ML081700102

Request for Additional Information for Extended Power Uprate - Round 18 (Ts-418) (TAC Nos. MD5263 and MD5264)

ML081700102
Person / Time
Site:	Browns Ferry
Issue date:	07/17/2008
From:	Ellen Brown NRC/NRR/ADRO/DORL/LPLII-2
To:	Campbell W Tennessee Valley Authority
Brown Eva, NRR/DORL, 415-2315
Shared Package
ML081990041	List: ML081700102
References
TAC MD5263, TAC MD5264
Download: ML081700102 (11)

Text

Ofic~ial Ubt Only - Pruprietauy I..formatien 74 UNITED STATES

°. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 17, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 18 (TS-418) (TAC NOS. MD5263 AND MD5264)

Dear Mr. Campbell:

By letter dated June 24, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted an amendment request for Browns Ferry Nuclear Plant (BFN), Unit 2 and 3, as supplemented by letters dated August 23, 2004, February 23, April 25, June 6, and December 19, 2005, February 1 and 28, March 7,9, 23 and 31, April 13, May 5 and 11, June 12, 15, 23 and 27, July 6, 21, 24, 26 and 31, December 1, 5, 11 and 21, 2006, January 31, February 16 and 26, and April 6, 18 and 24, March 6, July 27, August 13 and 21, September 24, Novemb6r 15 and 21, and December 14, 2007; January 25, February 11 and 21, March 6, April 4 and 9, and May 1, 2008. The proposed amendment would change the BFN operating licenses for Units 2 and 3 to increase the maximum authorized power level by approximately 15 percent.

A response to the enclosed Request for Additional Information (RAI) is needed before the Nuclear Regulatory Commission staff can complete the review. This RAI was discussed with Mr. James Emens of your staff on June 20, 2008, and it was agreed that TVA would respond'b&

September 15, 2008.

If you have any questions, please contact me at (301) 415-2315.

Sincerely,

• A Brown, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-260 and 50-296

Enclosures:

1. RAI (Public Version)

2. RAI (Non-Public Version) cc w/enclosure 1: See next page NOTE: THE PROPRIETARY ENCLOSURE 2 HAS BEEN SENT TO THE ADDRESSEE ONLY.

ALL OTHER DISTRIBUTION CONTAINS A NON-PROPRIETARY ENCLOSURE 1 ONLY.

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Official 8. - PIP. i a Y 2111, .lllptiou..

July 17, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 18 (TS-418) (TAC NOS. MD5263 AND MD5264)

Dear Mr. Campbell:

By letter dated June 24, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted an amendment request for Browns Ferry Nuclear Plant (BFN), Unit 2 and 3, as supplemented by letters dated August 23,.2004, February 23, April 25, June 6, and December 19, 2005, February 1 and 28, March 7, 9, 23 and 31, April 13, May 5 and 11, June 12, 15, 23 and 27, July 6, 21, 24, 26 and 31, December 1, 5, 11 and 21, 2006, January 31, February 16 and 26, and April 6, 18 and 24, March 6, July 27, August 13 and 21, September 24, November 15 and 21, and-December 14;-2007;-January 25TFebruary 11- and-21,-;March-6,-April4-and-9-and -. .

May 1, 2008. The proposed amendment would change the BFN operating licenses for Units 2 and 3 to increase the maximum authorized power level by approximately 15 percent.

A response to the enclosed Request for Additional Information (RAI) is needed before the Nuclear Regulatory Commission staff can complete the review. This RAI was discussed with Mr. James Emens of your staff on June 20, 2008, and it was agreed that TVA would respond by September 15, 2008.

If you have any questions, please contact me at (301) 415-2315.

Sincerely, IRA/!

Eva A. Brown, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-260 and 50-296 Enclosures

1. RAI (Public Version)

2. RAI (Non-Public Version) cc w/enclosure 1: See next page DISTRIBUTION: See next page ADAMSAccessionNo.: Public: 'M1L081700102 Pkg: ML081990041 Non-Public: .i C19°M7 N.-088 OFFICE 7PZ2-2IPM LPL2-2/LA DSS/SNPB LPL2-2IBC NAME EBrown M B/CI0Yt8 on n, AMendiola FSaba for TBoyce j14j DATE 7 /11 /08 7/16/08 107/16/08 7 OFFICIAL RECORD COPY Off.icia Us C..Iy - Propreetary .f 1 Ictit

Offieelial Usero ly PEArowpnidaty I 7,nfra2i00 Letter to William R. Campbell, Jr. from Eva A. Brown dated July 17, 2008

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 18 (TS-418) (TAC NOS. MD5263 AND MD5264)

DISTRIBUTION:

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REQUEST FOR ADDITIONAL INFORMATION EXTENDED POWER UPRATE ROUND 18 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN), UNITS 2 AND 3 DOCKET NOS. 50-260 AND 50-296 Table 1.3 in Enclosure 5 to the letter dated June 25, 2004;' indicates that the COTRANSA2 Version AAPR03 computer code was used to evaluate the anticipated transient without scram (ATVS) -. overpressurization event. The licensee cites a May 31,2000, letter from the Nuclear Regulatory Commission (NRC) to Framatome (now AREVA) to support the use of COTRANSA2 for the ATWS-overpressurization abnormal operating occurrence (AOO).

SRXB (Units 2 and 3)

91. In Enclosure 1 of the letter dated March 7, 2006, Tennessee Valley Authority (TVA) provides information in support of the use of the Ohkawa-Lahey void quality correlation against ATRIUM-1 0 test data in response to SRXB-A.35. The Ohkawa-Lahey void quality correlation appears to under-predict the void fraction for the majority of the thermodynamic qualities tested at 6.9 Megapascal (MPa). The void reactivity coefficient is sensitive to the instantaneous void fraction, generally becoming more negative with increasing void fraction.

Provide a quantitative determination of the impact of the bias in the void fraction in COTRANSA2 on ATWS overpressure analysis results for the bottom head peak pressure. This should include a comparison of the impact of the void bias to the margin between the peak calculated pressure and the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME) acceptance criterion of 1500 pounds per square inch gage.

In addition, address how known biases are taken into account for future cycle specific calculations and for bundle designs other than ATRIUM-1 0.

92. Subcooled boiling is a'phenomenon that can have a significant impact on the efficacy of a code system to accurately predict the axial power shape and bundle flow (by impacting the two phase pressure losses). Using a subset of the KATHY data, provide a similar comparison for the XCOBRA-T Levy subcooled void model as provided for the Ohkawa-Lahey correlation in response to NRC-RAI-SRXB-A.35.

Enclosure 1

93. Provide justification for the application of the Ohkawa-Lahey void-quality correlation to pressures above 6.9 MPa.

94. The initial steam flow rate at extended power uprate (EPU) conditions is higher than at pre-EPU conditions, and the transient power pulse is expected to be higher during the pressurization. The suppression pool temperature for Units 2 and 3 is based on an analysis for GE14 fuel. Provide a discussion on the means used to confirm that the.

results of the GE-14 analysis are bounding for ATRIUM-10 fuel. This justification should contain qualitative discussion regarding the impact of the differences in nuclear characteristics and should consider the timing and nature of the transient power response during pressurization, relief, and boration.

95. Provide details regarding the bypass flow during the ATWS overpressure transient as predicted by COTRANSA2. In particular determine if the bypass flow is downward during any portion of the transient. Justify the applicability of these results given any inherent constraints in the COTRANSA2 code.

96. Provide a plot of the total safety relief valve (SRV) flow during ATWS overpressurization for EPU and pre-EPU conditions.

97. Condition 2 of the safety evaluation dated May 23, 1990, approving Advanced Nuclear Fuels (ANF) 913(P)(A), COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, August 1990 requires consideration of time step. Provide details regarding the nodalization of the steam line and time step used for ATWS overpressure analysis. Specifically provide results of sensitivity studies to verify that the node size and time steps are sufficient to preclude numerical errors in the calculation of the pressure wave propagation to the reactor core from the main steam isolation valves.

98. It appears that COTRANSA2 has two centrifugal pump models, the first pump model neglects the inertia and the second pump model is based on homologous input. Identify which model option is used. If the second model option is. used, verify that it is used to model the dual recirculation pump trip during ATWS evaluations. Verify that the homologous input for the recirculation pumps for the Unit 2 analyses have been benchmarked against operational data at Unit 2.

99. The description of the steam line model in ANF-913(P)(A) does not provide descriptive details of the SRV model. Provide a description similar to Section 2.3.4 in ANF-913(P)(A) describing these models. In the description address the use of calculated pressures in modeling the SRV lift. When apportioning flow from a steam line node between an SRV and the downstream node, discuss how the SRV flow is calculated.

100. Section 2.1 of ANF-913(P)(A) states that cross sections are interpolated based on both controlled and uncontrolled states at [H 11 void fraction. These void cases appear to not be consistent with the void cases used to develop cross section response surfaces for MICROBURN-B2 E .]1, explain this discrepancy.

101. The Doppler coefficient is stated to be dependent on the broadening of the fast group cross section and to be a function of fuel temperature.

MICROBURN-B2 calculates the nodal fuel temperature based on quadratic fitting function. Provide this function. Discuss how the initial nodal fuel temperature is calculated. Provide a comparison of the quadratic function predicted nodal fuel temperature to results predicted using a more sophisticated thermal rod conduction model and heat transfer coefficient, such as XCOBRA-T.

Expand on the discussion provided in ANF-913(P)(A) and describe what combination of calculations is performed to determine the reactivity contribution from Doppler for ATWS overpressure analysis, for example, specify if a lattice calculation is performed to determine a coefficient relating microscopic cross sections to average fuel temperature.

Discuss whether the- rod -temperatures in Section 2.1.3-of ANF-913(P)(A) are calculated based on a nodal average rod or for each rod in the node. Clarify how the transient nodal average fuel temperature is calculated.

Provide a description of any differences between the COTRANSA2 thermal conduction models, including material properties, and the RODEX2 models.

Discuss whether the RODEX2 code was used to develop input for COTRANSA2 similar to XCOBRA-T.

102. 'In the response to SRXB-87 contained in the letter dated March 6, 2008, as supplemented in the letter dated May 1, 2008, TVA provided a sensitivity analysis to quantify the impact on the safety limit minimum critical power ratio (SLMCPR) of a potential increase in pin power peaking uncertainty at the harder spectrum EPU conditions.

XCOBRA-T is used to determine the transient effect on the critical power ratio (CPR).

To calculate the critical heat flux ratio XCOBRA-T requires S-factors. The S-factor accounts for lattice peaking and bundle geometry effects. Describe.how the additive constants and lattice calculations are used to determine the S-factors for use by XCOBRA-T. In particular, address whether the analyses are performed for local peaking using XFYRE, CASMO-4, or MICROBURN-B2.

103. Provide the relationship of the term Fe, to the S-factor. If axial integration is required to determine the S-factors, specify how this is performed. Address whether the S-factors are sensitive to the bundle void distribution. Describe how the S-factors are determined for conditions typical (or bounding) for operation at EPU conditions.

104. Describe how S-factors are determined for part length rods.

105. Verify that the Unit 2 transient analyses were performed using input options for closure relationships that are consistent with the NRC approval of XCOBRA-T. This includes specifying the Levy subcooled boiling model, the Martinelli-Nelson two phase friction

multipliers, the two phase component loss multiplier, the-wall viscosity model, and thermodynamic properties from the ASME steam tables.

106. At EPU conditions, the core steam flow rate is increased. The pressure response to events such as turbine trip and load rejection is expected to be exacerbated at EPU conditions relative to pre-EPU conditions.

The NRC staff notes that the SPCB critical power correlation for ATRIUM-1 0 fuel is not qualified above [f ].. In the analysis of the pressurization transients, address whether the pressure exceed [L J1 in the reactor core. Similarly compare the analysis conditions for those parameters listed in Condition 3 of the safety evaluation July 3, 2000, for EMF-2209(P)(A), SPCB CriticalPower Correlation.

107. Address how the wall friction and component loss coefficients were determined for Unit 2. Address whether these parameters were input in the analysis to account for friction. Provide these parameters and the technicalbasis.for their selection.. Relative to pre-EPU conditions, channel flow tends to redistribute at EPU conditions as there are fewer low resistance bundles in the core. Address whether the friction parameters were selected to be consistent with this expected trend.

108. At EPU conditions there are a higher number of higher powered bundles. It is possible,

-and likely, for large axial sections of these bundles to be in an annular flow regime.

Calculating pressure losses near bundle features such as fuel spacers can be important in the prediction of critical heat flux, which tends to occur below fuel spacers where the liquid film is typically thinnest.

On page 25 of Exxon Nuclear Company's XN-NF-84-105(P)(A), XCOBRA-T: A Computer Code for BWR Transient Thermal-HydraulicCore Analysis, it is stated that

"[t]his [Martinelli-Nelson] formulation was developed for horizontal flow, but is reasonably accurate for vertical flow where both phasic flow rates are high enough to ensure turbulent co-current flow." Justify why the Martinelli-Nelson two phase friction multipliers are applicable in annular flow regimes.

109. Section 3.3 of the Technical Evaluation Report attached to the NRC's safety evaluation approving XN-NF-84-105(P)(A) states that critical power calculations may be inaccurate if the inlet flow is negative or if the inlet quality is above zero. Verify that for the transient analyses that the bundle inlet flow is positive and that the inlet qualities are less than zero.

110. Transient cladding heat flux during transients will be a function of the heat hold-up in the fuel pins during AQOs. Identify all changes that have been made to the fuel rod thermal conduction models since the NRC's review and approval of RODEX2. Provide a comparison of the XCOBRA-T fuel rod models to those approved to the NRC during review of EMF-85-74(P), RODEX2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model, or RODEX2A. Also, provide a comparison of these models to those in BAW-1 0247(P), Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, or RODEX4.

111. If different historical models are preserved in XCOBRA-T relative to RODEX2A, justify the use of XCOBRA-T to model transients for fuel above the previously established burnup limits for RODEX2.

112. Some models may have been updated to conservatively bound experimental data collected subsequent to the NRC review and approval of RODEX2. The staff notes that certain assumptions may be conservative in the assessment of linear heat generation rate limits that may not be conservative when evaluating transient heat flux during AOO simulation -due to the competing effects of reactivity feedback and heat flux flow mismatch. If a model is "conservatively bounding" in RODEX2, and translated to XCOBRA-T, provide a discussion of the performance of the model for thermal margin transient calculations.

113. At EPU conditions a core contains a higher number of higher powered bundles. At these conditions the heat transfer may be driven by phenomena such as liquid entrainment and redeposition... The XCOBRA-.."T heat.ttransfer is priedicted according to Dittus-Boelter and Thom heat transfer correlations for forced convection and nucleate boiling, respectively. The NRC is aware that AREVA has temperature data from full scale critical power tests. Provide qualification of the heat transfer correlations to predict fuel rod surface temperatures for test conditions representative of higher powered bundles at flow rates similar to-EPU conditions for ATRIUM-10 [i.e., -8 megawatts thermal].

Particularly, provide the relevant qualification data near the top of the bundle where liquid entrainment and droplet redeposition is expected to have an impact on heat transfer.

114. XCOBRA-T accepts input from COTRANSA2 to capture the transient variation in power during AQOs. EPU cores are high energy cores and may have a higher peak hot excess reactivity, resulting in changes in SCRAM worth relative to pre-EPU conditions.

Provide a more detailed description of how effects such as transient variation in axial power shape during SCRAM are captured in XCOBRA-T. Address whether detailed nodal power histories translated from COTRANSA2 to XCOBRA-T.

115. At EPU conditions a larger number of bundles are operated at high power levels. It is expected, therefore, for more bundles to be near their thermal limits at the onset of a transient. Describe how the Unit 2 radial channel nodalization was evaluated. Address whether the potentially limiting bundles were grouped with non-limiting bundles. Provide the radial bundle power distribution (as predicted by MICROBURN-B2) and radial channel group assignments (in XCOBRA-T) for the initial conditions for the limiting transient analysis. Also, compare the COTRANSA2 radial grouping with the XCOBRA-T grouping. Provide justification that the resolution is sufficient to model the transient behavior in all potentially limiting bundles.

116. Address whether XCOBRA-T was used to demonstrate acceptable fuel rod thermal mechanical performance during transients. If XCOBRA-T is not used for this purpose, address how acceptable thermal mechanical performance is demonstrated during transients. If the method is not consistent with the models in RODEX2 or later NRC-approved thermal mechanical code, justify the approach.

117. Enclosure 4 of the letter dated June 25, 2004, references NEDO-32047-A. In particular it is noted that operation at EPU conditions is generally achieved by flattening radial core power. As a result of this flattening the second harmonic eigenvalue separation is'likely to be greatly reduced. Therefore, under non-isolation ATWS conditions it is expected that the core will be more susceptible to regional mode oscillations that at pre-EPU conditions..

Given the information provided in the NRC's contractors' technical evaluation report attached to the safety evaluation approving NEDO-32047-A dated February 5, 1994, Appendix C: "Consequences of Out-of-Phase Instability Mode Not Proven More Favorable than In-Phase Mode." Provide an evaluation of the likelihood of a regional mode oscillation to develop under non-isolation ATWS conditions. It is acceptable to evaluate the regional and core wide mode decay ratios for these conditions for an equilibrium ATRIUM-10 Unit 2 core using STAIF to respond to this request for additional information (RAI). Based on the available analyses, determine if such an oscillation at BFN would result in a significant-increase in the fuel damage relative to the results in NEDO-32047-A.

The analyses in NEDO-32047-A were performed for General Electric (GE) fuel. The analyses are generally applicable for pre-EPU core designs since hydraulic stability of the fuel products has improved or at least remained the same. Provide a comparison of

'the. channel stability characteristics of ATRIUM-10 to GE 8x8 fuel. If ATRIUM-10 is less

. /stable than GE 8x8 fuel, consider any impact on the projected consequences of a

.non-isolation ATWS instability event.

The following are related to the June 3, 2008, response to SRXB-88.

18*. In the'supplemental response to RAI SRXB-88, TVA provided the results of sensitivity analyses to evaluate the impact of void fraction uncertainty on the calculated delta-critical power ratio (DCPR) and the safety limit minimum critical power ratio (SLMCPR). In the void fraction reduction case, the DCPR is apparently unaffected and is accompanied by an increase in SLMCPR.

If the void fraction were reduced throughout the core by a fixed bias, the result would be to redistribute the reactor power according to the change in reactivity associated with the void perturbation. Since those bundles with the higher bundle average void fractions will have a greater reactivity response, a reduction in the void fraction will tend to increase, slightly, the power in those bundles with a higher bundle average void fraction relative to the bundles that had a lower void content prior to the perturbation. The bundles with a higher bundle average void fraction are the high powered bundles. Therefore, a fixed reduction in void fraction will increase the radial power peaking factor. The increased radial power peaking factor for a given steady state power level would result in fewer rods entering boiling transition as a result of a transient initiated from this state.

When this effect is considered, it is the equivalent of increasing the radial power peaking and-reducing the SLMCPR since fewer rods are at the limiting end of the pin power statistical distribution. In effect, the span of pin powers to account for the 0.1 percent of highest powered pins increases. Results of the TVA sensitivity analysis demonstrate the

opposite trend. It is expected that the imposition of a fixed void fraction reduction would result in a lower SLMCPR. Explain this discrepancy.

119. Continuing with the void fraction reduction case, the decrease in void fraction would simultaneously result in a redistribution of the axial power. Since those higher void nodes would have a greater reactivity response than low void nodes, the axial power distribution would shift upwards in the core. The upward shift in the axial power distribution has the effect of increasing the reactor adjoint in the upper portions of the core. As pressurization transients are typically limiting, the impact of an upward shift in axial power on the transient power prediction should be considered. The upward shift in reactor adjoint directly affects the core void reactivity coefficient and tends to increase the sensitivity of the core reactivity to a pressure wave, since the back pressure wave is dissipated by void collapse in the upper parts of the core. Therefore, the core wide transient power would be increased relative to the base case, which appears to result in an increase in the DCPR.

T'he resulIts of the TVA sensitivity analysis do not demonstrate this trend. Address why imposing a fixed void fraction reduction does not result in a higher DCPR.

120. The void increase cases exhibited opposite trends relative to the void reduction cases.

The staff found that the void reduction cases were not consistent with the staff's expectations. Provide information similar to the information requested in SRXB-1 18 and 119 for the fixed increase in void fraction sensitivity analyses.

For each case in Study 1 provide:

• The limiting bundle: core location, initial radial peaking factor and axial power shape

• Plots of the perturbed axial and radial core power shapes

• Plots of transient limiting bundle peak rod heat flux and mass flow rate

• Plots of transient CPR

• A comparison of the predicted power pulse heights and widths.

121. It should be noted that the increase in the operating limit minimum CPR (OLMCPR) for the increased void fraction cases is substantial relative to the base case. The Study 1 increase in'OLMCPR is 0.014 and the Study 2 increase in OLMCPR is 0.027.

In response to RAI SRXB-88, TVA stated that COTRANSA2 includes a 110 percent multiplier on integral thermal power as a conservative assumption. However, XN-NF-80-19(P)(A) Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX:

Thermal Limits Methodology Summary Description Section 4.4 states: "In developing the methodology for the COTRANSA code Exxon Nuclear addressed uncertainties in the code through the integral power variable. The revised methodology uses a more conservative deterministic bounding value (+10 percent) for the integral power uncertainty." TVA's evaluation of the 110 percent conservatism found that the OLMCPR margin afforded by the conservatism is [r )).

While analysis of pre-EPU reactor conditions, such as the Peach Bottom turbine trip tests, indicate that the integrated effect of all conservatisms in COTRANSA2 is adequate. The response to RAI SRXB-88 appears to indicate that at EPU conditions, the integral thermal power response to a 5 percent uncertainty in void fraction may not be bounded by the conservatism afforded by the 110 percent multiplier. This is evidenced by an increase in the OLMCPR in Study 2 that exceeds the conservatism afforded by the total 110 percent multiplier.

It should be noted that the approved method is based on an approach to conservatively bound all uncertainties, including uncertainties in other important variables such as flow and friction factors.

Given that the OLMCPR increase exceeds the 110 percent multiplier margin, provide a demonstration that the integrated effect of all conservatisms in COTRANSA2 for Unit 2 at EPU conditions is adequate. This demonstration may be provided by qualification against relevant operating plant transient data to ensure conservatism of the methodology for EPU or near-EPU conditions or by comparison against a rigorous statistical treatment of all uncertainties or by some alternative quantitative and applicable means.

122. The modified correlations are based on constant slip models. Provide a discussion regarding the treatment of subcooled boiling. This discussion should address void fraction continuity at the boiling boundary. Describe any impact on the transient analyses arising from SCRAM reactivity worth if significant differences are expected based on treatment of subcooled boiling.