative for the hot channel calculations.

Provide, with quantitative examples, how these competing effects between the core and hot channel calculations are balanced to minimize the overall impact of thermal conductivity degradation.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-38 AREVA 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 time constant.

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 the fluid response to the changing neutron power. For a limiting pressurization event, the lower conductivity results in an increase in reactor power due to the lag in void formation that would otherwise mitigate the power rise. This increase in transient reactor power results in a larger reduction of thermal margin during the event; therefore, the lack of modeling TCD in the system is non-conservative.

However, an increase in the rod thermal time constant causes a hold up of heat 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 heat flux 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 Table SNPB RAI 11-1. In the table, base refers to unaltered pellet conductivity.

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

As noted, the reduction in pellet conductivity for both the system and hot channel tend to minimize the overall impact of thermal conductivity degradation; for this specific example, the thermal margins slightly decreased when degraded conductivity was utilized.

This overall trend is consistent with the discussion in AREVA NP Inc.

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

Table SNPB RAI 11-1 Sensitivity of Pellet Conductivity on COTRANSA2/XCOBRA-T COTRANSA2 XCOBRA-T Change in ACPR ACPR from Pellet Conductivity Pellet Conductivity Original Case AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-40 Figure SNPB RAI 11-1 Fuel Thermal Conductivity Relative to No Burnup Degradation as a Function of Temperature and Exposure SNPB RAI-12 ANP-3145P, ANP-2637 Section 3.0 Nuclear 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 1 Fuel Transition Page 2-41 (a) Provide the details of the FDL and PAPT limits as a function of exposure and show that sufficient margin exists between the steady state and transient LHGR limits, AREVA Response: The RODEX4 methodology used for the ATRIUM T M IOXM fuel differs significantly from the previous 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 design limit (FDL) itself is established to ensure that the thermal-mechanical criteria are met using power histories from expected operation but generated assuming an anticipated operational occurrence (AO0) occurs during the life of the assembly.

Specifically, transient responses for the quasi steady-state events off] The RODEX4 licensing topical report (Reference

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

The transient evaluation approach is also described in some 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 flow dependent multipliers (or set-downs) to the LHGR limits where needed. These power and flow dependent multipliers are identified as LHGRFACp and LHGRFACf, respectively.

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

The required LHGR limits and associated set-downs are provided in the cycle specific Reload Analysis 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 1 Fuel Transition Page 2-42 LHGR limits), 8.5 (LHGRFACp), and 8.6 (LHGRFACf) for the reference cycle provided for the LAR (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 fast transients.

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 for cladding strain (1.0% strain limit) and fuel overheating (fuel melt limit). A summary of the analysis 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 slow or 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 power calculated 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 slow transients would be a CRWE (control rod withdrawal error), flow runup, or LOFWH (loss of feedwater heater).Fast transients, by nature, occur over a very short period of time with potentially high neutron flux levels. These fast transients are typically caused by a pressurization event and are evaluated 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.

Slow transients, 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 derive flow-dependent multipliers on the LHGR limit to protect the cladding transient strain and fuel temperature criteria from anticipated transients occurring during operation.

AREVA NP Inc.

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-44 Figure SNPB RAI 12-1 Allowable Transient Overpower Ratio versus Rod Nodal Exposure In 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 fuel rod includes a description of fuel operation at the applicable FDL LHGR immediately prior to the fast 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 is performed at various exposure increments.

A series of[ I AREVA NP Inc.

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

The analysis performed per this methodology is used to establish power-dependent limits on the maximum LHGR that can be allowed for the fuel design and operating cycle of interest.As indicated in part (a) above, the combination of the slow transient allowable overpower ratio and the fast transient methodology are used to develop the LHGRFACp and LHGRFACf multipliers 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 a postulated AOO.(c) Confirm that the fuel Thermal Conductivity Degradation (TCD) with exposure has been taken into account in generating/adjusting the LHGR limits.AREVA Response: The FDL and LHGR multipliers are evaluated using RODEX4 following the approved methodology.

RODEX4 takes into account thermal conductivity degradation.

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

Revision 0, Nuclear Fuel Design Report BFE-19 LAR ATRIUM 1OXM, August 2012. This report is not available to the NRC staff. Please submit a copy of this report to the NRC or provide a complete description 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 1 Fuel Transition Page 2-46 SNPB RAI-13 ANP-3148P.

Section 3.3 Describe 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 of the Browns Ferry units.AREVA Response: TVA to provide response.SNPB RAI-14 ANP-3140, Section 3 In BWR terminology, assembly critical power is defined as the minimum assembly power that results in onset of Boiling Transition (BT) (dryout) at any location in the assembly.

The acceptance criterion related to BT in BWRs is that the limiting value of Minimum Critical Power Ratio (MCPR) such that at least 99.9 percent of the fuel rods in the core are not expected to experience BT during normal operation and AOOs. The SLMCPR is determined such that at least 99.9 percent of the fuel rods in the core are expected to avoid BT if the MCPR is greater than or equal to the SLMCPR. The SLMCPR methodology uses a critical power correlation to calculate 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 local effects such as spacing and geometry) is determined based on predictions of the critical power correlation and comparisons to test data. Figure 3-2 of ANP-3140P identifies the test bundle where majority of the rods were dryout was observed.

Explain how the test results (with majority rods in dryout) are used in the accurate determination of CPR correlation 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 had no indications of dryout. If one examines a single test, [ J are peaked and it is common to observe only a single rod indicate dryout. As described in section 3.3. 1, an experimental additive constant is determined for each of the rods as the value that causes the critical power correlation to exactly predict the critical power for that measurement.

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

In the case AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-47 that the rod was not observed to dryout in that measurement, the experimental additive constant is conservative (it is assumed in the computation of the experimental additive constant that dryout 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 observed changes in additive constants are generally small. However, a random check by staff on the changes in additive constants has shown that the changes vary from 2 percent to 80 percent in magnitude.

Please explain why there are larger variations contrary to what is stated in Section 3.3.4.AREVA Response: Because additive constants have both positive and negative values, percent differences can be quite exaggerated when the base value of the additive constant is near zero. If a difference in additive constant is calculated for each rod position (revised value -original value), the difference in additive constant ranges from f 1, with an average magnitude of change oft ].[] Thus, the conclusion is that the observed changes in additive constant are generally small.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-48 SNPB RAI-15 ANP-3140, Section 5.0 Provide 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 report and 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.

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

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-49 I I SNPB RAI-16 ANP-3152, ANP-3167P While reviewing ANP-3167P, Browns Ferry Unit 2 Cycle 19 Reload Analysis, the NRC staff noticed that the Reference 7 listed in Section 9.0 of ANP-3167P, ANP-3153(p)

Browns Ferry Units 1, 2 and 3 LOCA-ECCS (emergency core cooling system) Analysis MAPLHGR Limit for ATRIUM 1OXM Fuel, is NOT included in your LAR package. Submit this reference to the NRC for staffs review and evaluation of LOCA-ECCS analysis with the MAPLHGR limits for the ATRIUM 1OXM fuel design.AREVA Response: TVA to provide the requested report.SNPB RAI-17 ANP-3152P Section 2.0 Provide detailed summary results of the analyses that are referred to in the second and third sentences of page 2-2.AREVA Response: I I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-50 I] The results support the statement

[I Table SNPB RAI 17-1 [I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM IOXM Fuel Transition ANP-3248NP Revision 1 Page 2-51 Table SNPB RAI 17-2 [I Figure SNPB RAI 17-1 [I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-52 Figure SNPB RAI 17-2 [I SNPB RAI-18 ANP-3152P, ANP-3170P RODEX4 computer code was used to assess the potential impact of exposure (burnup)degradation of U0 2 thermal conductivity (TCD) on the fuel parameters and this assessment was used to make adjustments to input for RODEX2 TCD with exposure that was used in the LOCA analysis.(a) Provide details of the adjustments made to RODEX2 input for LOCA analysis.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-53 AREVA 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, the limiting PCT was not impacted since it occurred at beginning of life (BOL) where TCD is not present.The approach taken to address TCD at the time of the analyses was an interim process. The impact 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 were determined from the RODEX4 runs and then applied to the RODEX2 output that is used in HUXY. Specifically, the RODEX2 output parameters modified by the ratios were: Specifically for Browns Ferry, the ATRIUM IOXM LOCA analyses of Reference 8 were repeated with the new analysis process for evaluating the effects of TCD for comparison.

The new AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-54 results 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 the cladding 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),[I AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-55 Table SNPB RAI 18-1 Impact of Thermal Conductivity Degradation on Browns Ferry ATRIUM IOXM LOCA Analysis Results Temperature Temperature (OF) Difference (OF)Average Planar Exposure (GWd/MTU)PCT Without TCD PCT With TCD Increase in PCT Margin to due to Limiting TCD PCT SNPB RAI-19 ANP-3152 Figure 6.19 of ANP-3152P illustrates the variation of cladding temperature with time for the limiting recirculation line break. The cladding temperature experiences the first peak at around 190 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 vessel during the blowdown period. As shown in Table 6.2 and Figures 6.1 and 6.3 of Reference 6, the ADS valves depressurize the vessel at approximately 190 seconds. This occurs before the end of blowdown/time of rated core spray. This depressurization causes a lower plenum flash through 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 1 Fuel Transition Page 2-56 Following the flash, the node of interest returns to a quality of 1, the heatup then continues until PCT is reached.SNPB RAI-20 ANP-3152.

Section 8.0 Explain 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 either loop, 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 to a similar break occurring during TLO. This occurs because there will be an immediate loss of jet 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 heat transfer will result in higher stored energy in the fuel rods at the start of heatup. Applying an SLO multiplier to the MAPLHGR limits can reduce the increased severity of an SLO LOCA.f] To achieve this goal, a SLO multiplier of 0.85 was chosen. This resulted in a SLO PCTof 1747 0 F, compared to a TLO PCT of 1909 OF. Since SLO is at reduce power and flow, operation with a reduced MAPLHGR limit is not overly restrictive.

SNPB RAI-21 ANP-3152 Explain the impact on LOCA (EGOS performance) analysis and Title 10 of Code of Federal Regulations, 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 heaters out-of-service (FHOOS).AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-57 AREVA Response: Coastdown with FFWTR or FHOOS does not create a limiting condition for the LOCA analyses.I] Therefore, the combination of coastdown with FFTR or FHOOS is not a limiting condition for 10 CFR 50.46 acceptance criteria.SNPB RAI-22 ANP-3167P.

Section 4.2 BFN Unit 2 Cycle 19 SLMCPR calculated from Reference 12 (ANP-10307P) resulted in a value of 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 1.08 for SLO. Provide basis for the adoption of these new values.AREVA Response: TVA to provide response.SNPB RAI-23 ANP-3167P.

Section 4.3 Provide a summary of the analysis and results for BFN units operation with FFTR and FHOOS that 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 BFN operation with backup stability protection regions, if the Option III OPRM system is declared inoperable.

AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-58 AREVA Response: BFN is licensed for the FFTR and FHOOS modes of operation, with a reduction in final feedwater temperature of 55 degrees F at the current licensed power of 3458 MW. The use of reduced feedwater temperature was included as an assumption in the engineering evaluation performed in support of the power uprate program (which increased the licensed power level from 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 request was made as Supplement 1 to Technical Specification change TS-431, and was submitted on September 22, 2006 (ADAMS Accession:

ML062680459).

The use of FFTR and FHOOS was also considered in this uprate request, and the use of reduced feedwater temperature was specifically discussed within the associated NRC SER approving this uprate request (ADAMS Accession:

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 of Reference

5. This calculation was performed in accordance with the requirements of the NRC approved licensing topical report NEDO-32465-A, (Reference 9), with the addition of the calculation 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 topical report 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 in feedwater temperature.

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

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

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

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

f]AREVA NP Inc.

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

5. The results provided 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 when operating 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 maximum calculated decay ratios for the scram and controlled entry regions defined in Table 4.4 of Reference 5.SNPB RAI-24 ANP-3167P, Section 5.1.3 For 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 sharp reduction in relative steam flow at about 15 seconds in to the event. Please describe the cause of this behavior during the FWCF event.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-61 AREVA Response: Due to a failure of the feedwater control system to maximum demand, there is a continual rise in vessel water level that eventually reaches the high water level trip setpoint.

The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam lines. Valve closure creates a compression wave traveling back to the core, causing void collapse and subsequent rapid power excursion.

In this example, the valve closure occurs just prior to 15 seconds. The closure of the turbine stop valves also initiates a reactor scram and a recirculation pump trip. The reactor scram results in a loss of steam production and the sharp decrease in steam flow occurs.SNPB RAI-25 ANP-3167P, Section 6.1 Section 6.1 reports that ATRIUM 1OXM LOCA analysis for BFN Units PCT is 1903 OF and the peak local metal water reaction is 1.16 percent. However, ANP-3152P, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel, Table 6.1 reports that the PCT is 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 obtained from the MAPLHGR report (Reference 7).SNPB RAI-26 ANP-3172P.

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

Please specify whether the pressure results AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 2-62 include any other analysis/measurement uncertainties in addition to the 7 pounds per square inch 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 other analysis/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 2 Cycle 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 for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-63 Table SNPB RAI 26-1 Browns Ferry Unit 2 Cycle 19 Overpressurization Biases and Results.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-64 RAIs by Reactor Systems Branch (SRXB)SRXB RAI-1 In 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 K factor, which is an input to the correlation

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

Explain if updated 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 IOXM Configuration with Nodal K-Factor Model Critical Power Correlation 1.04 0.0834 0.0820 TLO 1.06 0.0417 0.0360 1.05 0.0921 0.0719 SLO_______ 1.08 0. 0331 0. 0331 AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1 OXM Revision 1 Fuel Transition Page 2-65 SRXB RAI-2 Since Unit 2 Cycle 19 core is expected to include both ATRIUM-10 and ATRIUM-10 XM fuel designs, 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 than 0. 1% of the fuel rods are expected to experience boiling transition during normal operation, or an abnormal operational occurrence.

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

17. One value is calculated for TLO and one is calculated for SLO.The SLMCPR analysis is performed at each cycle exposure with a power distribution that conservatively represents expected operating states that could both exist at the operating limit MCPR (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 is at 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 nuclear fuel type are presented in Table SRXB RAI 2-1.AREVA NP Inc.

AREVA RAI Responses for Browns Ferry ATRIUM 1OXM Fuel Transition ANP-3248NP Revision 1 Page 2-66 Table SRXB RAI 2-1 Contribution of Total Predicted Rods in BT by Nuclear Fuel Type SRXB RAI-3 Regarding the calculations performed for Unit 2 Cycle 19 reload safety analysis, as provided in ANP-3167(P), Revision 0, AREVA NP Inc., November 2012 (Attachment 12), confirm that most recently approved methodologies were used, including RODEX4, which accounts for the degradation of thermal conductivity with increasing fuel burnup using upper limit on calculated clad oxide thickness, and the updated methodology for ACE correlation that addresses a nonconservatism in the axial averaging process used to determine the K factor, which is an input to the correlation (as discussed in page 2 of Enclosure to the LAR). If the most recently updated and approved methodologies were not used for BFN Unit 2 Cycle 19 reload safety analyses, then please provide justification.

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

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

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 3-1 3.0 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-478 Regarding Addition of Analytical Methodologies to TS 5.6.5 for Browns Ferry Nuclear Plant Units 1, 2, and 3, and Revision of TS 2.1.1.2 for BFN Unit 2 (TAC Numbers MF0877, MF0878 and MF0879)," USNRC, August 30, 2013. (38-9211081-001)

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

Revision 0, Browns Ferry Unit 2 Cycle 19 LAR Fuel Cycle Design, AREVA NP, August 2012.4. ANP-2860P Revision 2, Browns Ferry Unit 1-Summary of Responses to Request for Additional 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 Spectrum Analysis for A TRIUMTM IOXM Fuel, AREVA NP, October 2012.7. ANP-3153(P)

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

Revision 0, Evaluation of Fuel Conductivity Degradation for ATRIUM IOXM Fuel for Brown Ferry Units 1, 2 and 3, AREVA NP, November 2012.9. NEDO-32465-A, Licensing Topical Report, "Reactor Stability Detect and Suppress Solutions 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 of STAIF 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 of Responses to Request for Additional Information Extension for ATRIUM I OXM, AREVA NP, November 2012.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1OXM Revision 1 Fuel Transition Page 3-2 15. ANP-1 0298PA Revision 0 Supplement 1 P Revision 0, Improved K-factor Model for ACE/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 for ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, August 2012.17. ANP-1 0307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.18. ANF-89-98(P)(A)

Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR 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 not approved.)

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

Supplement 1 Revision 2, Qualification of Exxon Nuclear Fuel for Extended Bumup, Supplement 1, "Extended Burnup Qualification of ENC 9x9 BWR Fuel", May 1988.21. "Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations 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-1O BWR Reload Fuel, Siemens Power Corporation, December 1998.23. EMF-92-116(P)(A), Revision 0, Generic Mechanical Design Criteria for PWR Fuel Designs, February 1999.24. ANP-3150P Revision 0, "Mechanical Design Report for Browns Ferry ATRIUM IOXM Fuel Assemblies", AREVA Inc., October 2012.25. ASME Boiler and Pressure Vessel Code,Section III, Division 1, American Society of Mechanical Engineers.

26. EMF-93-177(P)(A), Revision 1, Mechanical Design for BWR Fuel Channels, August 2005.27. EMF-2971 (P), Revision 1, Mechanical and Thermal-Hydraulic Design Report for Browns Ferry 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 Batch BFE2-14 ATRIUM-10 Fuel Assemblies, September 2004.29. ANP-2537(P), Revision 0, Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-15 ATRIUM- 10 Fuel Assemblies, May 2006.30. XN-81-51 (P)(A), LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly, Exxon Nuclear Company, May 1986.AREVA NP Inc.

AREVA RAI Responses for ANP-3248NP Browns Ferry ATRIUM 1 OXM Revision 1 Fuel Transition Page 3-3 31. XN-NF-84-97(P)(A), LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly, Exxon Nuclear Company, August 1986.32. 51-9191258-001, Browns Ferry Unit 2 Cycle 19 MCPR Safety Limit Analysis With SAFLIM3D 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 4 License Condition Related to Treatment of Channel Bow Uncertainty Amendment Number XXX (Unit 1)License Condition Implementation Date The fuel channel bow standard deviation component Upon implementation of of 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) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

Amendment License Condition Number Implementation Date XXX (Unit 2)The fuel channel bow standard deviation component Upon implementation of of 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) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

Amendment License Condition Number Implementation Date XXX (Unit 3)The fuel channel bow standard deviation component Upon implementation of of 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) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.