ML082690016

Specifications (TS) Change TS-418 - Extended Power Uprate (EPU) - Supplemental Response to Request for Additional Information (RAI) Round 3 & 18 & Response to Round 20 Fuels Methods RAIs

ML082690016
Person / Time
Site:	Browns Ferry
Issue date:	09/19/2008
From:	Langley D Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MD5263, TAC MD5264, TS-418, TVA-BFN-TS-418
Download: ML082690016 (100)

Text

Tennessee Tennessee Valley Authority, Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 35609-2000 September 19, 2008 TVA-BFN-TS-418 TVA-BFN-TS-418 10 CFR 50.90 U.S. Nuclear Nuclear Regulatory Commission Commission ATTN: Document Control Desk

. Mail Stop OWFN, OWFN, P1-35 Washington, D. C. 20555-0001 Gentlemen:

Matter of In the Matter )) Docket Nos. 50-260 Tennessee Valley Authority ) 50-296 50-296 BROWNS FERRY NUCLEAR BROWNS NUCLEAR PLANT (BFN) (BFN) - UNITS UNITS 2 ANDAND 3 - TECHNICAL SPECIFICATIONS (TS) CHANGE TS-418 - EXTENDED EXTENDED POWER UPRATE (EPU) -

UPRATE (EPU)-

SUPPLEMENTAL SUPPLEMENTAL RESPONSE RESPONSE TO REQUEST FOR ADDITIONAL ADDITIONAL INFORMATION INFORMATION (RAI) (RAI)

ROUNDS 3 AND ROUNDS AND 18 AND RESPONSE TO ROUND AND RESPONSE ROUND 20 FUELS METHODS METHODS RAlsRAIs MD5263 AND (TAC NOS. MD5263 AND MD5264)

By letter letter dated June June 25, 2004 (ADAMS Accession No. ML041840301),

25,2004 ML041840301), TVA submitted aa license amendment amendment application application to the NRC NRC for EPU operation of BFN Units 2 and 3. The operation The pending EPU amendment increases the maximum authorized amendment increases authorized power power level by approximately 14 percent from 3458 megawatts thermal (MWt) to 3952 MWt.

approximately On July 17, 2008, NRC NRC issued a Round 18 RAI (ML081700102) regarding AREVA fuel fuel methods used in support of Units 2 and 3 EPU operations. Round 18 18 consists of 32 RAI questions, SRXB-91 through SRXB-122. To facilitate the review of TS-418, a meeting meeting was held on August 7,2008, 7, 2008, with NRCNRC staff to review draft responses responses to SRXB-91 through SRXB-116. Subsequently, on August 15, 2008, TVA submitted submitted aa partial response response (ML082330187)

(ML082330187) to Round 18; specifically RAIs SRXB-92, 93, 95, 96, 97, 99, specifically to RAls 100, and 99,100, 102 through 116. This submittal responds to the remainder remainder of the Round RAIs.

Round 18 RAls.

Additionally, NRC NRC staff conducted an audit of AREVA AREVA fuel methods methods from August 18 through August 28, 2008, at the AREVA engineering August 28,2008, engineering facilities in Richland, Washington.

As a result of the audit, TVA agreed agreed to provide supplemental supplemental responses to a number number of the August 15,2008, 15, 2008, Round 18 RAI responses responses and also to Round 3 RAls RAIs SRXB-A.34 and and SRXB-A.42.

SRXB-A.42. The original TS-418 Round 3 response response was submitted on March 7, 2006 (ML060680583). Lastly, this submittal also provides (ML060680583). provides responses to the five fuels related RAIs, RAls, SRXB-123 SRXB-123 through SRXB-127, from the NRC Round 20 RAI dated dated September 16,2008.

September 16, 2008.

U.S. Nuclear Regulatory Commission Nuclear Regulatory Page 2 September September 19, 2008 As discussed discussed with the NRC Project Manager Manager for BFN, Ms. Eva Brown, on September 17, September 17,2008, 2008, responses to remainder of the Round 20 RAls RAIs related to steam dryers, along with supplemental supplemental responses to Round 19 RAls RAIs EMCB.147 EMCB.147 and EMCB.192/150 pertaining to steam dryer analyses EMCB.192/150 analyses will be provided at a later later date. is a proprietary proprietary response response to the subject RAls RAIs and contains contains information that AREVA NP, Inc. (AREVA) considers to be proprietary in nature nature and subsequently, pursuant to 10 CFR 9.17(a)(4), 2.390(a)(4) and 2.390(d)(1),

pursuant 2.390(d)(1), AREVA requests that such information information be withheld from public disclosure. Enclosure 2 is a redacted redacted version of Enclosure with the proprietary proprietary material removed and is suitable for public disclosure.

Enclosure contains an affidavit from AREVA supporting supporting this request for withholding withholding from public disclosure.

determined that the additional TVA has determined additional information information provided provided by this letter letter does not affect affect the no significant significant hazards considerations associated hazards considerations associated with the proposed TS changes. The The proposed proposed TS changes still qualify for a categorical categorical exclusion from environmental environmental review pursuant to the provisions provisions of 10 CFR 51.22(c)(9).

No new regulatory regulatory commitments commitments are made in this submittal. If If you have any questions questions regarding regarding this letter, please contact me at (256)729-7658.

declare under penalty I declare penalty of perjury perjury that the foregoing is true and correct. Executed Executed on this this 19th day of September, 2008.

Sincerely, T. L ngle

0. Licensing and Site and Industry Affairs Manager Industry Manager

Enclosures:

1. Supplemental Supplemental Response to Request for Additional Information Information (RAI) Rounds 3 and 18 (RAI) Rounds and Response Response to Round 20 Fuels Methods RAIs RAls (Proprietary Information Version)

2. Supplemental Supplemental Response to Request for Additional Information Information (RAI) Rounds 3 and 18 (RAI) Rounds and Response to Round 20 Fuels Methods Methods RAIs (Non-Proprietary Information RAls (Non-Proprietary Information Version)

3. Affidavit Affidavit

u.s.

U.S. Nuclear Nuclear Regulatory Regulatory Commission Page 3 September September 19, 2008 2008

Enclosures:

cc (Enclosures):

State Health Officer Officer Alabama Alabama State Department Department of Public Health RSA Tower - Administration Administration Suite 1552 1552 P.O. Box 303017 303017 Montgomery, Alabama 3(3130-3017 36130-3017 Ms. Eva Brown, Manager Brown, Project Manager U.S. Nuclear Regulatory Commission Commission (MS 08G9)

One White Flint, North North 11555 Rockville Pike 11555 Rockville, Maryland Rockville, Maryland 20852-2739 20852-2739 Eugene F. Guthrie, Branch Branch Chief U.S. Nuclear Regulatory Commission Region IIII Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Atlanta, Georgia Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Nuclear Plant 10833 Shaw Road 10833 Road Athens, Alabama 35611-6970

u.s.

U.S. Nuclear Regulatory Commission Page 4 September 19, 2008 September JEE:BCM:BDL cc (w/o Enclosures):

G. P. Arent, EQB 1B-WBN W. R. Campbell, Campbell, Jr., LP 3R-C S. M.M. Douglas, POB 2C-BFN R. F. Marks, Jr., Jr., PAB 1C-BFN D. C. Matherly, BFT 2A-BFN L. E. Nicholson, L. Nicholson, LP 4K-C 4K-C L. E. Thibault, LP 3R-C L.

R. G. West, NAB 2A-BFN B. A. Wetzel, Wetzel, LP 4K-C S. A. Vance, WT 6A-K E. J. Vigluicci, ET 11A-K 11 A-K NSRB Support, LP 5M-C 5M-C EDMS WT CA-K, s:licensing/lic/submit/subs/EPU/RAI Round 3 and 18 and s:licensing/lic/submitlsubs/EPU/RAI and Round 20 Fuels Methods Methods FAls/Supplemental FAIs/Supplemental Response Response to to Request for Additional Request Information (RAI)

Additional Information (RAI) Rounds Rounds 3 and 18 and Response Response to Round 20 Fuels Methods Methods RAIs RAls

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION ENCLOSURE 2 ENCLOSURE TENNESSEE VALLEY AUTHORITY TENNESSEE BROWNS FERRY NUCLEAR NUCLEAR PLANT (BFN)

(BFN)

AND3 UNITS 2 AND 3 TECHNICAL SPECIFICATIONS (TS) CHANGE CHANGE TS-418 TS-418 EXTENDED POWER UPRATE EXTENDED UPRATE (EPU)

(EPU)

SUPPLEMENTAL RESPONSE SUPPLEMENTAL RESPONSE TO REQUEST REQUEST FOR ADDITIONAL ADDITIONAL INFORMATION INFORMATION (RAI)

(RAI)

ROUNDS 3 AND 18 AND RESPONSE ROUNDS RESPONSE TO ROUND 20 FUELS METHODS RAIs METHODS RAls (NON-PROPRIETARY INFORMATION (NON-PROPRIETARY INFORMATION VERSION)

VERSION) enclosure provides TVA supplemental This enclosure RAIs SRXB-A.34 supplemental responses to Round 3 RAls SRXB-A.34 and SRXB-A.42, a supplemental supplemental response response to NRC's July 17, 2008, Round 18 RAI, RAI, and a response response to the five fuels methods related RAIs, SRXB-123 related RAls, SRXB-123 through SRXB-127, NRC's SRXB-127, from NRC's September 16, 2008, Round September Round 20 RAI.

RAI.

,/

7

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION conducted an audit of NRC staff conducted AREVA fuel methods from August 18 of AREVA 18 through August 28, through August 28, 2008, 2008, AREVA engineering at the AREVA facilities in Richland, engineering facilities Washington. As Richland, Washington. As a result result ofof the audit, audit, TVA TVA agreed to provide supplemental agreed supplemental responsesresponses Round 3 RAIs SRXB-A.34 and SRXB-A.42.

RAls SRXB-A.34 SRXB-A.42. The The responses were originally previous Round 3 responses previous submitted originally submittecj on March March 7, 7, 2006 (ML060680853). A (ML060680853).

revised revised response 11, SRXB-A.34 was also submitted on May 11,2006 (ML061360148).

response to SRXB-A.34 2006 (ML061360148).

NRC RAI SRXB-A.34 (From RAI SRXB-A.34 (From Round 3) 3) qualitatively the Describe qualitatively cross-section reconstruction the cross-section incorporated in CASMO-4 process incorporated reconstruction process CASMO-4 and and MICROBURN-B2. The response MICROBURN-B2. response should reflect the information provided in information provided in the slides (1-35)

(1-35) of the August presentations, including August 44 presentations, fraction effects including high void fraction and accuracy.

effects and Provide flow accuracy. Provide flow chart(s), road map(s) and any chart(s), means to demonstrate any other means demonstrate the the process, starting from the the gathered raw void fraction gathered fraction data, how that data is used CASMO-4 to generate used by CASMO-4 generate the required required cross-sections. In addition, briefly describe In addition, development of the void fraction describe the development correlation and fraction correlation and associated uncertainties.

uncertainties.

Supplemental Response to SRXB-A.34 Supplemental SRXB-A.34 MICROBURN-B2 versions MICROBURN-B2 versions prior to 2003 treated cross section dependency on spectral section dependency spectral history between the fuel nuclide differently between differently depletion module and the neutron flux calculation module.

nuclide depletion module used ((

depletion module The fuel nuclide depletion calculation module used

)) while the neutron flux iteration calculation aa ((]. ]. This This inconsistency was remedied starting in 2003 by changing the depletion depletion module to the the*

(( ]. Starting from 2006,

]. 2006, converted to the ((

both modules were converted

].

].

changes over the years were mainly due to code These changes maintenance concerns and did not code maintenance impact any result due to the ((

].3. Unlike the cross section dependency instantaneous void, the [

dependency on the instantaneous

)) is rather rather weak. This is shown in Figure SRXB-A.34.1SRXB-A.34.1 for Pu-239 and and in in Figure SRXB-A.34.2 for Pu-240. The ((

1]. At the high end of (( ], the difference difference between the ((

difference is

]. This kind of difference uncertainty of nuclear cross section measurement entirely within the uncertainty measurement and its evaluation process evaluation process including including the CASMO-4 lattice CASMO-4 .lattice code. It code.' It has no observable effect on the reactor reactor nodal power distribution and the reactor criticality evaluation evaluation as has been verified in the code maintenance maintenance MICROBURN-B2.

record of MICROBURN-B2. .

E2-1

NON-PROPRIETARY NON-PROPRIETARY INFqRMATION INFORMATION rr"

..J

. Figure SRXB-A.34.1 SRXB-A.34.1 PU-239 sigma-1 Dependence Dependence on Spectral Spectral History at 20 Gigawatt-days Gigawatt-days per ton (GWd/T) r r-

-U Figure SRXB-A.34.2 SRXB-A.34.2 PU-240 sigma-1 Dependence sigma-1 Dependence on Spectral Spectral History at 20 GWd/T E2-2 E2-2

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION SRXB-A.42 (From Round 3)

NRC RAI SRXB-A.42 In August 30,2004, In 30, 2004, General Electric Nuclear Nuclear Energy (GENE) issued issued a 10 CFR Part Part 21 21 report ML042720293), stating (ADAMS ML042720293), stating that that using limiting control rod blade patterns developed developed for lessless than rated flow at rated power conditions could sometimes yield more limiting bundle-by-bundle bundle-by-bundle distributions and/or more limiting bundle MCPR distributions bundle axial power shapes than using limiting control rod patterns developed for rated flow/rated power in the SLMCPR calculation. The affected plants submitted amendment requests increasing their SLMCPR value. The staff understands understands that Framatome did not issue a Part 21 reporting on the SLMCPR methodology that addresses the calculation calculation of the SLMCPR at minimum core flow and off-rated conditions conditions similar to GENE's GENE's Part 21 report.

Reference the applicable sections of the ANF-524P-A SLMCPR methodology that specify the Reference the requirement to calculate the SLMCPR at the worst case conditions for minimum minimum core flow conditions for rated power. DemonstrateDemonstrate that the SLMCPR is calculated at different statepoints statepoints of the licensed operating operating domain, including including the minimum core flow statepoint and that the calculation is performed for different different exposure points.

Supplemental Response Supplemental Response to SRXB-A.42 AREVA AREVA NP1 NP 1 performs performs the safety limit Minimum Critical Power Ratio (SLMCPR) analysis on a cycle-specific cycle-specific basis. As discusseddiscussed in the original response to SRXB-A.42 SRXB-A.42 (Reference SRXB-A.42.1), the core power distributions used in the SLMCPR analyses are obtained from MICROBURN-B2 cycle-specific the MICROBURN-B2 cycle-specific design basis step-through step-through calculation. SLMCPR analyses analyses performed with these power distributions are performed distributions at the minimum minimum and maximum core flow allowed at rated power.

The SLMCPR analyses analyses supporting supporting the BFN Unit 2 EPU submittal were performed for an ATRIUMTM-10 2 equilibrium cycle ATRIUMTM-10 that assumed Maximum Extended Extended Load Line Limit Analysis Analysis plus (MELLLA+)

(MELLLA+) operation.

operation. The BFN EPU SLMCPR SLMCPR analyses considered considered the minimum and maximum flow at rated power for planned planned MELLLA+ cycle-specific SLMCPR MELLLA+ operation. The cycle-specific SLMCPR analyses supporting current operating operating cycles for BFN Units 2 and 3 were performed performed consistent with the currently allowed power/flow maps allowed power/flow maps for these cycles and did not include include the MELLLA+

MELLLA+

flow window. Future cycle-specific BFN Future cycle-specific SLMCPR analyses will be performed BFN SLMCPR consistent with performed consistent with the allowable allowable power/flow power/flow map map for the cycle.

The AREVA AREVA SLMCPR SLMCPR methodology uses a design basis core power distribution. distribution. The The criteria for for selecting the design design basis power power distribution distribution are specified specified in Reference Reference SRXB-A.42.2 SRXB-A.42.2 and state and state that analyses performed with power analyses be performed distributions that" power distributions that "...conservatively

... conservatively represent represent expected operating states reactor operating states which which could both both exist at the MCPR MCPR operating operating limit and produce produce a MCPR MCPR equalequal to the MCPR safety limit during an anticipated operational occurrence."

anticipated operational occurrence." CandidateCandidate design basis design basis power distributions are obtained from the cycle-specific cycle-specific design step-through.

step-through. The The design step-through reflects the cycle design step-through cycle design design energy and and operating operating strategy planned planned by thethe utility and is the best projection projection of how how the cycle cycle will operate.

The The design design step-through step-through is required to to have have margin margin to the operating operating limit limit MCPR MCPR (OLMCPR).

(OLMCPR).

Flatter Flatter (less peaked) peaked) radial power power distributions distributions areare conservative conservative for thethe SLMCPR SLMCPR analysis. The The radial power distributions power distributions from the cycle step-through cycle step-through are flatter than than the radial power power 1 AREVA 1

AREVA NPNP Inc. is an an AREVA and Siemens company.

company.

22 ATRIUM is aInc. is trademarkAREVA and of AREVA Siemens NP.

ATRIUM is a trademark of AREVA NP.

E2-3 E2-3

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION distributions that would result from adjusting the control rod patterns until the core OLMCPR OLMCPR is reached. These control rod adjustments adjustments would result in a more peakedpeaked radial power distribution distribution and increased increased margin margin to the SLMCPR. The design margin to the OLMCPR ensures ensures that the the power distributions from the cycle step-through are conservative conservative relative to the power power distributions distributions that may occur occur during during actual operation operation of the cycle.

Figure SRXB-A.42.1 comparison of the core radial power distribution from the design SRXB-A.42.1 provides a comparison design step-through and from actual operation for a BWR/4 BWRl4 at EPU.

EPU. The power power distributions are at thethe cycle exposure exposure that was limiting limiting for the SLMCPR analysis. The figure shows that the actual distribution had a higher radial power power distribution power distribution and is less flat than the design step-through power distribution. In step-through In addition, for actual operation there was still 5.1 5.1% % MCPR MCPR margin. These comparisons demonstrate comparisons demonstrate that the radial power distribution used power distribution used in the SLMCPR SLMCPR conservative relative analysis is conservative relative to the required required SLMCPR SLMCPR design basis power distribution and bounds actual operation.

operation.

Reference:

SRXB-A.42.1 Correspondence, W.D. Crouch Correspondence, Crouch (TVA) to U.S. Nuclear Regulatory Regulatory Commission, "Browns Ferry Nuclear Nuclear Plant (BFN)

(BFN) - Units 2 and 3, Response

Response

to NRC NRC Round 3 Requests Requests for Additional Information Information Related to Technical Specifications (TS) Change Specifications Change No. TS-418 - Requests Requests for Extended Power Power Uprate Operation (TAC Nos. MC3743 MC3743 and MC3744)," March March 7, 2006 (M L060680583).

(ML060680583).

SRXB-A.42.2 SRXB-A.42.2 ANF-524(P)(A) Revision 2 and Supplements ANF-524(P)(A) Revision Supplements 1 and 2, ANF Critical CriticalPower Methodology for for Boiling Boiling Water Reactors, Reactors, Advanced Nuclear Fuels Nuclear Fuels Corporation, November November 1990 1990 E2-4

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION 1.45 1.45 r r --

135 K-1.4 1.4 ** Design Design Step-Through Step-Through

• Actual Operation 1.35 1.3 1.3 " -

is u

U. 1.25 u.. 1.25 +---'-'...-~o;:------------------

0

~

c;;

-a 1.2

=s a:"'

1.15 + - - - - - - - - - - - - - --"""11..- - - - - - -

1.15 1.1 --

1.05 f-==----------- -----------~iiI!ooo.

o0 50 100 150 200 250 300 350 400 Assembly Rank Ran k Figure SRXB-A.42.1 SRXB-A.42.1 Design vs.

Factors Actual Radial Power Factors E2-5 E2-5

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 17, 2008, NRC issued a Round 18 RAI (ML081700102)

On July 17,2008, (ML081700102) regarding regardingAREVAAREVA fuel methods used in support used support of Units 2 and 3 EPU EPU operations.

operations. Round 18 consists consists of 32 RAI questions, questions, SRXB-91 through SRXB-122. To facilitate SRXB-91 through facilitate the review of TS-418, TS-418, a meeting was held on August 7, 7, 2008 with NRC staff to reviewreview draft draft responses responses to SRXB-91 SRXB-91 through SRXB-116.

through SRXB-1 16.

Subsequently, on August 15, 2008, Subsequently, 2008, TVA submitted a partial TVA submitted partialresponse response (ML082330187)

(ML082330187) to Round 18; specifically RAIs SRXB-92, 93, 95, specifically to RAls 95, 96, 97, 99, 96, 97, 99, 100, 100, and through 116. Below and 102 through are responses are remainderof the Round 18 RAIs.

responses to the remainder Additionally, NRC staff conducted RAls. Additionally, conducted an audit AREVA fuel methods from August 18 through audit of AREVA through August 28, 2008, at 28, 2008, at the AREVA engineeringfacilities engineering facilities in Richland, Richland, Washington.

Washington. As a resultresult of the audit, audit, TVA TVA agreed agreed to provide provide supplemental responses to a number of the August 15, 2008, supplemental responses 2008, Round 18 RAI responses, responses, which are provided below as are also provided indicated.

as indicated.

NRC Introduction Introduction to RoundRound 18 RAI Enclosure 5 to the letter dated June Table 1.3 in Enclosure 25, 2004, indicates June 25,2004, indicates that the COTRANSA2 COTRANSA2 Version AAPR03 computer computer code was used to evaluate evaluate the anticipated anticipated transient transient without scram (ATWS) - overpressurization overpressurization event. The licensee cites a May 31,2000, 31, 2000, letter letter from the Nuclear Nuclear Regulatory Commission (NRC) to Framatome (now AREVA) to support Regulatory support the use of COTRANSA2 COTRANSA2 for the ATWS-overpressurization ATWS-overpressurization abnormal operating operating occurrence occurrence (AOO).

NRC RAI SRXB-91 In Enclosure 1 of the letter dated March 7, 2006, Tennessee Tennessee Valley Valley Authority Authority (TVA) provides (TVA) provides information in support of the use of the Ohkawa-Lahey information Ohkawa-Lahey void quality correlation against against ATRIUM-10 ATRIUM-10 test data in responseresponse'to Ohkawa-Lahey void quality correlation to SRXB-A.35. The Ohkawa-Lahey correlation under-predict the void fraction for the majority appears to under-predict majority of the thermodynamic thermodynamic qualities tested at 6.9 Megapascal Megapascal (MPa). The void reactivity coefficientcoefficient is sensitive sensitive to the instantaneous instantaneous voidvoid fraction, generally becoming more more negative negative with increasing increasing void fraction.

Provide a quantitative quantitative determination determination of the impact of the bias in the void fraction in COTRANSA2 on ATWS overpressure COTRANSA2 overpressure analysis results for the bottom head peak pressure. This This should include comparison of the impact of the void bias to the margin include a comparison between the peak margin between calculated pressure pressure and the American Society of Mechanical Engineers Boiler &

Mechanical Engineers & Pressure Pressure Vessel Code (ASME) acceptance acceptance criterion of 1500 poundspounds per square square inch gage.

In addition, address address how known biases are taken into account for future cycle specific specific calculations and for bundle designsdesigns other than ATRIUM-1 ATRIUM-10. 0.

Clarificationsprovided Clarifications Providedby the NRC following a meeting on August 7, 7. 2008 Address Address the void bias for both the anticipated anticipated transient without scram (ATWS) overpressureoverpressure as well as ASME overpressure.

Response to SRXB-91

Response

AREVA AREVA performs cycle-specific cycle-specific ATWS analyses of the short-term reactor vessel peak pressure pressure using the COTRANSA2 COTRANSA2 computer code. The ATWS peak pressure calculation calculation is a core core wide wide pressurization pressurization event that is sensitive to similar phenomena as other pressurization pressurization transients.

Bundle design is included in the development development of input for the coupled neutronic and thermal E2-6 E2-6

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION hydraulic COTRANSA2 COTRANSA2 core model. Important Important inputs to the COTRANSA2 system model are biased in a conservative conservative direction.

The AREVA analysis analysis methods methods and the correlations correlations used by the methods are applicable for both pre-EPU pre-EPU and EPU conditions conditions as discussed discussed in responses (ML060680583)

(ML060680583) to previous previous RAIs RAls (SRXB-A.15, SRXB-A.26 through SRXB-A.29, and SRXB-A.35). The transient analysis (SRXB-A.15, SRXB-A.26 analysis methodology methodology is a deterministic deterministic bounding approach that contains bounding approach contains sufficient conservatism conservatism to offset biases and uncertainties uncertainties in individual phenomena. For bundle designs other than ATRIUM-10, individual phenomena. ATRIUM-10, the void-quality void-quality correlation correlation is robust as discussed discussed in the response (ML082330187) to RAI response (ML082330187)

SRXB-93 for past and present present fuel designs. For future fuel designs, the void-quality correlation would be reviewed reviewed for applicability, which may involve additional verification and validation.

A sensitivity study was performed for the limiting ATWS pressurization pressurization event event for BFN Unit 3 Cycle 14 with EPU to assess between the ATRIUM-1 assess the bias between ATRIUM-100 test data and the void-quality correlation. The event event was a pressure pressure regulator failure-open failure-open (PRFO), whichwhich is a depressurization depressurization event, followed by pressurization pressurization due to main steam line isolation valve (MSIV) (MSIV) closure. The neutronics input included included the impact impact of the fuel depleted depleted with the changes the changes in the void-quality correlation. To remove the bias in the MICROBURN-B2 MICROBURN-B2 neutronics input, the the

[ ] void-quality void-quality correlation correlation was modified.

modified. To address the bias in the Ohkawa-Lahey Ohkawa-Lahey void-quality correlation for the COTRANSA2 COTRANSA2 code, the void-quality void-quality relationship was changed changed to a (( ].]. Additionally, the sensitivity sensitivity study was repeated without depleting depleting the the fuel with the changes in the void-quality void-quality correlation (the change in the void-quality correlation was instantaneous instantaneous at the exposure of interest).

The reference reference ATWS case had a peak vessel pressure of 1477 1477 pounds per square inch gauge gauge (psig). The change change in the void-quality correlations resulted in a 10-psi void-quality correlations 10-psi increase in the peak vessel pressure. The results for an instantaneous instantaneous change change in the void-quality void-quality correlation correlation showed the same impact. A study was also performed performed for the ASME overpressure overpressure event for BFN BFN Unit 3 Cycle 1414 with EPU. The event was the MSIV MSIV closure with flux scram. The change in the the correlations resulted in a 7 psi increase in the peak vessel void-quality correlations vessel pressure. The impact impact of a change change in the bias of the void-quality void-quality correlations on peak pressurepressure is expected expected to be more than offset by the model conservatisms. However, until quantitative conservatisms quantitative values of the conservatisms can be demonstrated, demonstrated, TVA has directed directed AREVA to include a 10-psi 10-psi increase increase to the peak vessel pressure pressure for the EPU ATWS overpressureoverpressure analysis analysis and aa 7-psi 7-psi increase increase to the peak vessel pressure pressure for the EPU ASME overpressure overpressure analysis.

NRC SRXB-94 NRC RAI SRXB*94 The initial steam flow rate at extended extended powerpower uprate uprate (EPU)

(EPU) conditions is higher than at pre-EPU conditions, and the transient power power pulse is expected to be higher during during the pressurization.

pressurization.

The suppression suppression pool temperature 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 ATRIUM-1 0 fuel. This justification justification should contain qualitative discussion regarding qualitative discussion the impact of the differences differences in nuclear characteristics and should consider the timing and nuclear characteristics nature of the transient power response during pressurization, nature pressurization, relief, and boration.

Response to SRXB-94SRXB*94 The higher higher initial steam flow at EPU conditions will result in aa slightly higher power during power pulse during the initial relatively short pressurization pressurization phase of the ATWS event. However, the total energy released to the suppression suppression pool is dominated dominated by the later much longer phase of the event event where power is reduced reduced after the recirculation recirculation pumps trip and the core power is slowly reduced E2-7

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION as boron injection injection occurs. The ATWS analyses performed performed for BFN BFN Units 2 and 3 included included the the impact of the higher initial steam flow at EPU conditions.

conditions. As shown in Table 9-4 of Reference Reference SRXB-94.1, the impact of EPU operation SRXB-94.1, operation on the maximum maximum suppression pool temperature temperature is not significant <<1°F).

significant supports the conclusion that the initial power

(<1 OF). This supports power pulse, which is higher forfor EPU operation, is not significant significant relative to the total energy transferred transferred to the suppression suppression pool.

The suppression suppression pool temperature performed for BFN Units 2 and 3 with GE fuel temperature analyses were performed (Reference SRXB-94.1). An evaluation evaluation was performed performed to compare fuel neutronic parameters parameters important for the ATWS analysis (void coefficient, boron worth) for ATRIUM-1 ATRIUM-100 and GE fuel.

The boron worth characteristics characteristics of ATRIUM-10 ATRIUM-1 0 were were slightly better better while the void reactivity characteristics characteristics were slightly worse relative to the impact on the ATWS suppressionsuppression pool temperature temperature analysis.

Additional analyses were performed performed to assess assess the impact of the difference difference in fuel assembly assembly characteristics on the suppression reactivity characteristics temperature during an ATWS. ((

suppression pool temperature

]I All A" fuel types in the core designs designs including the GE fuel were explicitly explicitly modeled in the above above analyses analyses consistent with the approved approved methodology.

methodology. The GE fuel was modeled with a level of detail equivalent to that used for the ATRIUM-10 ATRIUM-10 fuel. CASMO-4 CASMO-4 analyses analyses explicitly modeled the water rod configuration of the GE fuel. MICROBURN-B2 MICROBURN-B2 was used to calculate calculate the core characteristics provided to the COTRANSA2 reactivity characteristics COTRANSA2 analysis. The GE fuel assembly geometric geometric and nuclear characteristics (enrichment and gadolinia distribution) were based on nuclear characteristics design design data provided to AREVA by TVA. The hydraulic hydraulic characteristics characteristics for the GE fuel assemblies were were based on GE fuel assembly pressure drop tests performed performed by AREVA.

The BFN ATWS analyses describeddescribed above were performed for cycles operating at pre-EPU power levels. However, as shown in Table 9-4 of Reference Reference SRXB-94.1, SRXB-94.1, the impact impact of EPU operation on the maximum maximum suppression pool temperature temperature is not significant. Therefore, the the trends observed for ATRIUM-10 ATRIUM-10 fuel in the above analyses are equally applicableapplicable for EPU operation.

operation.

The analyses analyses described described above confirm that the suppression suppression pool temperature analysis temperature analysis documented in Reference documented Reference SRXB-94.1 is slightly conservative conservative for ATRIUM-10 ATRIUM-10 fuel. In addition, the analyses show that the difference difference in reactivity reactivity characteristics characteristics between between ATRIUM-1 0 and GE fuel do not have a significant significant impact relative to the large margin to the suppression suppression pool temperature limit shown in Reference SRXB-94.1.

temperature SRXB-94.1.

The conclusions of the Reference SRXB-94.1 suppression Reference SRXB-94.1 suppression pool temperature temperature analysis are applicable for ATRIUM-10 ATRIUM-1 0 fuel and the acceptance acceptance criteria will wi" be met for BFN Units 2 and 3 EPU operation operation with ATRIUM-10 ATRIUM-10 fuel.

E2-8 E2-8

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION SRXB-94.1 NEDC-33047P Revision Revision 2, Browns Ferry Ferry Units 22 and 3 Safety Analysis Report for Extended Power Uprate, June 2004. (ML041840301)

(ML041840301)

Table SRXB-94.1 Energy Table SRXB-94.1 Release to Energy Release Suppression Pool Suppression Pool

[

]

NRC RAI SRXB-98 ItIt appears appears that COTRANSA2 COTRANSA2 has two centrifugal pump models, the first pump model neglects neglects the inertia and the second second pump model is based on homologous input. Identify which which model option is used.

used. IfIf the second second model option is used, verify that it is is used to model the dual recirculation pump trip during ATWS evaluations. Verify that the homologous homologous input for the the recirculation recirculation pumps for the Unit 2 analyses have have been benchmarked benchmarked against operational operational data at Unit 2.

Response to SRXB-98 SRXB-98 The second pump model model based on homologous homologous input input is used to model the dual recirculation recirculation pump trip during ATWS evaluations. The homologous curves are from the pump manufacturer.

The pump speed and flow are initialized initialized from operational plant data. Frictional torque and pump pump inertia are tuned to model the plant coastdown inertia coastdown rate.

E2-9

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NRC RAIRAI SRXB-100 SRXB-100 ANF-913(P)(A) states that Section 2.1 of ANF-913(P)(A) that cross sections are are interpolated interpolated based on both both controlled and uncontrolled states at (((( )) void fraction.

These void cases appear to not be consistent with the void cases used to develop cross cross section MICROBURN-B2 ((

response surfaces for MICROBURN-B2 )), , explain

))

this discrepancy.

SRXB-100 Supplemental Response to SRXB-100 In order to produce the COTRAN neutronic parameters, a series of MICROBURN-B2 MICROBURN-B2 calculations are performed. These successive calculations are:

(1) conditions Nominal initial conditions (2)

E2-10 E2-10

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION

]i The 11/2energy group diffusion equation in steady-state can be written as The 1Y2 energy group diffusion equation in steady-state can be written as

+ E-2. V"f2 (I),

2:a2 eP1 ++/-

V fIN =0

':a2

)

keff leakage. This equation The first term is a leakage. integrated over the cylindrical node depicted in the equation is integrated the following figure.

H 01j+1 fý Dr"+

1j11 H

D1,i H

Dl 1 ,i-1 The leakage term is approximated approximated as:

2DI,iDI,j(0

_ L3 2D] .0] .(r!J].

,I,} 1,i--r!J]l,j)-) A

, I , } __

A j=1 j=] (DO,i (Dl,i + D],j)

OD,j) HV E2-11 E2-11

NON-PROPRIETARY NON-PROPRI INFORMATION ETARY INFORMATION where Do D1 ,i == 0D for plane of interest Dj 01,j = 0D for the nodes adjacent to the plane of interest

=

01,i

([J1,i = flux in the plane of interest interest 014j C/J1,j =

= flux in the regions adjacent adjacent to the plane of interest A

A = surface surface area between nodesnodes ii and j H =

= distance distance between nodes i and nodes jj V

V =

= node volume volume

[

E2-12 E2-12

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION E2-13 E2-13

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION

,I E2-14 E2-14

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION These final one-group one-group cross section and leakage parameters are used in a new 1-dimensional leakage parameters flux solution and the axial power power distribution is updated updated for the next thermal hydraulic solution.

Iterations between Iterations between the 1-dimensional 1-dimensional flux solution and the thermal hydraulic hydraulic solution are repeated until converged converged results are obtained for core power, power distribution, distribution, temperature temperature distribution, and density distribution.

E2-15 E2-15

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr..

.J Figure SRXB-100.1 Comparison of Scram SRXB-100.1 Comparison Scram Bank Worth for

[ ]I E2-16 E2-16

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION NRC RAJ RAI SRXB-101 Doppler coefficient is stated to be dependent The Doppler dependent on the broadening broadening of the fast group cross cross section and to be a function of fuel temperature.

• • MICROBURN-B2 calculates the nodal fuel temperature MICROBURN-B2 calculates temperature based on quadratic quadratic fitting function.

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

XCOBRA-T.

• • Expand Expand on the discussion provided provided in ANF-913(P)(A)

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

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

• • Provide Provide aa description of any differences differences between between the COTRANSA2 COTRANSA2 thermal conduction conduction models, including material material properties, and the RODEX2 models. Discuss whether the the RODEX2 RODEX2 code was used to develop develop input for COTRANSA2 similar to XCOBRA-T.

Response

Response to SRXB-101

[

E2-17 E2-17

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION E2-18 E2-18

NON-PROPRIETARY INFORMATION NON-PROPRI ETARY INFORMATION E2-19 E2-19

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY I

• The RODEX2 computer code provides initial input information relative to core average average fuel-to-cladding gap fuel-to-cladding COTRANSA2 computer Gode.

gap heat transfer coefficients for the COTRANSA2 code. As As steady-state heat conduction models. The heat such, RODEX2 uses steady-state heat conduction model employed employed by COTRANSA2 COTRANSA2 includes includes transient terms.

The fuel thermal conductivity correlations used by COTRANSA2 are equivalent to the the RODEX2 models.

COTRANSA2 computes a fuel temperature COTRANSA2 temperature for each axial plane in the core. Based on assumption of aa core composition the assumption composition primarily consisting of uranium dioxide, COTRANSA2 does not account COTRANSA2 gadolinium in the fuel thermal conductivity account for gadolinium conductivity calculation.

capacities of fuel components Heat capacities cladding) are not components (uranium dioxide, gadolinium, and cladding) not steady-state calculations, but are used in the COTRANSA2 required for the RODEX2 steady-state COTRANSA2 transient transient calculations.

The fuel pellet-to-cladding coefficient used in COTRANSA2 pellet-to-cladding gap heat transfer coefficient COTRANSA2 is the the product of a RODEX2 calculation.

calculation.

E2-20 E2-20

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

..J Figure SRXB-101.1 Evolution of the SRX8-101.1 RODEX Evolution the Effective Fuel Temperature Doppler Effective Temperature for SPC Fuel at Constant Power*

Power-E2-21

NON-PROPRI ETARY INFORMATION NON-PROPRIETARY INFORMATION r

..J Figure SRXB-101.2 RODEX Evolution of the Doppler Effective Fuel Temperature Temperature for for SPC Fuel vs. LHGR and Burnup Burnup E2-22 E2-22

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

Figure SRXB-101.3 SRXB-101.3 MICROBURN-B2 MICROBURN-B2 Correlation Evolution of the Correlation Evolution the Doppler Effective Doppler Effective Fuel Temperature for Temperature for SPC Fuel vs. LHGR and Burnup Burnup E2-23 E2-23

r NON-PROPRIETARY NON-PROPRI INFORMATION ETARY INFORMATION NRC RAI SRXB-103 SRXB-103 Provide the relationship of the term Feff to the S-factor. If Provide If axial integration integration is required required to to determine determine the S-factors, specify how this is performed.performed. Address whether the S-factors are sensitive to the bundle void distribution.

distribution. Describe Describe how the S-factors are determined determined for conditions typical (or bounding) bounding) for operation operation at EPU conditions.

Supplemental Response Supplemental Response to SRXB-103 SRXB-103 Evaluations performed to assess the impact on ACPR Evaluations were performed ~CPR of a change change in Feff resulting from the the variation in the lattice void fraction during pressurization event. MICROBURN-B2 during a pressurization MICROBURN-B2 analysesanalyses were performed correlation and an adjusted void correlation to assess performed using the nominal void correlation assess the change change in Feff as void changes. The MICROBURN-B2 MICROBURN-B2 cases were run to reflect an instantaneous instantaneous change change in core average void fraction of +0.05. For the limiting MCPR bundle in in the core, the changes in void, local peaking factor (LPF), and Feff were:

~void Avoid == +0.0441

+0.0441 (node 24)

~void Avoid == +0.0456 (node 23)

~LPF ALPF == -0.0026

-0.0026 (node 24)

~LPF ALPF == -0.0030

-0.0030 (node 23)

~Feff AFeff == 0.0000 (assembly)

For other potentially potentially limiting bundles (10% highest powered (10% highest powered bundles) in the core, the change change in in Feff was between -0.0002 and +0.0011 for aa +0.05 core average Avoid. ~void. InIn general, an increase increase in void fraction resulted in an increase in Feff for high power, low exposure (end of first cycle) assemblies assemblies and aa decrease decrease in Feff for low power, high exposureexposure assemblies.

A decrease decrease in Feff during the transient will improve the CPR during the transient and result in aa reduced ~CPR.

ACPR. The converse converse is true for an increase increase in Feff Feff during during the transient. The sensitivity of MCPR to Feff is about 2 to 1; therefore, the sensitivity ACPR is about twice the ~Feff sensitivity of ~CPR AFeff during during the transient. The change in ~CPR ACPR would be between between 0.000 and +0.002 for a +0.05 core average Avoid.

average ~void.

During a pressurization pressurization event, the core void will initially decrease followed followed by an increase increase in core core void. Therefore, the effect effect of the change in void on fuel rod peaking factors Feff) will factors (and Feff) tend to be offset during the transient.

The assessment assessment above for the impact of a void change change on ~Feff A(ACPR) is based on AFeff and ~(~CPR) assuming assuming the nuclear nuclear power is instantly instantly converted converted to surface heat flux. Because Because the time of MCPR (-1.25 sec) is less than the fuel rod thermal time constant (- (- 5 sec), the actual impact impact on Feff and ACPR

~CPR from the void change will be much less. At the boiling transition transition plane, there is an insignificant insignificant change change in void until after the time of peak power. Because the increase increase in void void and the corresponding corresponding increase increase in Feff occur close to the time of MCPR, the slight change change in rodrod power will not significantly significantly change the rod heat flux at the time of MCPR. Therefore, the effect effect on ~CPR ACPR will be much less than estimated estimated based on the MICROBURN-B2 MICROBURN-B2 analyses.

In summary, the above results show that the effect effect of the variation in void fraction during a transient on the Feff has an insignificant insignificant effect effect on ~CPR.

ACPR.

E2-24 E2-24

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION NRC RAI SRXB-105 SRXB-105 Verify that the Unit 2 transient analyses were performed using input options for closure closure relationships relationships that are consistent with the NRC approval approval of XCOBRA-T. This includes specifying includes specifying the Levy subcooled subcooled boiling model, the Martinelli-Nelson Martinelli-Nelson two phase friction multipliers, multipliers, the two phase component loss multiplier, the wall viscosity model, and thermodynamicthermodynamic properties from from the ASME steam tables.

Revised Response Revised Response to SRXB-105 SRXB-105 The BFN BFN Units 2 and 3 EPU transient analyses used the default default models of XCOBRA-T.

XCOBRA-T. The The default models include the Levy subcooled boiling model, the Martinelli-Nelson Martinelli-Nelson two phase phase friction multipliers, the two phase component component loss multiplier, and the heated heated wall viscosity correction model.

correction model. [

discussed in a meeting

] as discussed meeting with the NRC on May (Reference SRXB-1 May 4,1995, (Reference SRXB-105.1).

05.1). Thermodynamic properties from the ASME Thermodynamic properties ASME steam tables were used. The code providesprovides a message message ifif the default default models are not used. Per AREVA's licensing licensing analyses requirements, use of default models is required. required.

Reference:

SRXB-105.1 SRXB-105.1 Correspondence, R.A. Copeland (Siemens) to R.C. Jones (NRC), "ATRIUM-10 Correspondence, "ATRIUM-10 Presentations," RAC:95:080, May May 4, 1995 (38-9091703-000).

NRC SRXB-107 NRC RAI SRXB-107 Address how the wall friction and component loss coefficients coefficients were determined determined for Unit 2.

Address whether whether these parameters parameters were input in the analysis to account account for friction. Provide friction. Provide these parameters parameters and the technical basis for their selection. Relative Relative to pre-EPU pre-EPU conditions, channel flow tends to redistribute at EPU conditions as there are fewer low resistance resistance bundles bundles in the core. Address whether whether the friction parameters parameters were selected selected to be consistent this consistent with this expected expected trend.

Supplemental Supplemental Response Response to SRXB-107 SRXB-107 During the NRC audit of AREVA codes and methods methods in Richland, Richland, Wa.,

Wa., from August 18 throughthrough August 28, 2008, the NRC requested additional 28,2008, information regarding additional information regarding the background background and process that [

process

].

Spacer Pressure Drop Testing Spacer Pressure Testing The Portable Hydraulic Hydraulic Test Facility (PHTF) is used by AREVA to obtain single phase loss loss coefficients coefficients for the spacers. The friction factor correlation is based on previous tests performed at the PHTF that remain applicable applicable for current fuel designs (rods and channelchannel have a consistent surface surface condition). The pressure pressure drops across across the spacers spacers are measured in the PHTF for each new new fuel design. The PHTF has pressure taps just upstream upstream of the spacers spacers so that the flow will be fully developed. component of pressure drop developed. The component drop due to friction is calculated calculated and subtracted subtracted from the total measured pressure drop. The remaining pressure drop is due due to the spacers spacers and is used to determine determine the spacer spacer pressure loss coefficients.

coefficients.

E2-25 E2-25

NON-PROPRIETARY NON-PROPRIETARY INFORMATIONINFORMATION Preliminary ATRIUM-10 Preliminary Spacer Loss Coefficients ATRIUM-10 Spacer Coefficients Development ATRIUM-10 fuel design took place in Germany. Because PHTF pressure Development of the ATRIUM-10 pressure drop testing was not complete, single phase pressure drop data for ATRIUM-10 ATRIUM-10 was obtained development effort. For the use in preliminary ATRIUM-10 design from the German development assessments, the German data was used to develop develop single phase spacer pressure loss loss coefficients appropriate coefficients use with Richland hydraulic models. Analyses using these single appropriate for use single phase phase losses resulted in an under under prediction prediction of the pressure drop data as shown in Figure Figure SRXB-107.1.

SRXB-1 07.1. The spacer loss coefficients (K) (K) used to generate generate the results presented presented in in Figure Figure SRXB-1 07.1 are of the form SRXB-107.1

+ B Re c K=A +BReC where A, B, and C are constants constants and Re is the Reynolds Reynolds number number based on local local fluid conditions conditions and geometry.

Until PHTF data was available for the ATRIUM-10ATRIUM-10 design, a means the means of adjusting the German-based German-based pressure loss coefficients to better predict predict the pressure drop data using Richland methods was developed. ((

Richland

)) are are shown in Figure SRXB-107.2. The spacer loss coefficients coefficients (K)(K) used generate the results used to generate results presented presented in Figure SRXB-107.2 SRXB-107.2 are of the form

[ ]

where (( )) for the ATRIUM-10 ATRIUM-10 design.

Further development development of ATRIUM-1 ATRIUM-100 spacer loss coefficients coefficients was subsequently subsequently performed performed based on PHTF ATRIUM-10 ATRIUM-10 pressure pressure drop data.

PHTF ATRIUM-10 Based Spacer Loss Coefficients Coefficients The ATRIUM-10 ATRIUM-10 PHTF tests form the basis for the single single phase loss coefficients coefficients currently used for design licensing analyses supporting design and licensing supporting domestic domestic BWRs. PHTF data was reduced to determine determine single phase losses for the spacers spacers in the lower (fully-rodded)

(fully-rodded) region of the bundle, the spacer in the transition transition (end of part-length rods) region of the bundle, and the spacers in the the (partially-rodded) region of the bundle.

upper (partially-rodded)

Assessments Assessments of the predicted pressure drop relative to measured measured two phase pressure pressure drop data data confirmed confirmed the applicability applicability of the [

[ ]

] for use with spacer pressure loss loss coefficients coefficients based on PHTF data. Results of analyses for each region of the bundle (lower, transition, upper) when usingusing the PHTF spacer coefficients [

spacer loss coefficients ] are are shown shown in Figures Figures SRXB-107.3, SRXB-107.4, and SRXB-107.5. SRXB-107.5.

NRC Interactions Interactions 4, 1995, a meeting On May 4,1995, meeting was held with the NRC to describe ATRIUM-10 design and the describe the ATRIUM-10 the application application of the approved AREVA methodology methodology for the design. Two view graphs extracted from those presented at the meeting are provided provided in Figures SRXB-107.6 and SRXB-107.7.

Figures SRXB-107.6 SRXB-1 07.7.

A summary summary of the May 4, 1995 meeting 1995 meeting was provided provided to the NRC in Reference SRXB-107.1.

Reference SRXB-107.1.

E2-26 E2-26

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Applicability Applicability for EPU Operation Operation ATRIUM-10 hydraulic models have been verified over a range of conditions that bound both The ATRIUM-10 pre-EPU pre-EPU and EPU operating conditions. The applicability of the models is described and supported supported by data presented in the thermal hydraulics section of the response to RAI RAI SRXB-A.15 (Reference SRXB-A.15 (Reference SRXB-1 07.2).

SRXB-107.2).

References:

SRXB-107.1 SRXB-107.1 Correspondence, Correspondence, R.A. Copeland (Siemens) to R.C. Jones Jones (NRC), "ATRIUM-10 "ATRIUM-10 Presentations," RAC:95:080, May 4, 1995 (38-9091703-000).

SRXB-107.2 SRXB-107.2 Correspondence, w.o. Crouch (TVA) to U.S. Nuclear Correspondence, W.D. Nuclear Regulatory Regulatory Commission, "Browns Ferry Nuclear Plant (BFN)

(BFN) - Units 2 and 3, Response to NRC NRC Round 3 Requests for Additional Information Related Additional Information Related to Technical Specifications Technical Specifications (TS)

Change No. TS-418 - Requests for Extended Extended Power Uprate Operation (TAC Nos.

MC3743 and MC3744)," March 7, 2006 (ML060680583).

MC3743 E2-27

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

.J SRXB-107.1 ATRIUM-10 Figure SRXB-107.1 ATRIUM-10 Bundle Drop Bundle Pressure Drop

[I ]I E2-28

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

..J Figure SRXB-107.2 ATRIUM-10 Bundle Pressure SRXB-107.2 ATRIUM-10 Pressure Drop Drop

((] . ]

E2-29

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY r

.J SRXB-107.3 ATRIUM-10 Figure SRXB-107.3 Drop Lower Region Spacer Pressure Drop ATRIUM-10 Lower Using PHTF Loss Coefficients Coefficients I[ I]

E2-30 E2-30

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

..J Figure SRXB-107.4 SRXB-107.4 ATRIUM-10 ATRIUM-10 Transition Transition Region Spacer Pressure Pressure Drop Drop Coefficients Using PHTF Loss Coefficients

[I ]I E2-31

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

..J Figure SRXB-107.5 ATRIUM-10 Upper Region Spacer SRXB-107.5 ATRIUM-10 Pressure Drop Spacer Pressure Drop Using PHTF Loss Coefficient Coefficient

[I I]

E2-32

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-Viewgraph From May Figure SRXB-107.6 Viewgraph May 4,1995 Presentation to NRC Presentation NRC Regarding ATRIUM-10 Regarding ATRIUM-10 Fuel r

r-Figure SRXB-107.7 Viewgraph From May 4,1995 SRXB-107.7 Viewgraph 4, 1995

..J rn-i Presentation to NRC Presentation Regarding ATRIUM-10 ATRIUM-10 Fuel E2-33

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NRC RAI SRXB-108 SRXB-108 At EPU conditions number of higher powered bundles. It conditions there are a higher number It is possible, and and likely, for large axial sections of these bundles to be in an annular flow regime. Calculating Calculating pressure losses near bundle features features such as fuel spacers can be important important in the prediction of critical heat flux, which tends to occur occur below fuel spacers where the liquid film is typicallytypically thinnest.

On page 25 of Exxon Nuclear Company's XN-NF-84-105(P)(A),

Exxon Nuclear XN-NF-84-105(P)(A), XCOBRA-T: A Computer Code A Computer Code for BWR for Transient Thermal-Hydraulic BWR Transient Thermal-HydraulicCore Core Analysis, it is stated that "[t]his [Martinelli-Nelson]

[Martinelli-Nelson]

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

Supplemental Supplemental Response to SRXB-108 SRXB-108 When applying the XCOBRA-T XCOBRA-T two phase pressure drop models implemented in the models implemented the 1-dimensional hydraulic model of the COTRANSA2 1-dimensional COTRANSA2 code, the local (spacer grid) pressure losses are automatically automatically adjusted to preserve preserve the core pressure pressure drop predicted by the more more detailed 3-dimensional representation in MICROBURN-B2.

3-dimensional hydraulic representation MICROBURN-B2. The XCOBRA-T XCOBRA-T initial flow rate is defined defined by a hydraulic hydraulic demand curve predicted by XCOBRA, which defines the the relationship between relationship between assembly power and the initial flow rate and accounts assembly power accounts for the lack of a core bypass model in XCOBRA-XCOBRA-T. T.

The orifice loss coefficient is automatically automatically adjusted adjusted in XCOBRA-T XCOBRA-T to preserve COTRANSA2 preserve the COTRANSA2 MICROBURN-B2) initial core (and MICROBURN-B2) core pressure drop and the initial flow rate defined defined by the the hydraulic demand hydraulic demand curve. Therefore, Therefore, adjustments adjustments made to the local (spacer grid) grid) pressure losses in COTRANSA2 appear appear in the adjustments to the orifice loss coefficient coefficient in XCOBRA-T.

The hydraulic channel nodalization nodalization of each code is discussed discussed in the previous response to RAI previous response SRXB-1 15 (ML082330187).

SRXB-115 NRC RAI SRXB-109 SRXB-109 Section 3.3 of the Technical Evaluation Report Technical Evaluation Report attached to the NRC's safety evaluation evaluation approving XN-NF-84-1 XN-NF-84-105(P)(A) 05(P)(A) states that critical power calculations calculations may inaccurate if be inaccurate if the the inlet flow is negative negative or if if the inlet quality is above zero. Verify that for the transient analyses analyses that the bundle inlet flow is positive positive and that the inlet qualities are less than zero.

Supplemental Response Supplemental Response to SRXB-109 SRXB-109 The transient code, XCOBRA-T, evaluates Reynolds Reynolds number for each node for each step of the the calculation.

calculation. If the flow becomes becomes negative at any node, the code stops the calculation.

SRXB-112 NRC RAI SRXB-112 Some models may have been updated to conservatively conservatively bound experimental experimental data collected data collected subsequent to the NRC review and approval of RODEX2. RODEX2. The staff notes that certain certain assumptions may be conservative assumptions conservative in the assessment of linear linear heat heat generation generation rate limits that may not be conservative conservative when evaluating transient transient heat flux during AOO simulation simulation due to the the E2-34

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION competing effects of reactivity feedback competing feedback and heat flux flow mismatch. If If aa model is "conservatively bounding" "conservatively bounding" in RODEX2, and in RODEX2, and translated translated to XCOBRA-T, provide provide a discussion of performance of the model for thermal the performance thermal margin transient calculations.

Clarificationsprovided Clarifications Providedby bv the NRC following a meeting on August 7, 7. 2008 The draft response for SRXB-1SRXB-112 12 deals with changes changes to the RODEX2 RODEX2 code in its first part, but requests additional additional information regarding the use of conservative assumptions in the abnormal conservative assumptions occurrence (AOO) transient response. The discussion regarding operating occurrence regarding the conservatism of the gap properties should be addressedaddressed in the response response to the second second part of RAI 112. See the the second and third sentences:

The staff notes that certain certain assumptions assumptions may be conservative conservative in the assessment assessment of linear linear heat generation rate limits that may not be conservative when evaluating evaluating transient heat flux flux during AOOAOO simulation simulation due to the competing competing effects reactivity feedback and heat effects of reactivity flux/flow heat flux/flow mismatch. If If a model is "conservatively bounding" in RODEX2, and translated to XCOBRA-T, provide provide a discussion discussion of the performance performance of the model for thermal margin transient calculations.

Summary Summary of staff concern:

The NRC staff considered the coupling of the neutron neutron flux and fluid conditions for AOO transienttransient evaluations evaluations for both aa reduced thermal time constant and an increased increased thermal time constant.

When the time constant is over predicted, the fluid response response to changing neutron neutron power is lagged. A pressurization pressurization transient, therefore, therefore, would result in an increase increase in the reactor reactor power that is not impeded impeded by subsequent subsequent rapid void formation due to hold up of the heat flux in the the pellet. An over prediction of the time constant will tend to increase increase the fission power for such aa transient. However, the same effect effect of holding the heat up in the fuel pellet has the dual effect of reducing the cladding heat flux response; therefore, the ultimate ultimate effect effect on the transient critical power ratio (CPR) is a combination of the conservative conservative prediction prediction of peak neutron flux with the the non-conservative prediction non-conservative prediction of the transient cladding heat flux.

For the case where the time constant is under predicted predicted the inverse inverse is true, the gross reactor power increase due to pressurization pressurization is limited limited due to more rapid void formation in response to the increasing increasing neutron neutron flux, but this is countered countered by a prediction prediction of higher cladding surface surface heat heat flux relative to the pin power throughout throughout the transient.

The input assumptions assumptions regarding the gas gap may increase increase or decrease decrease the thermal thermal resistance, and similarly, an increase increase or decrease decrease in the thermal thermal resistance does not have a clear impact on the transient predicted CPR due to competing effects effects in the cladding cladding heat flux and void void reactivity.

Supplemental Response to SRXB-112 SRXB-112 A gap conductance conductance sensitivity study was performed for the 100% 100% power/105%

power/1 05% flow BFN load load rejection with no bypass (LRNB) (LRNB) transient transient event from Reference SRXB-1 12.1. The purpose of Reference SRXB-112.1.

the sensitivity study was to show the ~CPR ACPR trend for changes in gap conductanceconductance for COTRANSA2 versus XCOBRA-T.

COTRANSA2 XCOBRA-T. The gap conductance conductance change considered considered was [ I].

The results are provided SRXB-1 12.1. As seen from the results, an increase in provided in Table SRXB-112.1. in COTRANSA2 core average COTRANSA2 average gap conductance conductance results in a decrease decrease in ~CPR; ACPR; whereas an increase in XCOBRA-T increase XCOBRA-T gap hot channel conductance conductance results in an increase in ~CPR.

increase ACPR. A E2-35

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION decrease decrease in gap conductance conductance shows the opposite trend. The XCOBRA-T ATRIUM-10 hot XCOBRA-T ATRIUM-10 channel model is slightly more sensitive to the change conductance than the change in gap conductance the COTRANSA2 ATRIUM-100 average core model. When both COTRANSA2 COTRANSA2 ATRIUM-1 COTRANSA2 and XCOBRA-T gap conductance conductance are changed by an equivalent amount, the net impact is no significant in significant change in flCPR.

ACPR.

Reference:

SRXB-112.1 SRXB-1 12.1 EMF-2982(P) Revision 0, Browns Ferry EMF-2982(P) Ferry Units and 3 Safety Analysis Report for Units 2 and Extended Power Extended A TRIUM TM-1O Fuel Uprate ATRIUMTM-10 Power Uprate Supplement, Framatome ANP, Fuel Supplement, June 2004.

Table SRXB-112.1 SRXB-112.1 Gap Conductance Conductance Study Study Increase in Gap Conductance Increase Conductance Gap conductance conductance condition A(ACPR) fl(flCPR)

Core average average(( ] -0.011

-0.011 Hot channel channel [ )) +0.012

+0.012 average and hot channel [

Core average ] 0.000 Decrease in Gap Conductance Gap Conductance Gap conductance conductance condition condition A(ACPR) fl(flCPR)

Core average average(( ] +0.015

+0.015 Hot channel channel [ ] -0.016

-0.016 average and hot channel ((

Core average ] -0.001 NRC RAI SRXB-116 SRXB-116 Address Address whether XCOBRA-T was used to demonstrate acceptable fuel rod thermal mechanical demonstrate acceptable performance during transients.

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

Clarifications Provided bv the NRC following a meeting on August 7, provided by 7, 2008 describing the method for normalization Aside from describing normalization of the transient LHGR LHGR to the initial initial LHGR, provide some additional minor clarifications:

(1) The decay heat contribution will remain essentially static during the transient, address address whether the normalization whether normalization capture capture the varying varying rod decay decay heat sources; (2) Specify Specify the source of the decay decay heat constants (i.e. ANS standard);

constants (i.e.

E2-36

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION (3) The rod power distribution is flattened gamma smearing of the thermal power, flattened due to gamma address how these gamma smeared power fractions address fractions are calculated; calculated; and (4) Address Address how the direct moderator moderator heatheat is accounted for.

The response response should also provide a detailed description description of the rod heat flux calculation for for bundles bundles with part length fuel rods, and address address the code change as well as items 1-4 1-4 for each region (fully rodded, plena region, above above plena region).

Supplemental Response Supplemental Response to SRXB-116 SRXB-116 For bundles with part-length fuel rods (PLFRs), the rod heat flux calculation calculation begins begins by computing time-dependent heat flux generation computing the time-dependent generation rate at each axial section section in the fuel rod.

The updated corresponding to equation 2.130 of Reference updated equation corresponding Reference SRXB-116.1 SRXB-1 16.1 is:

q"(t) = P(t) I (ff + fc )FriFliFa 7FDrodj LNaNn where where P(t) pet) == transient reactor power power fff, =

fraction of power produced in the fuel fýfc

= fraction of power produced in the claddingcladding Na == assemblies in the core total number of assemblies core Nri N,. == total number of heated heated rods for type i assembly assembly at the axial plane plane Fri F; == radial peaking factor of type i assembly assembly F/i F,, == local peaking factor of type ii assembly Fa Fa == axial peaking factor at the axial plane plane Drodj Orad'; == fuel rod diameter diameter of type i assembly L == axial heated length length This equation differs from that in Reference Reference SRXB-116.1 SRXB-1 16.1 by replacing replacing the initial reactor power in the denominator with TT. Tr. In addition, the variable definitions have have been modified modified to identify identify that the total number of heated heated rods is dependent dependent on both the assembly assembly type and axial elevation, elevation, and the definition definition of LL has been corrected to the axial heated heated length of the assembly. This This equation is substituted into equations equations 2.129a 2.129a and 2.129b 2.129b in Section Section 2.5.5 of Reference Reference SRXB-1 16.1 to define the volumetric heat SRXB-116.1 heat deposition deposition rate for the fuel pellet and cladding, respectively. This volumetric deposition rate is used in the right hand side of equation volumetric heat deposition equation 2.85 of Reference SRXB-1 Reference SRXB-116.1 16.1 to iteratively iteratively solve the transient heat conduction equation and the the hydraulic hydraulic conservation equations for the new time step temperatures temperatures and surface heat flux. The The heat flux is introduced into the channel energy equation (2.2 of Reference SRXB-1 16.1)) through Reference SRXB-116.1 through the term q'. q'. This linear heat deposition rate is a summation summation of the energy energy added by direct energy deposition deposition and surface heat flux:

q ,(t) q'(t)

J=

1P(t) fcoolFriFa+Hsurf"(TNodesT-Tfluid)'%'Drod,i"Nri

= {P(t)

-NaL- fcool FriFa + Hsurf . (TNodesT - Tf/uid)*ff* Drad i . Nri }

Ni I Ni NLa '

E2-37

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION where where

f. 0 1o fcool = fraction of power power produced produced in the coolant coolant HSurd Hsurf = heat transfer coefficient at the axial plane film heat plane TNodesT = = temperature at the axial plane cladding surface temperature plane Tr7,ud Tf/uid =

= temperature at the axial plane fluid temperature plane Ni N, = number assemblies in channel ii number of fuel assemblies In addition to axially varying number of heated proper modeling of PLFRs also requires heated rods, proper requires variations in the active flow area, the heated axial variations heated perimeter, and the wetted perimeter, and parameters are now defined as axially dependent these parameters dependent quantities quantities in AREVA methods.

references to these parameters parameters derived from the basic Consequently, all references parameters or parameters basic geometry data data in the approved topical reports should be interpreted as being being axially dependent dependent variables. The pressure drop due to the area expansion at the end of the PLFRs (or anywhere area expansion modeled using the specific volume for momentum as expressed in the active flow path) is modeled expressed inin Reference SRXB-1 equations 2.78 and 2.79 of Reference SRXB-116.1. contractions 16.1. For current designs, area contractions occur in the single phasephase region, but the coding was generalized address area contractions generalized to address contractions in in the two-phase region based on a solution of the two phase phase Bernoulli equation.

XCOBRA-T deposited power fraction sensitivity An XCOBRA-T sensitivity study was performed performed for the 100%

100%

power/1 power/105% 05% flow BFN LRNB transient event from Reference SRXB-1 16.2. The purpose of the Reference SRXB-116.2. the sensitivity study was to show the impact on L\CPR ACPR from using using generic ATRIUM-1 0 power generic ATRIUM-10 power case-specific power fractions versus case-specific case-specific power power fractions. The case-specific power fractions are used in in CASM0-4/MICROBURN-B2. AREVA is in the process of COTRANSA2 and are obtained from CASMO-4/MICROBURN-B2.

COTRANSA2 automating the transfer automating case-specific power fractions into XCOBRA-T transfer of the case-specific XCOBRA-T such that the the generic values will no longer be used.

generic used. [

] The power that would have been deposited ((

].]. A review of an ATRIUM-10 ATRIUM-10 power showed that the ((

deposition study showed power deposition

]. A study was was performed performed by taking [ ]. The results are provided provided in Table SRXB-1 16.1. The study shows no significant Table SRXB-116.1. significant change in ACPR.

L\CPR. ((

] This study demonstrates that the ATRIUM-10 demonstrates generic power fractions in XCOBRA-T ATRIUM-10 generic XCOBRA-T are adequate.

References:

SRXB-116.1 SRXB-116.1 XN-NF-84-105(P)(A) Volume XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements Supplements 1 and 2, XCOBRA-T:

Computer Code for BWR A Computer BWR Transient Transient Thermal-Hydraulic Thermal-HydraulicCore Exxon Core Analysis, Exxon Nuclear Nuclear Company, February February 1987.

SRXB-1 SRXB-116.2 16.2 EMF-2982(P) Revision EMF-2982(P) Ferry Units 2 and Revision 0, Browns Ferry and 3 Safety Analysis Report Report for Extended Power Uprate A Power Uprate TRIUM TM-1O Fuel ATRIUMTM-10 Fuel Supplement, Framatome ANP, Supplement, Framatome (ML041840301)

June 2004. (ML041840301)

E2-38 E2-38

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY Table SRXB-116.1 SRXB-116.1 Deposited Heat Heat Study Fuel Cladding Moderator Moderator Bypass Bypass Condition Heat Heat Heat Heat Heat A(ACPR) fl(flCPR)

Generic power power fractions [ ] (( ] [ ] [ ] NA NA Case-specific power fractions [ ] [ ] [ ] [ ] -0.0004

-0.0004 Case-specific power fractions fractions

[I ] [ ] [ ] [ ] [ ] +0.0008

+0.0008 E2-39 E2-39

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

Given the information provided in the NRC's contractors' technical technical evaluation evaluation report attached to the safety evaluation approving NEDO-32047-A safety evaluation NEDO-32047-A dated February February 5, 1994, Appendix C: C:

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

NEDO-32047-A.

The analyses analyses in NEDO-32047-A NEDO-32047-A were performed performed for General General Electric Electric (GE) fuel. The analyses analyses are generally generally applicable for pre-EPU pre-EPU core designsdesigns since hydraulic stability of the fuel products products has has improved improved or at least remained the same. Provide a comparison comparison of the channel channel stability characteristics characteristics of ATRIUM-10 ATRIUM-10 to GE 8x8 fuel. If If ATRIUM-10 ATRIUM-10 is less stable than GE 8x8 fuel, consequences of a non-isolation consider any impact on the projected consequences consider non-isolation ATWS instability event.

Response

Response to SRXB-1 SRXB-117 17 The pre-EPU stability stability analysis for BFN indicates that the global mode mode is dominant dominant over the the regional regional (out-of-phase) modemode where relatively large subcritical reactivity values are calculated with STAIF. For EPU cores with flatter radial power distributions, the calculated calculated subcritical reactivity reactivity values are noticeably lower in comparison. The resulting regional decay ratios decay ratios calculated calculated for the EPU core are larger than the corresponding corresponding globalglobal mode mode decay decay ratio in a minority minority of cases, which warrants the examination examination of the effect of regional mode oscillations oscillations dominating dominating postulated ATWS instability events.

The task of evaluating the impact of large regional regional versus global mode mode oscillations is first addressed addressed below from an analytical analytical point of view and calculations are presented using using aa reduced reduced order model. The calculations will also address address the effects of the parameters parameters of interest, namely namely the subcritical subcritical reactivity reactivity due to core radial power flatteningflattening for EPU, increase increase inin voidreactivity coefficient void reactivity coefficient due to increasing increasing the fresh fuel batch size, and fuel geometry effects geometry effects (part-length rods and reduced pin conduction time constant for an ATRIUM-10 ATRIUM-10 compared with with an 8x8 fuel bundle). These effectseffects will be demonstrated demonstrated to result in equivalent consequences consequences of a postulated ATWS event relative to the results in NEDO-32047-A NEDO-32047-A (Reference SRXB-1 17.1).

(Reference SRXB-117.1).

mitigation of the ATWS instability by reducing the core inlet subcooling as a Furthermore, the mitigation consequence consequence of water level reduction by operator operator action (Reference SRXB-1 action (Reference SRXB-117 17.2) be

.2) will be demonstrated demonstrated to be as effective in suppressing suppressing regional mode oscillations oscillations as for global mode mode oscillations.

oscillations.

Analytical Considerations Analytical Considerations Unstable global mode oscillations grow exponentially Unstable exponentially at a fixed rate (decay ratio) from a small perturbation. As the oscillation magnitude perturbation. magnitude increases, nonlinear effects become become important. The The E2-40 E2-40

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION average power level drifts to higher values as a consequence average power consequence of the nonlinearity neutron nonlinearity of the neutron kinetics, which results in a negative reactivity feedback due to the increaseincrease of void fraction. TheThe negative reactivity superimposed on the oscillating reactivity reactivity superimposed reactivity results in damping the neutron neutron kinetics (References (References SRXB-117.3. SRXB-117.4, and SRXB-117.5). ((

SRXB-117.3. SRXB-117.4, The regional mode mode oscillations are well understood understood in the linear limit where the power oscillation power oscillation is attributed to the excitation azimuthal harmonic excitation of the first azimuthal harmonic mode mode of the neutron flux.

Compared with the fundamental fundamental flux mode excitation associated associated with the global oscillation, thethe subcritical reactivity of the first azimuthal eigenfunction contributes a damping effect on the eigenfunction contributes the neutron kinetics feedback. The hydraulic response is less damped neutron damped compared compared to the global mode case case due to bypassing the dampingdamping effects effects of the recirculation recirculation loop. The regional modemode oscillations may become the preferred preferred oscillation oscillation mode for large-orificed large-orificed cores (hydraulic (hydraulic destabilization) and for small radial buckling destabilization) buckling (large core core diameter and radial power distribution power distribution that is relatively relatively flat or ring-of-fire with relatively low power in the center).

[

E2-41

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY I]

Description of the Reduced Description Reduced Order Model description of large phenomenological description The phenomenological modes large power oscillations in the global and regional modes which is used here to simulate is supported by the results of a reduced order model, which simulate large large global and regional regional mode oscillations. [

E2-42 E2-42

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION

]

The reduced reduced order model allows fast and robust simulation simulation of both the global and regional modes and helps to resolve resolve issues that were were not apparent at the time NEDO-32047-A NEDO-32047-A (Reference SRXB-117.1)

(Reference SRXB-117.1) was issued. Most importantly, itit helps to explore explore and provide provide insight into the differences differences between between the global and regional mode oscillations and their common mechanism.

ultimate limiting mechanism.

Results Results The results of several cases performed performed with the reduced order model are presented.

presented. All of these calculations calculations represent unstable unstable oscillations oscillations growing to large magnitudes with parameter parameter variations to address address the issues of global versus regional and the effect effect of EPU core loading loading with fuel design differing from the fuel type used in NEDO-32047-A NEDO-32047-A (Reference SRXB-117.1).

SRXB-1 17.1).

These cases are:

E2-43

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION I]

Conclusions Conclusions

• • Large regional mode oscillations oscillations have have (( ] effects compared compared with global mode.

• " ATRIUM-10 ATRIUM-10 bundle design differences differences from an older 8x8 [

]I

• coefficient) ((

EPU effects (lower subcritical reactivity and higher void reactivity coefficient)

))

• • [

I

References:

SRXB-1 17.1 SRXB-117.1 NEDO-32047-A, NEDO-3204 "ATWS 7-A, "A TWS Rule Issues Issues Relative to BWR Core Thermal-Hydraulic Thermal-Hydraulic Stability," June 1995.

SRXB-1 17.2 SRXB-117.2 NEDO-32164 NEDO-32164 Revision Revision 0, "Mitigation "Mitigation of BWR Core Thermal-Hydraulic Instabilities Thermal-Hydraulic Instabilities in ATWS," December 1992.

December 1992.

SRXB-1 17.3 SRXB-117.3 Wulff, W., H.H. S. Cheng, A.N. Mallen, and U.S. Rohatgi, "BWR Stability Analysis Stability Analysis with the BNL Engineering Engineering Plant Analyzer,"

Analyzer," NUREG/CR-5816, NUREG/CR-5816, October 1992.

October 1992.

SRXB-1 17.4 SRXB-117.4 March-Leuba, J., D.G. Cacuci, and R.B. Perez, "Nonlinear Dynamics March-Leuba, Dynamics and Stability of Boiling Water Reactors:

Reactors: Part I -- Qualitative I Qualitative Analysis," Nuclear Nuclear Science and Engineering:

Science Engineering: 93, 111-123 (1986).

93,111-123 SRXB-1 17.5 SRXB-117.5 March-Leuba, Instabilities in Boiling Water Reactors,"

March-Leuba, J., "Density-Wave Instabilities NUREG/CR-6003, September NUREG/CR-6003, September 1992.

E2-44

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY SRXB-117.6 SRXB-1 17.6 Farawila, Y.M.,

Y.M., and D.W. Pruitt, "A Study of Nonlinear Nonlinear Oscillation Oscillation and Limit Limit Cycles in Boiling Water Reactors Reactors --I:I: The Global Mode," Nuclear Nuclear Science Science and Engineering:

Engineering: 154, 302-315 302-315 (2006).

SRXB-117.7 SRXB-1 17.7 Y.M., and D.W. Pruitt, "A Farawila, Y.M., "A Study of Nonlinear Oscillation and Limit Nonlinear Oscillation Limit Cycles in Boiling Water Water Reactors Reactors --II:

I1: The Regional Regional Mode," Nuclear Nuclear Science and 154, 316-327 (2006).

Engineering: 154,316-327 Engineering:

E2-45 E2-45

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

r-

..J SRXB-1 17.1.1 Figure SRXB-117 .1.1 Relative Relative Power Power for Case 1 Oscillation Base Global Oscillation rr..

..J SRXB-117.1.2 Figure SRXB-117 .1.2 Relative Power for Case 2 Base Regional Oscillation Oscillation E2-46 E2-46

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

r-

.J Figure SRXB-117.1.3 SRXB-117.1.3 Relative Relative Power Power for Case 3 Global Oscillation Oscillation rr-Figure SRXB-117.1.4 SRXB-117.1.4 Relative Power for Case 4 Regional Oscillation Regional Oscillation E2-47

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY rr-

.J Figure SRXB-117.1.5 SRXB-117 .1.5 Relative Relative Power for Case 5 Oscillation With Decreased Regional Oscillation Decreased Subcriticality Subcriticality rr-a-

.J Figure SRXB-117 SRXB-117.1.6 Relative Power

.1.6 Relative Power for Case 6 Mitigated Mitigated Global Global Oscillation E2-48 E2-48

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-Figure SRXB-117.1.7 SRXB-117 .1. 7 Relative Power for Case 7 Mitigated Regional Oscillation Mitigated Oscillation rr" SRXB-1 17.1.8 Figure SRXB-117 .1.8 Relative Relative Power for Case 8

..J Late-Mitigated Oscillation Late-Mitigated Global Oscillation E2-49

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION r

r-

.J SRXB-1 17.1.9 Figure SRXB-117 Relative Power

.1.9 Relative Power for Case 9 Late-Mitigated Regional Oscillation Late-Mitigated rr-

..J

.2.1 Inlet Mass Flow Rate for Case I1 SRXB-117.2.1 Figure SRXB-117 Base Global Oscillation E2-50 E2-50

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-

-U Figure SRXB-117.2.2 SRXB-117.2.2 Inlet Mass Mass Flow Rate for Case 2 Oscillation Base Regional Oscillation r

r-SRXB-117.2.3 Figure SRXB-117 Inlet Mass Flow Rate for Case 3

.2.3 Inlet J

Global Oscillation E2-51

r r- NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION

..J SRXB-117.2.4 Inlet Mass Figure SRXB-117.2.4 Mass Flow Rate for Case 4 Regional Regional Oscillation rr-

..J Figure SRXB-117.2.5 SRXB-117.2.5 Inlet Mass Flow Rate for Case 5 Regional Regional Oscillation Oscillation With Decreased Decreased Subcriticality Subcriticality E2-52

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

r-SRXB-117.2.6 Inlet Mass Flow Rate for Case 6 Figure SRXB-117.2.6 Mitigated Global Oscillation Mitigated Global Oscillation rr-

..J Figure SRXB-117.2.7 Inlet Mass Flow Rate for Case 7 SRXB-117.2.7 Inlet Mitigated Oscillation Mitigated Regional Oscillation E2-53

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION r

r..

rn-

..J Figure SRXB-117.2.8 Inlet Mass Figure SRXB-117.2.8 Mass Flow Rate for Case 8 Late-Mitigated Global Oscillation Late-Mitigated Oscillation r

Figure SRXB-117.2.9 Inlet Mass Figure SRXB-117.2.9 Mass Flow Rate for Case 9 Late-Mitigated Regional Late-Mitigated Regional Oscillation Oscillation E2-54 E2-54

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION r

r.

.J Figure SRXB-117.3.1 Fraction for Case 1 SRXB-117 .3.1 Exit Void Fraction Oscillation Base Global Oscillation r

r..

.J Figure Figure SRXB-1 17.3.2 Exit Void Fraction SRXB-117.3.2 Fraction for Case 2 Regional Oscillation Base Regional Oscillation E2-55 E2-55

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

r-Figure SRXB-1 17.3.3 SRXB-117 .3.3 Exit Void Fraction Fraction for Case 3 Global Oscillation rr-

..J Figure SRXB-117.3.4 Exit Void Fraction Figure SRXB-117.3.4 Fraction for Case 4 Regional Regional Oscillation Oscillation E2-56

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-rn-Figure SRXB-117.3.5 Figure SRXB-117.3.5 Exit Void Fraction Fraction for Case 5 Regional Oscillation Oscillation With Decreased Decreased Subcriticality Subcriticality rr" rn-

.J SRXB-1 17.3.6 Exit Void Fraction for Case Figure SRXB-117.3.6 Case 6 Mitigated Global Oscillation E2-57

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

r.

Figure SRXB-1 17.3.7 Exit Void Fraction SRXB-117.3.7 Fraction for Case 7 Mitigated Mitigated Regional Regional Oscillation Oscillation rr..

-d

..J SRXB-1 17.3.8 Exit Void Fraction for Case 8 Figure SRXB-117.3.8 Late-Mitigated Regional Late-Mitigated Regional Oscillation Oscillation E2-58

rr- NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure SRXB-117.3.9 SRXB-117.3.9 Exit Void Fraction Fraction for Case 9 rr- Late-Mitigated Regional Oscillation Late-Mitigated Oscillation Figure SRXB-117.4.1 SRXB-117.4.1 Void Fraction in Selected Nodes for Case 1 Selected Nodes Base Global Oscillation Oscillation E2-59 E2-59

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-

.J Figure SRXB-117.4.2 Void Fraction in Selected Nodes for Case 2 Base Regional Oscillation Oscillation r

r-

.J SRXB-117.4.3 Void Fraction Figure SRXB-117.4.3 Fraction in Selected Selected Nodes for Case 3 Oscillation Global Oscillation E2-60

r-r NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION

..J SRXB-117.4.4 Void Fraction Figure SRXB-117.4.4 Fraction in Selected Selected Nodes Nodes for Case 4 Oscillation Regional Oscillation rr SRXB-117.4.5 Void Fraction Figure SRXB-117.4.5 Fraction in Selected Nodes Nodes for Case Case 5 Regional Oscillation With Decreased Decreased Subcriticality E2-61

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION NRC NRC Introduction Introduction The following are related to the June 3, 2008 response to SRXB-88. SRXB-88.

NRC SRXB-118 NRC RAI SRXB-118 In the supplemental supplemental response to RAI SRXB-88, TVA provided the results of sensitivity analyses analyses to evaluate uncertainty on the calculated delta-critical evaluate the impact of void fraction uncertainty delta-critical power ratio ratio (DCPR)

(DCPR) and the safetysafety limit minimum critical power power (SLMCPR). In In the void fraction fraction reduction case, the DCPR apparently unaffected DCPR is apparently accompanied by an increase in SLMCPR.

unaffected and is accompanied If the void fraction were reduced If reduced throughout throughout the core by a fixed bias, the result would be to redistribute the reactor reactor power according to the change according change in reactivity reactivity associated associated with the voidvoid perturbation.

perturbation. Since those bundles with the higher bundle average void fractions will have have a greater greater reactivity response, a reduction in the void fraction will tend to increase, slightly, the the power in those bundles with a higher higher bundle average void fraction fraction relative to the bundles that had a lower void content content prior to the perturbation.

perturbation. The bundles with a higherhigher bundle average average void fraction are the high powered bundles. Therefore, Therefore, a fixed reduction in void fraction will increase the radial power peaking factor. The increased radial power power peaking factor for aa given steady 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 effect is considered, 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.

distribution. In effect, the span of pin powers powers to account for the 0.1 percent of highest powered powered pins increases. Results of the TVA sensitivity sensitivity analysis demonstrate demonstrate the opposite opposite trend. It is expected expected that the imposition of aa fixed void fraction reduction would result in a lower SLMCPR.

Explain this discrepancy.

Response to SRXB-1 SRXB-118 18 It should be noted It noted that the sensitivity analyses presented in the SRXB-88 response were not analyses presented not based on "fixed" void fraction changes. Rather, the analyses were based on modifications modifications to the void-quality correlation that resulted in a new nominal fit and offsets that were on average average

+0.05 void. The discussion

+/-0.05 discussion above in the RAI question for SRXB-118 SRXB-1 18 is based on a comparison comparison of trends trends for an instantaneous instantaneous change in void fraction.

fraction. The RAI SRXB-88 response included included the the impact of the fuel depleted depleted with the changes changes in the void-quality correlation. The difference difference in depletion depletion changes changes the sensitivity of void friction modifications modifications considerably considerably due to the feedback of modified modified power distributions distributions on exposure exposure distribution.

For the RAI SRXB-88 case, the change change in the void-quality void-quality correlation correlation was imposed over all fuel fuel in the core from beginning-of-life.

beginning-of-life. No changes were made to the fuel loading and rod patterns.

The result of SRXB-88 SRXB-88 was that aa reduction in void resulted in more assembliesassemblies at higher higher peaking factors of the high-powered power. The radial peaking high-powered assemblies that contributed to rods in boiling transition "flat" and resulted in a slightly higher SLMCPR.

transition were slightly more "flat" Figure SRXB-1 SRXB-118.1 differences in radial distributions.

18.1 shows the slight differences The sensitivity analysis analysis of SRXB-88 repeated for an instantaneous SRXB-88 was repeated instantaneous change in voids. For an instantaneous change instantaneous change in voids, the SLMCPR SLMCPR trends were were the same as SRXB-88; SRXB-88; however, the the change is small for both depleted and instantaneous instantaneous void change, i.e., an SLMCPR SLMCPR changechange of

-0.003

-0.003 for +0.05 voids and +0.002 for -0.05 -0.05 voids. The sensitivity can be explained by the the E2-62

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION small radial power distribution distribution shifts in Figures SRXB-1 18.2 and SRXB-118.3.

Figures SRXB-118.2 SRXB-1 18.3. ItIt is concluded that the radial distribution is not significantly significantly changed changed for the SLMCPR SLMCPR analysis; the analysis; therefore, the impact impact of the prescribed void-quality correlation prescribed void-quality correlation changes insignificant on SLMCPR.

changes is insignificant It is very difficult It difficult to identify the expected direction direction of the radial power distribution change due to to a modification of the void-quality void-quality correlation.

correlation. In addition to the void coefficient coefficient dependency dependency on void fraction, there is an even stronger dependency dependency of the void coefficient coefficient on exposure. For the the limiting SLMCPR the highest limiting case of SLMCPR highest radial powers come from a range of assembly assembly exposures.

The importance of void changes in different assemblies of different different exposures exposures cannot be be analyzed with simplified models and isolated trends.

analyzed Independent of the trend, the analyses demonstrate insignificant impacts on SLMCPR.

Independent r

E2-63 E2-63

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 0.99

~° 0.98 OJ

~

]

OJ E 0.97 o

0 z

Reference E

- - Modified

- - - - - . Modified +0.05 0.96 - - - - Modified -0.05 0.95 L -_ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ ___ ~

0.95 o

0 100 100 Bundle Index Figure SRXB-118.1 SLMCPR Radial Power Distribution SRXB-118.1 SLMCPR Distribution High-Powered Assemblies High-Powered Assemblies Depleted Depleted Voids Voids

\

\

'\. n\

0.99

~ 0.98

a. 0.98 OJ

~

.~

~* 0.97 ~ ---~~- -------

z 0~

Z

- - Modified Reference~

- - - Reference Modified 0.96 t ......

Modified

- - - - - -Modified +0.05

+0.05

- --- - - Modified -0.05 1---------------

0.95 I - - - - - -- - - - - - - - -- - - - - - -- - - - - -- - - - - - - ----l o0 100 100 Bundle Index Figure SRXB-118.2 SRXB-118.2 SLMCPR Radial Power Distribution High-Powered Assemblies High-Powered Assemblies Instantaneous Voids Instantaneous Voids E2-64 E2-64

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION 0.8 0.8 , -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - ,

0.7 f 0.6 a.

0 II.

iii 0.5

'6 (i

~ 0.4.

04 E

0o z

1

---

Reference Reference 0.3

- - ModifiedModified

- - - - - - Modified +0.05

- - - Modified -0.05

-0.05 0.2 -

0.1 0.1 ~--------------~--------------~----------------~------------~

400 400 500 600 700 800 Bundle Bundle Index Index Figure SRXB-118.3 SRXB-118.3 SLMCPR Radial Power Distribution Low-Powered Assemblies Low-Powered Assemblies Instantaneous Voids Instantaneous Voids E2-65

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NRC RAI SRX8-119 SRXB-119 Continuing with the void fraction reduction case, the decrease Continuing decrease in void fraction would simultaneously simultaneously result in aa redistribution of the axial power. Since those higher void nodes nodes would have a greater reactivity 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 upwards effect of increasing increasing the reactor adjoint in the upper upper portions portions of the core. As pressurization pressurization transients are typically typically

limiting, limiting, the impact of an upwardupward shift in axial power on the transient transient power power prediction prediction should be be considered.

considered. The upward upward shift in reactor reactor adjoint directly directly affects affects the core void reactivity reactivity coefficient coefficient and tends to increase increase the sensitivity of the core reactivity reactivity to aa pressure wave, since the back pressure pressure wave is dissipated by void collapse collapse in the upper upper parts of the core. Therefore, the core core wide transient power transient power would be increased increased relative to the base case, which which appears to result in an increase in the DCPR.

increase The results of the TVA sensitivity analysisanalysis do not demonstrate demonstrate this trend. Address why imposing a fixed void fraction reduction reduction does not result in a higher DCPR.

Response

Response to SRXB-1 SRX8-119 19 The trend described described above in the RAI questionquestion for SRXB-119 SRXB-1 19 is for an instantaneous instantaneous change in in discussed in the previous response to SRXB-118, voids. As discussed SRXB-1 18, the results of the SRXB-88SRXB-88 sensitivity sensitivity analyses analyses werewere based on fuel depleted depleted with the changes changes in the void-quality correlations.

correlations. AREVA concurs concurs with the general general trend as described described above for an instantaneous instantaneous change change in void. The analysis of SRXB-88 SRXB-88 was repeated repeated for an instantaneous change in voids.

Relative to the Reference Reference case, the change in .'1CPR ACPR was -0.002

-0.002 for +0.05 voids and +0.01 +0.01 for

-0.05

-0.05 voids, which is consistent with the staff's observations.

observations.

NRC RAI SRXB-120 SRX8-120 The void increase increase cases exhibited exhibited opposite trends relative to the void reduction cases. The The staff found that the void reduction reduction cases were not consistent with the staff's staffs expectations.

Provide information information similar to the information information requested requested in SRXB-1 SRXB-118 18 and 119 for the fixed increase in void fraction sensitivity increase sensitivity analyses.

For each case in Study 1 provide:

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

• " Plots of the perturbed perturbed axial and radial core power shape shape

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

• • Plots of transient critical CPR

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

Response to SRX8-120 SRXB-120 For an increase increase in void fraction, the responses to SRXB-1 SRXB-118 18 and SRXB-119 SRXB-1 19 provide the provide the requested information.

information. That is, the void trend was explained explained and the change did not result in a significant impact impact to the SLMCPR SLMCPR and the transient analyses are consistent with the staff's staff's expectations when instantaneous expectations instantaneous voids are considered.

considered.

E2-66

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Below are the requested results for Study 1 and are based based on depleting depleting the fuel with the change change in the void correlations (the results are not for an instantaneous instantaneous change in the void correlations).

• • The limiting bundle: core location location and initial radial peaking peaking factor SXRB-1 20.1).

factor (Table SXRB-120.1).

• • Initial axial power (Reference case) (Figure SXRB-120.1).

power shape (Reference SXRB-1 20.1).

• • Plots of the perturbed perturbed axial power power shapes (initial conditions) (Figure SXRB-1 20.1).

SXRB-120.1).

• • perturbed radial core power shapes (initial Plots of the perturbed (initial conditions) (Figure SXRB-120.2).

SXRB-120.2).

• • Plots of transient limiting bundle peak rod heat flux at axial heights (x/L) of 25%, 50%,

and 75% (Figures SXRB-120.3 SXRB-1 20.3 to -120.5).

-120.5).

• • Plots of Plots of transient transient limiting limiting bundle bundle mass mass flow rate at flow rate at axial axial heights heights of 0%, 25%,

of 0%, 25%, 50%,

50%, 75%

75%

and 100%

100% (Figures (Figures SXRB-120.6 to -120.10).

• • Plots of transient CPR (Figure SXRB-1 20.11).

SXRB-120.11).

• • A comparison comparison of the predicted power pulse heights and widths (Table SXRB-120.2). SXRB-120.2).

E2-67 E2-67

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Table SXRB-120.1 SXRB-120.1 Limiting Bundle Data Data Study 1 Study 1 Study 1 Modified Modified Modified Modified Modified Modified Reference Reference v-a V-Q v-a V-Q v-a V-Q Calculation Calculation (-0.05) (0.0) (+0.05)

Location Location in Core I,J W,J 23,24 21,22 21 ,22 23,24 23,24 37,24 MB2 initial radial peaking factor 1.314 1.295 1.295 1.329 1.346 1.346 XCT converged initial peaking factor radial 4 (( )) (( )) (( )) (( ]

REGIONS REGIONS lR:

M1 1 3579U13 3 5 7 9 11 13 15 15 17 17 1919 21 23 25 27 'Z7 29 31 31. 33 33 35 35 37 39 41 43 45 47 49 51 51 53 55 57 59 59 JR:

~: WA H

60 ro 1 2 33 44 5 66 7 7 88 99 10 ro 11 12 U 13 14 M 58 15 16 17 18 19 20 20 21 22 23 24 25 26 'Z7 27 28 29 30~

56 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 31.~33~35~37~39~41~43"45~~484950 54 51 52 53 54 55 56 57 58 59 6 61 62 63 64 65 66 67 68 69 70 5152535455565758~ro~~~~~~~m~w 52 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 n~UN~~TI~~~~~~~~~~~~~~~

50 93 94 95 96 97 98 99100 101 102 103104 105 106107 108 109110 11132 113 114 115116117 118

~~~%~~~~www~~~m~Dillillrumillllillimm 48 119 120 121 122 123 124 125 126 371 28129 130131 1 133 134 135 136 137 138 139 140 141 142 143144 mmmmmm~~m~~rnm~m~~~m~m~ill~ill~

46 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 16W 161 162 163 164 165 166 167 168 169 170 171 172

~~rn~w~m~m~~5m55~m~m~~~~~~mrnm

" 173 174 175 176 177 178 179 180 181 182 183 184 185 186 187 188 189 190 191 192 193 194 195 196 197 198 199 200 201 202 mrn~mrnrnmwm~~~~~w~~mm~m~~~m~m~D~

~ 203 204 205 206 207 206 209 210 211 212 213 214 215 216 217 218 219 220 221 m2 223 224 225 226 227 228 229 230 231 23M

~~~~w~~~mmm~~mm~mmmmm~~~m~w~m~

~ 233 234 235 236 237 238 239 240 241 242 243 244 245 246 247 248 249 250 251 252 253 254 255 256 257 258 259 26D 261 262 m~~~m~~wm~w~~~w~~~&~m~~~~~~~E~

~ 263 264 265 266 267 26B 269 270 271 272 273 274 275 276 277 278 279 280 281 282 283 284 285 286 287 288 289 290 291 2W

~~~~w~~mmmmmmmmmm~~~~~~~w~~mm~

~ 293 294 25 296 297 298 299 300 301 302 303 304 305 306 307 308 309 310 311 312 313 314 315 316 317 318 319 310 321 3 m~~~m~mmE~m~~~Enmmmmm~mmm~m~m~

~ 323 324 325 326 3V7 328 329 330 331 332 333 334 335 M6 337 338 339 340 341 342 343 344 345 346 347 348 349 350 351 3M2

~~~~E~~Em~m~~~m~~~~~m~~~~~~~E~

~ 353 354 355 356 357 358 359 36D 361 362 363 364 365 366 367 36B 369 370 371 372 373 374 375 376 377 378 379 380 381 382 E~~~E~E~E~E~~~E~~mmmmmmmmmm~E~

~ 383 384 385 386 387 388 389 390 M1 3W E~~~E~~~E~m~m~m~m~G~~~~~~Q~mmm 394 3 396 397 398 399 400 40142 403 404 405 407 408 409 410 41142

~ 413 414 415 416 417 418 419 420 421 422 423 424 425 426 427 428 429 430 431 43V 433 434 435 436 437 438 439 440 441 442 m~m~m~~~m~~~~~mQ~~m~m~~~m~~~w~

~ 443 444 445 446 447 448 449 450 451 452 453 454 455 456 457 458 459 460 461 462 463 464 465 466 467 468 469 470 471 472

~~~~W~~~~~~~~6~~~~~~~~~~~~$mmm

~ 47 474 475 C76 477 478 479 480 481 482 483 ~

mmmmmmm~~~~ 484485 486 487 488 489 490 m

~~~~~~ 491.49Q 493 494 455 496 497 496 499 500 501 502

~~~~~~~~~D~

22 503 504 505 506 507 508 509 510 Sf11512 ~

~~~~~~~~~~ 513 ~~~m~~~2~~~~~~~~~m~

514 W15 516 517 518 519 520 521 =2 52 524 525 526 527 528 529 530 531 531 20 533 534 535 536 537 538 539 540 541 542 543 544 545 546 547 548 549 550 551 552 55 554 555556 557 558 559 56W 561 562 m~~~m~~~~~~~~~~~~~~~~~~~~~~~~~

18 563 564 565 566 567 568 569 570 571 572 573 574 575 576 577 578 579 580 581 582 583 584 585 586 587 588 589 590 591 592

~~~~~~~mmmm~~~m~m~~~~~~~~~~~s~

~ 593 594 595 596 597 598 599 600 601 602 603 604 605 606 W7 608 6W9 610 611 612 613 614 615 616 617 618 619 60

~~~~~~m~~~~~~~~~~~mrururum~m~~~

M 621 622 623 624 625 626 6V7 628 629 630 631 632 633 634 635 636 67 638 639 640 641 642 643 644 645 646

~~~~~~~~~~~~m~~~~~~~~~oo~~~

U 647 648 649 650 61 652 653 604 655 656 657 658 659 66W 661 662 663 664 665 666 667 669 669 670 671 672

~~~~~~~~~~~~~~~~~~~~~~~~mrn ro 673 674 675 676 677 678 679 680 681 682 683 684 685 686 687 688 689 690 691 692 693 694 rnrne~m~~~~~~~~~~~~~~~~~

8 695 696 697 698 699 700 701 702 703 704 705 706 707 708 709 710 711712 713 714

~~~~~DUU~~~DwnDmmmmm 6 715 716 717 718 719 720 721 722 723 724 725 726 727 728 729 730 731 73V 733 734 mmmmmmmmmm~mmmmmmmm~

4 735 736 737 738 739 740 741 742 743 744 745 746 747 748 749 750 mmm~mwm~ro~~_m~w~

2 751 752 753 754 755 756 757 758 759 760 761 762 763 764 EBm~~&m~sw~_~~

IR:

lR: 1 3 5 7 9 11 11 13 15 17 19 21 23 25 27 'Z7 29 31 31. 33 35 37 39 41 43 45 47 49 51 53 55 57 59 4

The XCOBRA-T XCOBRA-T radial that results in a [ ] during the transient transient event.

E2-68

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Table SXRB-120.2 Pulse SXRB-120.2 Power Pulse Pulse Width Pulse Height Pulse Width Pulse Height Void (seconds) (% rated)

Reference 0.53 318 318 Modified 0.56 330 Modified +0.05 0.58 328 Modified -0.05

-0.05 0.54 330 E2-69 E2-69

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION 2.00 r------------------------------------------------------------------,

1.80 .1--- - - - - - - - - - - - - - - - - - - -

1.80 -

-.-- - - Reference Reference

- -Modified

- Modified

- - - - - . Modified +0.05 --

1.60 1.60 +----------------------

-- --- - - Modified -0 .05

-0.05 1.40 -- - - - - - - - ------;

1.40

.:8 1.20 1.20 - - - - - - - - -

r:.

~

~

o 1.00 1.00 -I- - - - - - - - - - - - - -j'-J'-, <---~~~---------------------'7 '__'<

a,.

n;

~ 0_

,* 80 .1--- - - -

0.80 O.BO 0.60 i----r~---/-*--

0.40 \ -- '--r-.."--- - - - - - - - -

0.20 0.00 -I---------~--------~----------~--------~--------__r-----------J o

0 5 10 15 20 20 25 25 30 Axial Node Node Figure SRXB-120.1 SRXB-120.1 Initial Axial Power Shape

- Depleted VoidsVoids 1.6 1 . B *r-------------------------------------------------------------------~

1.4 *1- - - - - - - - - - - - - - - -

1.4

~~;:--;..:...:~

1.2 +---------=-..;~

1.2  :::::.-:---_,,-. _-

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

1 0.8 *1- - - - - - - - - - - -

1 0.8 E

8 0.6 z

O.B +--------- \

0.4 0.4 1--- - - - - - - - - - -

- - - Reference 0.2 0.2 - - Modified

- - - - - . Modified +0.05

- - - - Modified -0.05 O ~====~~-~--~--~--~-----~--~

o 0 100 200 300 400 500 600 BOO 700 800 Bundle Index Index Figure SRXB-120.2 SRXB-120.2 Initial Radial Power Distribution from Transient Transient Analyses Analyses - Depleted Voids Voids E2-70 E2-70

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 0.50

• r----------------------------------------------------------------'

O.~

0.40 OAO

~

.s:

l:i X

~

0.30

,. /, " ---. .:..:.

~ <;.

~~~.=~" ..::.'...0-.., '-' .... _* *",._ .

i:r 0.20 0.201- --

0.10 0.10

---

Reference

- - - Reference

-- - Modified Modified

...... . Modified Modified +0.05

- -. -. Modified -0.05

- . Modified 0.00 l===:::::::;=::::;::=--~_- __- _ _,_-,__----_--_-__.......J 23 23.5 23.5 24 24 24.5 24.5 25 25 Time (Sec)

Figure SRXB-120.3 SRXB-120.3 Heat Flux vs. Time Time 25% x/L - Depleted Depleted Voids Voids O . ~ r---------------------------------------------------------------_.

0.50 0.40 1- - - - - - - - __________",.c:::==~=

OAO f.s: 0.30 l:i K><

~"

0.20 - - - -

r 0.10 0.10

- - - Reference

- - Modified

..*... Modified +0.05

- . - . Modified -0.05 0.00 L--_ _ _ _ _ _ _ _ _ _ _ _ _ _.,--_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- - - - - - - - - -_ _ _ _ _ _- -_ ____l 0.00

~--~~--

23 23.5 24 24.5 24 .5 25 Time Time (Sec) (Sec)

Figure SRXB-120.4 SRXB-120.4 Heat Flux vs. Time Time 50%

50% x/L - Depleted Voids Voids E2-71

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 0.50 . , . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ,

0.40 ~-----------.-------- -

.~ .... ~ .'.' ~ ,

.r 0.30

.0 0.20 0.10 0.10

[--- .....

.-= = : = = = = ; - - -

~~

Reference

- Modified Modified Modified +0.05

- - - - - Modified Modified -005

-. - - - Modified -0.05 0.00 .L=--=.--=:::==::=~~--_-~_-~_-~

0.00 _ ___- __- _ _ __J 23 23.5 23.5 24 24.5 25 25 Time (Sec)

(Sec)

Figure SRXB-120.5 SRXB-120.5 Heat Flux vs. Time Time x/L - Depleted 75% xlL Depleted Voids Voids 50 50 r--------------------------------,

40 -------------------------------------------------------------~

30 0

0L oI 20 10 *~~=====>---------------------------

10

- - - Reference

- - Modified

- - - - - . Modified +0.05

- - - - Modified -0.05 o t=~~~~~--

0

____ ~ __ ~ __--____________--__ --~

23 23.5 23 .5 24 24.5 25 25 Time (Sec)

Figure SRXB-120.6 SRXB-120.6 Mass Flow vs. Time Time 0% x/L - Depleted Depleted Voids Voids E2-72

NON-PROPRIETARY INFORMATION NON-PROPRIETARY 50 r-------------------------------------------------------------~

50 40 ./--- - -- - - - - - - - - - - - - - - - - -

30 0

U= .--.-

I0 20 10 1;== = = = =

10

-Modified ~::~:ce Reference

~

- - - - - . Modified +0.05 ......

-....- - - Modified -0 .05

-0.05 0O L===~~~~-- ________~------____~--______~

23 23 23.5 24 24 24 .5 24.5 25 Time (Sac)

Time (Sec)

Figure SRXB-120.7 Mass Flow vs. Time Time 25% x/L - Depleted Depleted Voids Voids 50 50 r---------------------------------------------------------------,

40 - -- -

~ 30r-~~~~

30 - - - _.-- - - - - -

g.2U.

~

o

'A u::LU C-U=

In 2co

E 20 - - --

10 10 lr==== ===-,

- - - Reference

-- -Modified Modified 0 eeec

- - - - - -Modified Modified +0.05 ....

- - -- .Modified -0.05

- Modified -0 .05 23 23 23.5 23.5 24 24.5 24 .5 25 25 Time (Sec)

SRXB-120.8 Mass Flow vs. Time Figure SRXB-120.8 Time 50% x/L - Depleted Voids Voids E2-73

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 50 .-----------------------------------------------------------------~

40 40

.!!! 30 - - ..... &

g

~

0 u::U-

,T 0

<II

<II to

IE 20 1100 1r=======-,

- - - Reference

- - Modified

- - - - - - Modified +0.05

- - - - Modified -0.05 O t=~~~~~~~~~~~~~--~~~ __~~~~

23 23.5

23. 5 24 24.5 24 .5 25 25 Time (Sec)

Figure SRXB-120.9 Mass Flow Vs. Time Time 75% x/L - Depleted Depleted Voids Voids 50.------------------------------------------------------------------.

50 40 -- - -

40 F 30

~

o LL u::'aa-

<II a

<II 2to

IE 20 +--- - - - - - - - - -- - -- - -

10 t=====-~---------

10

- - - Reference

- - Modifi ed

- - - - - . Modifi ed +0.05 23

- - - - Modifi ed -0.05 O t=~~====~~~~--~~~------

23.5 24

~

24.5 24.5

~

25 25 Time (Sac)

(Sec)

Figure SRXB-120.10 SRXB-120.1 0 MassMass Flow vs. Time Time 100% x/L 100% x1L - Depleted Depleted Voids Voids E2-74 E2-74

INFORMATION NON-PROPRIETARY INFORMATION NON-PROPRIETARY r

f-'

.J Figure vs. Time SRXB-120.11 CPR VS.

Figure SRX8-120.11 Time Depleted Voids Depleted Voids E2-75 E2-75

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NRC RAI SRXB-121 It should be noted that the increase increase in the operating operating limit CPR (OLMCPR) for the increased increased voidvoid fraction cases is substantial substantial relative to the base base case. The Study 1 increaseincrease in OLMCPR OLMCPR is 0.014 andand the Study 2 increase in OLMCPR is 0.027.

In response response to RAI SRXB-88, TVA stated that COTRANSA2 COTRANSA2 includesincludes aa 110 percent multiplier 110 percent multiplier on integral integral thermal power power as a conservative assumption.

assumption. However, XN-NF-80-19(P)(A) Exxon XN-NF-80-19(P)(A) Exxon Nuclear Nuclear Methodology Methodology for Boiling WaterWater Reactors, THERMEX:

THERMEX: Thermal Limits Methodology Methodology Summary Description SectionSection 4.4 states: "In "In developing developing the methodology methodology for the COTRANSA COTRANSA code Exxon Nuclear addressed uncertainties in the code through addressed uncertainties through the integral power variable.

methodology uses aa more conservative The revised methodology conservative deterministic deterministic bounding value (+10 percent) for the integral power uncertainty."

uncertainty." TVA's evaluation of the 110 110 percent conservatism found that the OLMCPR OLMCPR margin afforded by the conservatism conservatism is (( )).

)).

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

It should be noted that the intent of the 110 110 percent multiplier is to conservatively conservatively bound all uncertainties, including uncertainties uncertainties, including uncertainties in other important variables variables such as flow and frictionfriction factors.

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

Response to SRXB-121 As discussed in the response to RAI SRXB-88 (Reference (Reference SRXB-121.1), the [ ]

correlation in MICROBURN-B2 MICROBURN-B2 was modified to adjust the mean to match the measured measured ATRIUM-10 void fraction data data for both high and low void fractions. The modified modified (( ]I correlation was then further modified modified to generate two bounding correlations correlations for the ATRIUM-10 ATRIUM-10 data of +/-0.05 meanmean void. The results of the modifiedmodified correlations correlations were shown in Figure SRXB-88.2 SRXB-88.2 and were used in sensitivity Study Study 1 and Study 2.

The sensitivity studies described described in the response response to RAI SRXB-88 are somewhat somewhat artificial and only capture capture the sensitivity sensitivity of portions of the methodology methodology to void correlation correlation uncertainty.

Study 2 is especially especially artificial artificial in that the results of the study only reflect the increaseincrease in core void void reactivity coefficient. Study 2 does not reflect that a different different change in void fraction would would occur for a given pressure change change with the modified modified void correlation. Study 1 included this this effect and resulted in a smaller effect effect on the OLMCPR. Other Other effects of using a different different void correlation uncertainty are not incorporated incorporated into Study 1 (e.g., pressure drop correlationcorrelation coefficients would be different). These sensitivity studies are not complete coefficients complete assessments of the the impact of void correlation correlation uncertainty uncertainty on OLMCPR.

OLMCPR.

E2-76

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION It should be noted that a +/-0.05 perturbation perturbation of the void correlation used in the SRXB-88 SRXB-88 sensitivity studies is substantial. For example, the +0.05 void scenario is equivalent equivalent to a

[ ].

]. The The measure measure of void correlation uncertainty uncertainty used used in the sensitivity analyses was somewhat somewhat arbitrarily arbitrarily defined as a value that would bound the ATRIUM-10 ATRIUM-10 test data. In a BWR, the core power and and power distribution are tightly coupled with the void fraction and a large errorerror in predicted core void fraction would have a significant significant effect on the predicted power distribution distribution measurements measurements operating reactors. If the error in void fraction was as large as assumed in the obtained from operating the SRXB-88 sensitivity studies, the effect would be observed observed in comparisons comparisons of predicted predicted to measured power measured distributions obtained power distributions obtained from operating operating reactors.

Additional calculations were performed ((

calculations were

]. These results confirm the conclusion conclusion stated above that the the increased increased void variation of +0.05 is not realistic.

Integral power is a parameter Integral parameter obtainable from test measurements measurements that is directly related to ACPR

~CPR and provides a means means to assess code uncertainty. The COTRANSA COTRANSA transient analysis analysis methodology methodology was a predecessor predecessor to the COTRANSA2 COTRANSA2 methodology. The integral integral power figure of merit merit was introduced introduced with the COTRANSA COTRANSA methodology as a way to assess (not account for) uncertainty impact code uncertainty impact on ACPR.

~CPR. From COTRANSA COTRANSA analyses analyses of the Peach Bottom turbine trip trip tests, the mean mean of the predicted predicted to measured measured integral integral power power was 99.7% with a standard E2-77

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION deviation of 8.1 deviation 8.1%.

%. AREVA AREVA (Exxon Nuclear Nuclear at the time) initially proposed to treat integral power statistical parameter. However, following discussions as a statistical discussions with the NRC, NRC, it was agreed agreed to apply apply a deterministic 110% integral integral power power multiplier multiplier (penalty) on COTRANSA calculations for COTRANSA calculations licensing analyses. That increase was sufficient licensing sufficient to make the COTRANSA COTRANSA predictedpredicted to measured integral measured integral power conservative conservative for all of the Peach Bottom turbine trip tests.

COTRANSA2 (Reference COTRANSA2 (Reference SRXB-121.2)

SRXB-121.2) was developed and approved as a replacement replacement for COTRANSA in the AREVA thermal limits methodology COTRANSA methodology (Reference (Reference SRXB-121.3). Initially itit was was not planned planned to use the 110% integral integral power multiplier with the COTRANSA2 methodology.

COTRANSA2 methodology.

COTRANSA2 COTRANSA2 predictions of integral power were conservative conservative for all Peach Peach Bottom turbine trip trip tests. The minimum conservatism was [

minimum conservatism )) and the meanmean of the predicted to measured integral power was [

integral ].]. The comparisons comparisons to the Peach Peach Bottom Bottom turbine trip tests demonstrated that the 110%

demonstrated 110% integral power power multiplier multiplier was not needed needed for COTRANSA2.

However, because the thermal limits methodology methodology that was approved independently independently of COTRANSA2 included discussion COTRANSA2 discussion of the 110% 110% integral integral power power multiplier, the use of the multiplier multiplier was retained for COTRANSA2 COTRANSA2 licensing calculations. With the 110%

110% multiplier, the the COTRANSA2 predicted to measured COTRANSA2 measured mean integral integral power power is [ ] for the the Peach Bottom turbine trip tests. Applying aa (( ] integral integral power power multiplier multiplier provides an OLMCPR conservatism OLMCPR conservatism of (( ] versus the (( ] reported in the response to RAI SRXB-88 for the 110% 110% multiplier alone.

summarize the above paragraphs, the sensitivity studies described To summarize described in the response to RAI SRXB-88 SRXB-88 overestimate overestimate the potential impact impact of uncertainty uncertainty in the void correlation correlation and thethe 110% integral integral power multiplier is just one part of the conservatism conservatism in the COTRANSA2 COTRANSA2 methodology and application process methodology process that covers methodology uncertainties.

methodology uncertainties.

COTRANSA2 is not a statistical COTRANSA2 statistical methodology methodology and uncertainties are not directly input to the the analyses. The methodology methodology is a deterministic deterministic bounding approach that contains sufficient sufficient conservatism to offset uncertainties in individual individual phenomena.

phenomena. Conservatism is incorporated in in the methodology in two ways: (1) computer code models are developed to produce produce conservative conservative results on an integral integral basis relative relative to benchmark benchmark tests, and (2) importantimportant input parameters parameters are biased biased in a conservative direction conservative direction licensingin licensing calculations. Justification Justification that the integrated integrated effect of all the conservatisms conservatisms in COTRANSA2 COTRANSA2 licensing analyses adequate for EPU analyses is adequate operation operation is provided provided below.

• The COTRANSA2 methodology results in predicted power increases that are bounding COTRANSA2 methodology bounding

([ ] on average) relative to Peach Bottom benchmark benchmark tests. In addition, for licensing licensing calculations calculations a 110% 110% multiplier is applied to the calculated integral power to provide conservatism. This approach provide additional conservatism. approach adds significant conservatism to the the calculated OLMCPR as discussed previously.

calculated OLMCPR

• important input parameters in licensing calculations Biasing of important calculations provides provides additional conservatism conservatism in establishing the OLMCPR. The Peach Peach Bottom turbine trips were performed assuming assuming the measuredmeasured performance performance of important input parameters parameters such as as control control rod scram speed and turbine valve closing times. For licensing licensing calculations, these (and other) parameters are biased in aa conservative conservative bounding direction. These These conservative conservative assumptions assumptions are not combined statistically; assuming all parameters are bounding at the same time produces produces very conservative conservative results.

• Assessments such as the Peach Bottom tests indicate that the integrated Assessments integrated effect of all the the conservatism COTRANSA2 is adequate conservatism in COTRANSA2 adequate for non-EPU non-EPU reactor conditions (as stated in in the RAI). To demonstrate demonstrate that the impact of the change in void-quality void-quality correlations correlations is E2-78 E2-78

NON-PROPRIETARY NON-PROPRIETARY INFORMATIONINFORMATION similar for EPU and non-EPU non-EPU conditions, the RAI SRXB-88 SRXB-88 sensitivity sensitivity analyses analyses (Study 1) repeated for BFN Unit 3 Cycle 14 without EPU. The change were repeated change in ACPR

~CPR relative relative to the reference reference cases for EPU and non-EPU non-EPU are shown in the table below:

~(~CPR)

A(ACPR)

Case EPU Non-EPU

+O.OS void

+0.05 +0.016 +0.024

-O.OS void

-0.05 -0.001

-0.001 -0.007

-0.007 Based Based on these results for EPU and non-EPU conditions,conditions, it is concluded concluded that EPU conditions do not increase the sensitivity to a change conditions change in the void correlation.

correlation.

• As discussed previously, the core axial power distribution is tightly coupled with the void void fraction. A large error in predicted predicted void fraction would have a significant significant effect effect on the the predicted predicted axial power distribution distribution measurements measurements obtained from operating reactors. The The very good comparisons between between predicted and measured measured axial power distributions distributions obtained from operating obtained operating reactors indicates indicates that the void distribution distribution within the core is being being predicted predicted well.

well.

• Minimal Minimal plant transient data at EPU conditions conditions is available to benchmark benchmark transient analysis methodologies.

methodologies. However, at the request of the NRC, a COTRANSA2 COTRANSA2 analysisanalysis was performed performed for a recent event that occurred occurred at a BWR/4 approved for EPU operation. operation.

The event involved a reduction reduction in pump speed in one of the recirculation loops loops followed by a sudden increase in the pump speed approximately approximately 40 seconds later. The event event did did not pose a challenge challenge to the fuel; however, the event did result in aa significant change in in core void fraction. Because of the tight coupling coupling between between core void fraction and core power, aa comparison of the predicted predicted to measured core power response during the the event is a good good way to assess the accuracy accuracy of the void correlation.

correlation. For this analysis, a best estimate approach approach was used and event event specific specific licensing conservatisms were not licensing conservatisms not applied (e.g., measured measured data used as boundary boundary conditions, realistic control system parameters, best estimate estimate core neutronics neutronics data). The recirculation recirculation pump speed versusversus time from the plant data was used as a boundary boundary condition for the analysis analysis (Figure SRXB-121.13). COTRANSA2 analysis SRXB-121.13). The COTRANSA2 analysis predicted predicted the core power and reactor pressure response very well (Figures SRXB-121.14 SRXB-121.14 and SRXB-121.1S).

SRXB-121.15). The The very good agreement agreement for the predicted predicted core power reached reached following the pump runback and the following pump runup indicates indicates a good prediction of the core void fraction during during the event.

Based on the above above discussions, the impact of void correlation uncertainty is inherently correlation uncertainty inherently incorporated in the analytical methods used to determine incorporated determine the OLMCPR. No additional adjustments adjustments to the OLMCPR OLMCPR are required to address address void correlation uncertainty.

correlation uncertainty.

References:

SRXB-121.1 June 3, 2008, TVA Letter to NRC,NRC, Browns Nuclear Plant (BFN)

Browns Ferry Nuclear (BFN) - Units 2 And 3 - Technical Specifications (TS) Change TT-418 - Extended Power Uprate Technical Specifications Uprate (EPU) Supplemental Response (EPU) - Supplemental Response To Round 16 16 Request Request For Additional Information (RAI) - SRXB-88 (TAC Nos. MDS263 Information (RAI) MD5263 AND MDS264)MD5264) (MI081640325).

(MI08164032S).

E2-79

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION SRXB-121.2 SRXB-121.2 ANF-913(P)(A) Volume 1 Revision 1 and Volume Volume 1 Supplements Supplements 2,2,3 3 and 4, COTRANSA2: A COTRANSA2: Computer Program A Computer Programfor Boiling Boiling Water Reactor Transient Water Reactor Transient Analyses, Advanced Nuclear Analyses, Nuclear Fuels Corporation, August 1990.

SRXB-121.3 SRXB-121.3 XN-NF-80-19(P)(A) Volume XN-NF-80-19(P)(A) Volume 3 Revision Revision 2, Exxon Nuclear Nuclear Methodology for for Boiling Water Reactors Water Reactors THERMEX: Thermal Thermal Limits Methodology Summary Description, Description, Exxon Nuclear Company, January 1987.

Exxon Nuclear E2-80 E2-80

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION rr-Figure SRXB-121.1 SRXB-121.1 BFN 2D 20 TIP Statistic Statistic Comparison for Variations of the Void Quality Correlation Correlation rr" Figure SRXB-121.2 SRXB-121.2 BFN 3D 30 TIP Statistic Comparison for Variations of the Void Quality Correlation Correlation E2-81

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION r

Figure SRXB-121.3 SRXB-121.3 BFN Core Average Axial TIP Comparison at MWd/MTU for Variations 9026 MWd/MTU Variations of the Void Quality Quality Correlation r

Figure Figure SRXB-121.4 SRXB-121.4 BFN Core Core Average Average Axial TIP Comparison Comparison atat 1755 MWd/MTU 1755 MWd/MTU for Variations Variations of the Void Quality Correlation Void Quality Correlation E2-82 E2-82

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr-U-

.J Figure SRXB-121.5 SRXB-121.5 BFN Core Average Axial TIP Comparison Comparison atat 9197 MWd/MTU MWd/MTU for Variations Variations of the Void Quality Correlation rr U-

.J SRXB-121.6 BFN Core Average Figure SRXB-121.6 Average Axial TIP Comparison Comparison atat 1340 MWd/MTU 1340 MWd/MTU for Variations of the Void Quality Correlation Correlation E2-83 E2-83

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION r

.J Figure SRXB-121. BWR/4 at EPU 2D SRXB-121. 7 A BWRl4 20 TIP Statistic Comparison Comparison for Variations of the Void Quality Variations Correlation Quality Correlation rr

.J SRXB-121.8 A BWRl4 Figure SRXB-121.8 BWR/4 at EPU 303D TIP Statistic Comparison Comparison for for Variations Variations of the Void Quality Correlation E2-84

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION rr.

-J

..J SRXB-121.9 A BWRl4 Figure SRXB-121.9 BWR/4 at EPU Core Average Axial TIP Comparison at 2127 MWd/MTU MWd/MTU for Variations of the Void Quality r

r

..J Figure SRXB-121.10 SRXB-121.10 A BWR/4 BWRl4 at EPU EPU Core Average Average Axial TIP Comparison Comparison at at 10621 MWd/MTU for Variations 10621 MWd/MTU Variations of the Void Void Quality Correlation Quality Correlation E2-85 E2-85

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION r

SRXB-121.11 A BWR/4 at EPU Core Average Figure SRXB-121.11 Average Axial TIP Comparison Comparison at 18459 MWd/MTU 18459 MWd/MTU for Variations of the Void Quality Correlation Correlation r

(.--

..J Figure SRXB-121.12 SRXB-121.12 A BWR/4 at EPU Core Average Axial TIP Comparison at at 2054 MWd/MTU MWd/MTU for Variations of the Void Quality Quality Correlation E2-86 E2-86

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 1.20 , -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ,

1.00 Il- - - - - - - - - - - - - - - - - - - - - - - - -r:rr--.;--:-----i'- - - - - - j

- - - - - - - Measured -------

-4a-

--<>-Analysis Analysis Input 0 . 80 +-~----------------------f------------1 0.80

~Q.

~ 0.60 +-- - --\-- - - - - - - - - - - - - - --+- - - - - - - - - - - - - - 1 A.

E 0.

l I.

D.

0.40 4- - - - - - - - - ' ~------------_/

0.20 +--- - - - - - - - - - - - - -- - - - - - -- - - - - - - - - - - -1 0.00 - ! - - - - - - . , - - - - - - , - - - - - - - - - - , - - - - - - , - - - - - - - - - , - - - - - - - - - j 0.0 20.0 20 .0 40.0 60.0 80.0 100.0 120.0 120.0 Time Time (sac)

(sec)

Figure SRXB-121.13 Pump Speed Figure SRXB-121.13 120 r - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

4 M

- -+ - - Measured easured

- - Calc ulated 100 -*~-------------------_r1 80 S-a.

II.

LU W

~

~S 60 *

~

~

0 II.

40 20 -~-------------------------------_i 00 + - ------.,------,----------,------,---------r--------1 0.00 20.00 20.00 40.00 60.00 80.00 100.00 120.00 Time (sec)

(sec)

Figure SRXB-121.14 SRXB-121.14 Core Power Power E2-87

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION 1100 1100 ,-r-------------------------------------------------------------------.

1050 i-----------------------------------------------------------------~

Ci +

'iii

.S:

~ 1000 +---------~-------------------------------~----------------------~

I/)

I/)

e C1. 9.

950 950 +-----------~

- - -+ --.

...- - Reactor Pressure, Measured Measured x Dome Pressure , Measured Dome Pressure, Measured

--Dome Dome Pressure, Pressure , Calculated 900 +---------~----------~----------r_--------_r----------~--------~

o0 20 40 60 60 80 100 120 120 (sac)

Time (sec)

Figure SRXB-1 21.15 Reactor SRXB-121.15 Reactor Pressure E2-88 E2-88

NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION NRC RAI SRXB-122 SRXB-122 The modified correlations correlations are based on constant constant slip models. Provide Provide a discussion regarding regarding the treatment of subcooled boiling.

boiling. This discussion discussion should address address void fraction continuity at the boiling boundary. Describe Describe any impact on the transient transient analyses analyses arising from SCRAM reactivity worth ifif significant differences differences are expected expected based on treatment of subcooled subcooled boiling.

SRXB-122 Response to SRXB-122 The thermal thermal hydraulic methodology incorporates the effects of subcooled boiling through use of methodology incorporates the Levy model. The Levy model predicts a critical subcooling subcooling that defines defines the onset of boiling.

The critical critical subcooling is used with a profile fit model to determine determine the total flow quality quality that accounts accounts for the presence presence of subcooled subcooled boiling. The total flow quality is used used with the the void-quality void-quality correlation determine the void fraction. This void fraction explicitly includes the correlation to determine the effects effects of subcooled subcooled boiling. Application of the Levy model results in a continuous continuous void fraction fraction distribution at the boiling boundary.

distribution The major influence that the void-quality models have on scram reactivity worth is through the the power shape. As discussed in previous predicted axial power previous responses (e.g., SRXB-121), the the void-quality void-quality models used for ATRIUM-10 ATRIUM-10 fuel result in a very good prediction of the axial power power shape.

Below are reponses to the five fuels related RAIs, SRXB-123 through SRXB-127, from NRC's Below are reponses to the five fuels related RAls, SRXB-123 through SRXB-127, from NRC's September 16, 2008, 2008, Round 20 RAI.

Introduction to Round 20 RAI NRC Introduction The following RAlsRAIs are based on proprietary proprietary draft responses provided during a public meeting responses provided held with the TVA regarding the BFN BFN Units 2 and 3 EPU review on August 7,2008. 7, 2008. These These questions focus on the proposed proposed response to SRXB 106.

The draft response states that the calculation terminates in the calculated calculation terminates calculated pressure pressure exceeds the the correlation bounds bounds ((( fl).

))). However, under anticipated transient without scram (ATWS) under anticipated expected to exceed this value (([

conditions the pressure is expected 1] pounds per square inch

))

gage (psig)].

SRXB-123 NRC RAI SRXB-123 Discuss what allows the code to continue its evaluationevaluation of the ATWS transient without without terminating.

Response to SRXB-123 SRXB-123 The response to SRXB-1 06 is relative to the XCOBRA-T XCOBRA-T computer code. The XCOBRA-T computer code is not used in the ATWS overpressurization overpressurization analysis. The COTRANSA2 COTRANSA2 computer code is the primary code used for the ATWS overpressurization overpressurization analysis. The ATWS ATWS overpressurization event is not used overpressurization used to establish operating operating limits for critical power; therefore, thethe SPCB critical power power correlation correlation pressure limit is not a factor in the analysis.

E2-89

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION SRXB-124 NRC RAI SRXB-124 Discuss how the core coolability under 10 10 CFR 50.46 is evaluated for this event.

Response to SRXB-124 Response SRXB-124 The ATWS A TWS event is not limiting relative to acceptance acceptance criteria identified identified in 10 CFR 50.46. The The core remains remains covered covered and adequately cooled cooled during the event. Following Following the initial power increase during increase pressurization phase, the core returns to natural during the pressurization natural circulation conditions conditions after the recirculation recirculation pumps trip and fuel cladding temperatures temperatures are maintained at acceptably acceptably lowlow levels. The ATWS event is significantly less limiting than the loss-of-coolant loss-of-coolant accident relative to 10 10 CFR 50.46 acceptance acceptance criteria.

SRXB-125 NRC RAI SRXB-12S Assuming that the pressure pressure is out of bounds, address address how does the code conservatively conservatively predicts the fuel temperature.

temperature.

Response to SRXB-12S Response SRXB-125 As indicated in the response to SRXB-1 SRXS-123, 23, the pressure limit is for application application of the SPCS SPCB critical power correlation. The SPCS power correlation. SPCB correlation correlation is not used in the ATWS overpressurization overpressurization analysis.

Dryout conditions conditions are not expected expected to occur for the core average channel that is modeled average channel modeled in in COTRANSA2. Dryout might occur in the limiting (high power) channels of the core during the COTRANSA2. the ATWS event; however, these channels are not modeled in COTRANSA2 COTRANSA2 analyses. For the the overpressurization analysis, ATWS overpressurization analysis, ignoring dryout for the hot channels conservative in that it channels is conservative maximizes the heat transferred maximizes transferred to the coolant and results in a higher calculated pressure.

higher calculated SRXB-126 NRC RAI SRXB-126 If If a fuel rod is predicted predicted in dryout, address address how the heat transfer is modeled.

modeled.

Response to SRXB-126 SRXB-126 Dryout conditions are not expected expeCted to occur for the core average average channel channel that is modeled in in COTRANSA2 for the ATWS overpressurization COTRANSA2 overpressurization analysis. Dryout might occur in the limiting limiting channels of the core during the ATWS (high power) channels ATWS event; event; however, these channels are not modeled in COTRANSA2 COTRANSA2 analyses. For the ATWS overpressurization overpressurization analysis, ignoring dryout ignoring dryout for the hot channels is conservative conservative in that itit maximizes maximizes the heat heat transferred transferred to the coolant and results in a higher calculated pressure.

higher calculated E2-90 E2-90

NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NRC RAI SRXB-127 SRXB-127 Discuss whether the heat transfer modeling approachapproach is conservative conservative in terms of the figure of merit (vessel pressure).

Response to SRXB-127 SRXB-127 conditions are not expected Dryout conditions expected to occur for the core average channel that is modeled average channel modeled inin COTRANSA2. For the ATWS overpressurization overpressurization analysis, ignoring dryout for the hot channels channels is conservative in that itit maximizes maximizes the heat transferred transferred to the coolant coolant and results in aa higher higher calculated pressure.

For BWRs, the fluid heat transfer coefficients are high and the thermal resistance of the fluid film transfer coefficients film is much smaller smaller than the thermal resistance of the cladding, cladding, the cladding-to-pellet cladding-to-pellet gap, and the the fuel pellet. Variations in the calculated calculated heat transfer coefficients will have insignificant effect have an insignificant on the calculated calculated peak vessel pressure.

E2-91

ENCLOSURE ENCLOSURE 3 TENNESSEE TENNESSEE VALLEY VALLEY AUTHORITY AUTHORITY BROWNS BROWNS FERRY FERRY NUCLEAR NUCLEAR PLANT PLANT (BFN)

(BFN)

UNITS 2 AND 3 UNITS2AND3 TECHNICAL TECHNICAL SPECIFICATIONS SPECIFICATIONS (TS) CHANGE CHANGE TS-418 TS-418 EXTENDED POWER UPRATE EXTENDED POWER UPRATE (EPU) (EPU)

SUPPLEMENTAL SUPPLEMENTAL RESPONSE RESPONSE TO TO REQUEST REQUEST FOR FOR ADDITIONAL ADDITIONAL INFORMATION INFORMATION (RAI)

(RAI) ROUNDS ROUNDS 33 AND 18 18 AND AND RESPONSE RESPONSE TO TO ROUND ROUND 20 20 FUELS FUELS METHODS METHODS RAIs RAls AREVA AREVA AFFIDAVIT AFFIDAVIT This This enclosure enclosure provides provides AREVA's AREVA's affidavit affidavit for for Enclosure Enclosure 1.

1.

AFFIDAVIT AFFIDAVIT COMMONWEALTH COMMONWEALTH OF VIRGINIA VIRGINIA )

) ss.

CITY OFOF LYNCHBURG LYNCHBURG )

1. My name is Gayle F. Licensing, for AREVA F. Elliott. I am Manager, Product Licensing, NP Inc. (AREVA NP) and as such II am authorized authorized to execute execute this Affidavit.

Affidavit.

2. am familiar with the criteria applied by AREVA NP to determine I am determine whether whether certain information is proprietary. I am familiar with the policies established certain AREVA NP information established by AREVA AREVA NP to ensure ensure the proper proper application application of these criteria.

3. I am familiar with the AREVA information contained AREVA NP information Responses to contained in the Responses NRC RAI for Round Round 18 and Round 20 for Browns Ferry EPU, dated September September 2008 and referred to herein as "Document." Information referred Information contained in this Document has been classified by AREVA AREVA NP as proprietary proprietary in accordance accordance with the policies established by AREVA AREVA NP for the the control and protection of proprietary proprietary and confidential information.

information.

4. This Document Document contains information of a proprietary contains information proprietary and confidential confidential nature nature and is of the type customarily held in confidence by AREVA available to the AREVA NP and not made available the public. Based on my experience, II am aware that other companies companies regard information of the the kind contained in this Document as proprietary and confidential.

confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory contained in this Document be Commission in confidence with the request that the information contained withheld from public disclosure. The request for withholding of of proprietary proprietary information is made in in information for which withholding from disclosure is accordance with 10 CFR 2.390. The information

requested qualifies qualifies under 10 CFR 2.390(a)(4) "Trade secrets under 10 secrets and commercial or financial information.""

information.

6. The following criteria are customarily customarily applied by AREVA NP to determine determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development development plans and programs programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly significantly reduce its expenditures, expenditures, in time or resources, to design, produce, produce, or market a similar product or service.

(c) The information includes test data or analytical analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

competitive NP.

(d) The information information reveals reveals certain distinguishing aspects of a process, distinguishing aspects methodology, or component, the exclusive exclusive use of which provides a competitive competitive advantage for AREVA AREVA NP in product optimization or marketability.

marketability..

(e) The information is vital to a competitive competitive advantage advantage held by AREVA NP, would be helpful to competitors competitors to AREVA AREVA NP, and wouldwould likely cause substantial harm to the competitive competitive position of AREVA AREVA NP.

The information in the Document Document is considered proprietary for the reasons set forth in considered proprietary paragraphs 6(b) and 6(c) above.

paragraphs

7. In accordance accordance with AREVA AREVA NP's policies policies governing the protection and control of information, of information, proprietary proprietary information information contained in this Document Document have been made available, on a on a limited limited basis, basis, to to others others outside AREVA NP only as required required and under under suitable agreement agreement providing providing for nondisclosure nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary proprietary information be kept in a secured file or area area and distributed distributed on a need-to-know need-to-know basis.

9. The foregoing foregoing statements are true and correct correct to the best of my knowledge, knowledge, information, and belief.

SUBSCRIBED before me this ----'1"Th*'

-It'_"tb__

day of September September 2008.

Sherry L. L. McFaden McFaden NOTARY PUBLIC, COMMONWEALTH NOTARY COMMONWEALTH OF VIRGINIA VIRGINIA COMMISSION EXPIRES: 10/31/10 MY COMMISSION 10/31/10 Reg. # 7079129 SHERRY L. MCFADEN

---~

NotIry Commonwealth of Commonwealth My Commission7079129 Public Noto"ry Public of V1rg.tnla Vlrginla Expires Oct 31, 20101 7079129 My Commission Expire. Oct 31. 2010 .

I I