Text

Areva Report ANP-2638NP, Revision 0, Applicability of Areva Np BWR Methods to Extended Power Uprate Conditions, Enclosure 6
	ML072180374
Person / Time
Site:	Brunswick
Issue date:	07/31/2007
From:	; AREVA NP
To:	; Office of Nuclear Reactor Regulation
References
	BSEP 07-0075, TAC MD4061, TAC MD4062, TSC-2006-06 ANP-2638NP, Rev 0
	Download: ML072180374 (79)
	v • d • e

BSEP 07-0075 AREVA Report ANP-2638(NP), Revision 0, Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions dated July 2007

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Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page i Nature of Changes Item Page Description and Justification

1.

All This is a new document.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page ii Contents 1.0 Introduction....................................................................................................................

1-1 2.0 Overview........................................................................................................................

2-1 3.0 Thermal-hydraulics.........................................................................................................

3-1 4.0 SPCB CHF Correlation..................................................................................................

4-1 5.0 Safety Limit MCPR.........................................................................................................

5-1 6.0 Mechanical Limits Methodology......................................................................................

6-1 7.0 Core Neutronics.............................................................................................................

7-1 8.0 Transient Analysis..........................................................................................................

8-1 9.0 LOCA Analysis...............................................................................................................

9-1 10.0 Stability Analysis..........................................................................................................

10-1 11.0 Summary.........................................

11-1 12.0 References...................................................................................................................

12-1 Appendix A Application of AREVA Methodology for Mixed Cores.................................

A-1 A.1 Discussion.....................................................................................................................

A-1 A.2 References....................................................................................................................

A-5 Appendix B Thermal Hydraulic Methods........................................................................

B-1 B.1 Hydraulic Characterization........................................................................................

B-1 B.2 References....................................................................................................................

B-8 Appendix C Neutronic Methods.....................................................................................

C-1 C.A Cross Section Representation.................................................................................

C-1 C.2 Applicability of Uncertainties.........................................................................................

C-4 C.3 Fuel Cycle Comparisons...............................................................................................

C-7 C.3.1 Bypass Voiding.............................................................................................................

C-9 C.3.2 Fuel Assembly Design............................................................................................

C-i I AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page iii C.4 References..................................................................................................................

C-12 Appendix D Stability Methods.............................................................................................

D-1 D.1 DIVOM Analysis............................................................................................................

D-1 D.2 ATW S/Instability............................................................................................................

D-1 D.2.1 Background...................................................................................................................

D-2 D.2.2 Review of the Im portant Phenomena............................................................................

D-3 D.2.2.1 -Global Oscillation Lim it Cycle Am plitude.....................................................................

D-3 D.2.2.2-Average Power Increase due to Oscillations...............................................................

D-4 D.2.2.3-Lim iting Mode...............................................................................................................

D-5 D.2.2.4-The Role of the Prompt-Critical Pulses........................................................................

D-5 D.2.2.5-The Role of Void-Reactivity Coeffi cient.......................................................................

D-6 D.2.3 Assessing the Change in Fuel Dam age Potential.........................................................

D-6 D.2.3.1 -Effect of the ATRIUM-10 Fuel Type.............................................................................

D-7 D.2.3.2-Effect of EPU Core Loading.........................................................................................

D-7 D.3 References....................................................................................................................

D-8 AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page iv Tables Table 7-1 CASMO-4/MICROBURN-B2 Operating Experience...........................................

7-2 Table 7-2 CASMO-4/MICROBURN-B2 Other Planned Uprate Support............................. 7-3 Table B-1 AREVA Multi-Rod Void Fraction Validation Database.......................................

B-5 Table C-1 KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A).................. C-13 Table C-2 Comparison of CASMO-4 and MCNP results for ATRIUM-10 D esign.............................................................................................................

C -13 Table C-3 Fuel Enrichment Description for the Initial Brunswick EPU ATRIUM -10 Fuel Cycle Design......................................................................

C-14 AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page v Figures Figure 3-1 Comparison of Karistein Two-Phase Pressure Drop Test Matrix and Typical Reactor Conditions..........................................................................

3-2 Figure 3-2 Comparison of Karlstein Two-Phase Pressure Drop Test Matrix and Brunswick Reactor Conditions with ATRIUM-1 0 Fuel.................................

3-2 Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis..................................................................................................

5-3 Figure 8-1 Typical Hydraulic Benchmarks to Karlstein Transient Simulations..................... 8-2 Figure B-1 Comparison of the Measured Local Quality for ATRIUM-10 Void Data and Exit Quality for Typical Reactor Conditions....................................

B-6 Figure B-2 Validation of [

] using FRIGG-2 and FRIGG-3 Void D ata...................................................................................................................

B -6 Figure B-3 Validation of [

] using ATRIUM-10 Void Data.................................

B-7 Figure B-4 Validation of Ohkawa-Lahey using FRIGG-2 and FRIGG-3 Void D ata...................................................................................................................

B -7 Figure B-5 Validation of Ohkawa-Lahey using ATRIUM-10 Void Data...........................

B-8 Figure C-1 Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches..............................................................................................

C-15 Figure C-2 Microscopic Fast Cross Section of U-235 from Base Depletion and Branches.....................................................................................................

C-15 Figure C-3 Microscopic Thermal Cross Section of U-235 at Beginning of Life................. C-16 Figure C-4 Microscopic Fast Cross Section of U-235 at Beginning of Life........................

C-16 Figure C-5 Microscopic Thermal Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions.................. C-17 Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions.................. C-17 Figure C-7 Macroscopic Diffusion Coefficient (Fast and Thermal) Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions..........................................................................................................

C -18 Figure C-8 Microscopic Thermal Cross Section of U-235 at 70 GWd/MTU...................... C-19 Figure C-9 Quadratic Interpolation Illustration of Microscopic Thermal Cross Section of U-235..............................................................................................

C-20 Figure C-10 Illustration of Final Quadratic Interpolation for Microscopic Thermal Cross Section of U-235.....................................................................

C-20 Figure C-I 1 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9............................................................................

C-21 AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page vi Figure C-12 Figure C-13 Figure C-14 Figure C-15 Figure C-16 Figure C-17 Figure C-18 Figure C-19 Figure C-20 Figure C-21 Figure C-22 Figure C-23 Figure C-24 Figure C-25 Figure C-26 Figure C-27 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction...............................................

C-21 Delta k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at 0.4 Historical Void Fraction and 0.9 Instantaneous Void Fraction............................................................................

C-22 Comparison of Interpolation Process Using Void Fractions of 0.0, 0.4 and 0.8 and Void Fractions of 0.0, 0.45 and 0.9........................................

C-22 EMF-2158(P)(A) TIP Statistics by Axial Level.................................................

C-23 Quad Cities Unit 1 Pin by Pin Gamma Scan Results......................................

C-23 Maximum Assembly Power in Topical Report EMF-2158(P)(A)...................... C-24 Maximum Exit Void Fraction in Topical Report EMF-2158(P)(A).................... C-24 Maximum Assembly Power Observed from Recent Operating Experience......................................................................................................

C-25 Void Fractions Observed from Recent Operating Experience......................... C-25 Axial Power and Void Profile Observed from Recent Design Experience......................................................................................................

C -26 Nodal Void Fraction Histogram Observed from Recent Design Experience......................................................................................................

C-26 Maximum Assembly Power in an EPU Brunswick Design..............................

C-27 Maximum Exit Void Fraction in an EPU Brunswick Design.............................

C-27 Brunswick EPU Design Axial Profile of Power and Void Fraction................... C-28 Brunswick EPU Design Nodal Void Fraction Histogram..................................

C-28 Hypothetical MICROBURN-B2 Multi-Channel Average Bypass V oid Distribution..............................................................................................

C-29 This document contains a total of 78 pages AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page vii Nomenclature Acronym BWR CHF EPU LHGR LOCA LRNB MAPLHGR MCPR NRC OLMCPR RAI SER SLMCPR Definition Boiling Water Reactor Critical Heat Flux Extended Power Uprate Linear Heat Generation Rate Loss of Coolant Accident Load Reject with no Bypass Maximum Average Planar Linear Heat Generation Rate Minimum Critical Power Ratio Nuclear Regulatory Commission Operating Limit Minimum Critical Power Ratio Request for Additional Information Safety Evaluation Report Safety Limit Minimum Critical Power Ratio AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page 1-1 1.0 Introduction This document reviews the AREVA NP licensing methodologies to demonstrate that they are applicable to operation of the Brunswick Nuclear Power Generating Stations at Extended Power Uprate (EPU) conditions. EPU conditions refer to power uprate to 120% of the originally licensed rated thermal power.

AREVA NP Inc. is an AREVA and Siemens company.

AREVA NP Inc-

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 2-1 2.0 Overview The first step in determining the applicability of current licensing methods to EPU conditions was a review of current AREVA BWR topical reports to identify SER restrictions on the BWR methodology. This review identified that there are no SER restrictions on power level for the AREVA topical reports. The review also indicated that there are no SER restrictions on the parameters most impacted by the increased power level: steam flow, feedwater flow, jet pump M-ratio, and core average void fraction.

The second step consisted of an evaluation of the core and reactor conditions experienced under EPU conditions to determine any challenges to the validity of the models. When the reactor power is increased, the resultant impact on operating margin is mitigated to a large extent by a decrease in the limiting assembly radial power factor. This decrease in the limiting assembly radial power factor is necessary since the thermal operating limits (MCPR, MAPLHGR and LHGR) that restrict assembly power are dependent on the limiting assembly power but are fairly insensitive to the core thermal power. From this fundamental constraint the following observations may be made about the EPU operating conditions:

1. The reduction in the hot assembly radial peaking factor leads to a more uniform radial power distribution and consequently a more uniform core flow distribution. The net result being less flow starvation of the hottest assemblies.

2. With the flatter radial power distribution, more assemblies and fuel rods are near thermal limits.

3. From a system perspective, there will be higher steam flow and feedwater flow rates.

4. With an increase in the average assembly power in the reactor, the core pressure drop will increase slightly resulting in a decrease in the jet pump M-ratio for a given core flow rate.

5. Core average void fraction will increase.

Based on these fundamental characteristics of power uprate, each of the major analysis domains (thermal-hydraulics, core neutronics, transient analysis, LOCA and stability) are assessed to determine any challenges to EPU application. A description of the analyses performed for a transition cycle is provided in Appendix A.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 3-1 3.0 Thermal-hydraulics AREVA assembly thermal-hydraulic methods are qualified and validated against full-scale heated bundle tests in the KATHY test facility in Karlstein, Germany. The KATHY tests are used to characterize the assembly two-phase pressure drop and CHF performance. This allows the hydraulic models to be verified for AREVA fuel designs over a wide range of hydraulic conditions prototypic of reactor conditions.

The standard matrix of test conditions for KATHY is compared to reactor conditions in Figure 3-1. This figure illustrates that the test conditions bound uprated assembly conditions.

These tests were performed on the same ATRIUMTm-1 0" fuel assembly that is to be loaded in the Brunswick core. The Brunswick data is based upon actual operating data at the uprated condition except B1C17. Figure 3-1 also shows that the key physical phenomena (e.g. fluid quality and assembly flows) for uprated conditions are within the scope of and bounded by current reactor experience. Figure 3-2 presents this same data from a core design with all ATRIUM-1 0 fuel at uprated conditions with an expanded power/flow map.

This similarity of assembly conditions is further enforced in AREVA analysis methodologies by the imposition of SPCB CHF correlation limits and therefore both current designs and uprated designs must remain within the same parameter space. Since the bundle operating conditions for EPU are within the envelope of hydraulic test data used for model qualification and operating experience, the hydraulic models used to compute the core flow distribution and local void content remain valid for EPU conditions.

A more detailed discussion about the thermal hydraulic analyses is presented in Appendix B.

ATRIUM TM-10 is a trademark of AREVA NP Inc.

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Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Page 3-2 I

I Figure 3-1 Comparison of Karlstein Two-Phase Pressure Drop Test Matrix and Typical Reactor Conditions I

I Figure 3-2 Comparison of Karlstein Two-Phase Pressure Drop Test Matrix and Brunswick Reactor Conditions with ATRIUM-10 Fuel AREVA NP Inc-

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 4-1 4.0 SPCB CHF Correlation The SPCB CHF correlation was approved by the NRC staff to be applicable over specified ranges of mass flow, pressure, inlet subcooling, and boiling transition enthalpy. The NRC staff also approved specific corrective actions when the computed conditions fall outside of the approved range to assure that conservative calculations are obtained. For both EPU and pre-EPU conditions, some analyses can predict assembly conditions to be outside the approved range of specified conditions for the CHF correlations. Consequently, the AREVA licensing methods are programmed to determine whether the computed assembly conditions fall outside of the approved range of applicability for the CHF correlation; if so, the methodology imposes approved corrective actions to conservatively assess the critical power margin for the assembly.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 5-1 5.0 Safety Limit MCPR The safety limit MCPR (SLMCPR) methodology is used to determine the Technical Specification SLMCPR value that ensures that 99.9% of the fuel rods are expected to avoid boiling transition during normal reactor operation and anticipated operation occurrences. The SLMCPR methodology is described in Reference 1. The SLMCPR is determined by statistically combining calculation uncertainties and plant measurement uncertainties that are associated with the calculation of MCPR. The thermal hydraulic, neutronic, and critical power correlation methodologies are used in the calculation of MCPR. The applicability of these methodologies for EPU conditions is discussed in other sections of this report. As discussed in Section 2.0, EPU operation will result in a flatter radial power distribution with more fuel rods operating near thermal limits. This is the most significant impact of EPU operation on the SLMCPR calculation and is explicitly accounted for in the methodology as discussed below.

AREVA calculates the SLMCPR on a cycle-specific basis to protect all allowed reactor operating conditions. The analysis incorporates the cycle-specific fuel and core designs. The initial MCPR distribution of the core is a major factor affecting how many rods are predicted to be in boiling transition. The MCPR distribution of the core depends on the neutronic design of the reload fuel and the fuel assembly power distributions in the core. AREVA SLMCPR methodology specifies that analyses be performed with a design basis power distribution that

"... conservatively represents expected reactor operating states which could both exist at the MCPR operating limit and produce a MCPR equal to the MCPR safety limit during an anticipated operational occurrence." (Reference 1, Section 5.0).

[

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Unrate Conditions ANP-2638NP Revision 0 Paae 5-2 I

I The impact that a flatter core power distribution may have on the SLMCPR is explicitly accounted for by the methodology. EPU operation will lead to a flatter core power distribution;

[

I AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Paae 5-3 I

I Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 6-1 6.0 Mechanical Limits Methodology The LHGR limit is established to support plant operation while satisfying the fuel mechanical design criteria. The LHGR limit is translated into power history inputs as described in the Reference 2 topical report (see the response to RAI, Question 3, in Reference 2). Then, the power histories are used as input to the RODEX2A, RAMPEX and COLAPX codes in evaluating the fuel rod thermal-mechanical criteria (References 3 and 4).

Higher voiding can have the effect of further offsetting the axial power and fast neutron flux profiles during operation. Because the power history methodology makes use of established axial profiles, changes to the power history due to higher voiding are not explicitly taken into account in the analyses. However, the profiles selected for the analyses are conservatively peaked to result in higher rod average power levels while attaining the LHGR limit.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 7-1 7.0 Core Neutronics The AREVA neutronic methodologies are characterized by technically rigorous treatment of phenomena and are very well benchmarked (>100 cycles of operation plus gamma scan data for ATRIUM-10). Recent operating experience is tabulated in Table 7-1 and Table 7-2. These tables present the reactor operating conditions and in particular the average and hot assembly powers for both US and European applications. As can be seen from this information, the average and peak bundle powers in this experience base exceed that associated with the power uprate for Brunswick.

The increased steam flow from power uprate comes from increased power in normally lower power assemblies in the core, operating at higher power levels. High powered assemblies in uprated cores will be subject to the same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits and restrictions as high powered assemblies in non-uprated cores.

The similarity of operating conditions between current and uprate conditions assures that the neutronic methods used to compute the nodal reactivity and power distributions remain valid.

Furthermore, the neutronic characteristics computed by the steady-state simulator and used in safety analysis remain valid.

Detailed analysis of the neutronic methodology and it is specific applicability to Brunswick is presented in Appendix C.

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Urrate Conditions ANP-2638NP Revision 0 Page 7-2 Table 7-1 CASMO-41MICROBURN-B2 Operating Experience Ave.

Approximate Bundle Peak Bundle Reactor Power, MWt

Power, Power, Fuel/Cycle Reactor Size #FA

(% Uprated)*

MWt/FA MWt/FA Licensing**

Uprate Comments A

592 2575 (0.0) 4.4 7.2 X

B 592 2575 (0.0) 4.4 7.4 X

C 532 2292 (0.0) 4.3 7.3 (X)

D 840 3690 (0.0) 4.4 7.5 X

E 500 2500 (15.7) 5.0 8.0 X

For 3 cycles oper.

F 444 1800 (5.9) 4.1 7.3 X

G 676 2928 (8.0) 4.3 7.4 (X)

H 700 3300 (9.3) 4.7 8.0 (X)I(X)

I 784 3840 (0.0) 4.9 8.1 (X)

J 624 3237 (11.9) 5.2 7.8 (X)

K 648 3600 (14.7) 5.6 8.6 (X)

With ATRIUM-10XM L

648 2500 (10.1) 3.9 6.9 (X)

M 624 3091 (6.7) 5.0 7.7 X

N 800 3898 (1.7) 4.9 7.7 X

0.

764 3489 (5.0) 4.6 7.2 X

Brunswickt 560 2923 (20.0) 5.2 8.0 X

Previously licensed I _by GNF

t t*

Latest power uprates.

(x)=currently fuel licensing only (Europe).

Brunswick added for comparison purposes only, i.e., not In the Operating Experience database.

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paqe 7-3 Table 7-2 CASMO-4/MICROBURN-B2 Other Planned Uprate Support Ave.

Approximate Bundle Peak Bundle Reactor Power, MWt

Power, Power, Fuel/Cycle Uprate Reactor Size #FA

(% Uprated)

MWt/FA MWt/FA Licensing*

Comments G

676 3253 (20.0) 4.8 7.8 (X)

In 2009 1

784 4000 (4.2) 5.1 8.1 (X)

In 2008 H

700 3896 (29.0) 5.6 8.6 (X)

P 764 3952 (20.0) 5.2 7.3 (X)

Analytical Study F

444 2300 (35) 5.2 7.7 Analytical Study (x)=currently fuel licensing only (Europe).

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 8-1 8.0 Transient Analysis The core phenomena of primary interest for limiting transients in BWRs are void fraction/quality relationships, determination of CHF, pressure drop, reactivity feedbacks and heat transfer correlations. One fundamental validation of the core hydraulic solution is separate effects testing against Karlstein transient CHF measurements. The transient benchmark to time of dryout for prototypic Load Reject with no Bypass (LRNB) and pump trip transients encompass the transient integration of the continuity equations (including the void-quality closure relation),

heat transfer, and determination of CHF. Typical benchmarks to Karlstein (Figure 8-1) illustrate that the transient hydraulic solution and application of SPCB result in conservative predictions of the time of dryout. The measured data is taken from ATRIUM-10 tests.

Outside of the core, the system simulation relies primarily on solutions of the basic conservation equations and equations of state. While there are changes to the feedwater flow rate and jet pump M-ratio associated with power uprate, the most significant change is the steam flow rate and the associated impact on the steamline dynamics for pressurization events. The models associated with predicting the pressure wave are general and have no limitation within the range of variation associated with power uprate.

The reactivity feedbacks are validated by a variety of means including initial qualification of advanced fuel design lattice calculations to Monte Carlo results as required by SER restrictions, steady-state monitoring of reactor operation (power distributions and eigenvalue), and the Peach Bottom 2 turbine trip benchmarks that exhibited a minimum of 5% conservatism in the calculation of integral power.

From these qualifications and the observation that the nodal hydraulic conditions during EPU are expected to be within the current operating experience, the transient analysis methods remain valid.

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page 8-2 I

Figure 8-1 Typical Hydraulic Benchmarks to Karlstein Transient Simulations AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 9-1 9.0 LOCA Analysis LOCA results are strongly dependent on local power and are weakly dependent on core average power. As discussed in previous sections, maximum local power is not significantly changed due to power uprate because the core is still constrained by the same thermal limits.

The parameters associated with power uprate that may impact LOCA results are: increased core average initial stored energy, decreased initial coolant inventory, relative flow distribution between highest power and average power assemblies, and increased core decay heat.

BWR LOCA analyses are not sensitive to initial stored energy. During the blowdown phase the heat transfer remains high and the stored energy is removed prior to the start of the heatup phase. Initial inventory differences may impact LOCA event timing and the minimum inventory during blowdown prior to refill of the reactor vessel. However, any impact on event timing or minimum inventory would be smaller than the impact associated with the different size breaks that are already considered in the break spectrum analyses. At EPU conditions, the difference in flow between the highest power assembly and the average power assembly is reduced relative to pre-EPU conditions. Therefore, these parameters do not change the range of conditions encountered or the capability of the codes to model LOCA at EPU conditions.

The potential impact of power uprate on LOCA analyses is thus primarily associated with the increase in decay heat levels in the core. Decay heat is conservatively modeled using industry standards applied as specified by regulatory requirements. The models used for decay heat calculations are valid for EPU.

From the above discussion and the observation that nodal thermal-hydraulic conditions during EPU are expected to be within the current operating experience, the LOCA methods remain applicable for EPU conditions.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 10-1 10.0 Stability Analysis The flatter radial power profile induced by the power uprate will tend to decrease the first azimuthal elgenvalue separation and result in slightly higher regional decay ratios. These effects are computed by STAIF as it directly computes the channel, global, and regional decay ratio and does not rely on a correlation to protect the regional mode.

STAIF has been benchmarked against full assembly tests (in KATHY facility) to validate the channel hydraulics from a decay ratio of approximately 0.4 to limit cycles. These tests or benchmarks exceed the bounds of EPU operation. These benchmarks include prototypical ATRIUM-10 assemblies. From a reactor perspective, STAIF is benchmarked to both global and regional reactor data as late as 1998, and therefore includes current reactor cycle and fuel design elements. This strong benchmarking qualification and the direct computation of the regional mode assure that the impact of the "flatter" core design for power uprate will be reflected in the stability analysis.

Detailed analysis of the stability methodology and it's specific applicability to Brunswick are presented in Appendix D.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 11-1 11.0 Summary This review concluded that there are no SER restrictions on AREVA methodology that are impacted by EPU. Since the EPU core and assembly conditions for the Brunswick units are equivalent to core and assembly conditions of other plants for which the methodology was benchmarked, the AREVA methodology (including uncertainties) remains applicable for EPU conditions at the Brunswick Units.

More specifically:

a) The steady state and transient neutronics and thermal-hydraulic analytical methods and code systems supporting EPU are within NRC approved applicability ranges because the conditions for EPU application are equivalent to existing core and assembly conditions in other plants for which the AREVA methodology was benchmarked.

b) The calculational and measurement uncertainties applied in EPU applications are valid because the conditions for EPU application are equivalent to existing core and assembly conditions for which the AREVA methodology was benchmarked.

c) The assessment database and uncertainty of models used to simulate the plant response at EPU conditions are equivalent to core and assembly conditions for which the AREVA methodology was benchmarked.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page 12-1 12.0 References

1. ANF-524(P)(A) Revision 2 and Supplements 1 and 2, "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation, November 1990.

2. XN-NF-85-67(P)(A) Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, September 1986.

3. EMF-85-74(P) Revision 0, Supplement 1(P)(A) and Supplement 2(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.

4. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page A-1 Appendix A Application of AREVA Methodology for Mixed Cores A.1 Discussion AREVA has considerable experience analyzing fuel design transition cycles and has methodology and procedures to analyze mixed cores composed of multiple fuel types. For each core design, analyses are performed to confirm that all design and licensing criteria are satisfied. The analyses performed explicitly include each fuel type in the core. The analyses consider the cycle-specific core loading and use input data appropriate for each fuel type in the core. The mixed core analyses are performed using generically approved methodology in a manner consistent with NRC approval of the methodology. Based on results from the analyses, operating limits are established for each fuel type present in the core. During operation, each fuel type is monitored against the appropriate operating limits.

Thermal hydraulic characteristics are determined for each fuel type that will be present in the core. The thermal hydraulic characteristics used in core design, safety analysis, and core monitoring are developed on a consistent basis for both AREVA fuel and other vendor co-resident fuel to minimize variability due to methods. Geometric data for the co-resident fuel is obtained from the utility, generally under a proprietary agreement. The hydraulic characteristics are based on flow test measurements performed for both AREVA fuel and co-resident fuel in the AREVA hydraulic flow test facility. For analyses performed to assess core thermal and hydraulic performance with the XCOBRA computer code, each fuel type in the core is explicitly modeled with the appropriate geometric data and hydraulic characteristics.

The GE-14 fuel design will be present during the ATRIUM-10 transition cycles at Brunswick. A GE-14 fuel design was flow tested by AREVA in 2002. The GE-14 design that will be co-resident with the ATRIUM-10 fuel at Brunswick was confirmed to be thermal hydraulically the same as the GE-14 assembly tested by AREVA.

For core design and nuclear safety analyses, the neutronic cross-section data is developed for each fuel type in the core using CASMO-4. Geometric and nuclear design data (e.g.,

enrichment distribution) that is required to prepare CASMO-4 input for the co-resident fuel is obtained from the utility, generally under a proprietary agreement. MICROBURN-B2 is used to AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page A-2 design the core and provide input to safety analyses (core neutronic characteristics, power distributions, etc.). Each fuel assembly is explicitly modeled in MICROBURN-B2 using cross-section data from CASMO-4 and geometric data appropriate for the fuel design.

Fuel assembly thermal mechanical limits for both AREVA and co-resident fuel are verified and monitored for each mixed core designed by AREVA. The thermal mechanical limits established by the co-resident fuel vendor continue to be applicable for mixed (transition) cores. The thermal mechanical limits (steady-state and transient) for the co-resident fuel are provided to AREVA by the utility. AREVA performs design and licensing analyses to demonstrate that the core design meets steady-state limits and that transient limits are not exceeded during anticipated operational occurrences.

The critical power ratio (CPR) is evaluated for each fuel type in the core using calculated local fluid conditions and an appropriate critical power correlation. Fuel type specific correlation coefficients for AREVA fuel are based on data from the AREVA critical power test facility.

Unless the co-resident fuel critical power correlation is provided to AREVA, the AREVA critical power correlation is applied to the co-resident fuel. An NRC-approved process for developing correlation coefficients and the associated uncertainty when applying an AREVA critical power correlation for co-resident fuel is described in Reference A.1. If adequate test data is available for the co-resident fuel, the correlation coefficients and uncertainty are determined using the process consistent with the approval of the correlation. When test data for the co-resident fuel is not available to AREVA, fuel type specific correlation coefficients and uncertainty are developed using calculated CPR data provided by the utility based on an alternate (indirect) approach described in Reference A.1.

The SPCB critical power correlation will be used for monitoring GE-14 fuel present during the ATRIUM-10 transition cycles at Brunswick. Consistent with the NRC-approved methodology, the indirect approach from Reference A.1 was used to develop correlation coefficients and uncertainty based on CPR data provided to AREVA by Progress Energy.

Analyses performed to determine the safety limit MCPR explicitly address mixed core effects.

Each fuel type present in the core is explicitly modeled using appropriate geometric data, thermal hydraulic characteristics, and power distribution information (from CASMO-4 and MICROBURN-B2 analyses). CPR is evaluated for each assembly using fuel type specific AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page A-3 SPCB correlation coefficients. Plant and fuel type specific uncertainties are considered in the statistical analysis performed to determine the safety limit MCPR. The safety limit MCPR analysis is performed each cycle and uses the cycle specific core configuration.

An operating limit MCPR is established for each fuel type in the core. For fast transients the COTRANSA2 code (Reference A.2) is used to determine the overall system response. The core nuclear characteristics used in COTRANSA2 are obtained from MICROBURN-B2 and reflect the actual core loading pattern. Boundary conditions from COTRANSA2 are used with an XCOBRA-T core model. In the XCOBRA-T model, a hot channel with appropriate geometric and thermal hydraulic characteristics is modeled for each fuel type present in the core. Critical power performance is evaluated using local fluid conditions and fuel type specific CPR correlation coefficients. The transient CPR response is used to establish an operating limit MCPR for each fuel type.

For transient events that are sufficiently slow such that the heat transfer remains in phase with changes in neutron flux during the transient, evaluations are performed with MICROBURN-B2 in accordance with NRC approval. Such slow transients are modeled with the MICROBURN-B2 core simulator code by performing a series of steady state solutions with appropriate boundary conditions using the cycle specific design core loading plan. Each fuel assembly type in the core is explicitly modeled. The change in CPR between the initial and final condition after the transient is determined, and if the CPR change is more severe than those determined from fast transient analyses, the slow transient result is used to determine the MCPR operating limit.

Stability analyses to establish OPRM setpoints and backup stability exclusion regions are performed using the cycle-specific core loading pattern. The stability analyses performed with RAMONA5-FA and STAIF explicitly model each fuel type in the core. Each fuel type is modeled using appropriate geometric, thermal hydraulic, and nuclear characteristics determined as described above. The stability OPRM setpoints and exclusion region boundaries are established based on the predicted performance of the actual core composition.

MAPLHGR operating limits are established and monitored for each fuel type in the core to ensure that 10 CFR 50.46 acceptance criteria are met during a postulated LOCA. The AREVA LOCA methodology is used to establish MAPLHGR limits for AREVA fuel. The RELAX code is used to determine the overall system response during a postulated LOCA and provides AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page A-4 boundary conditions for a RELAX hot channel model. Each fuel type in the core can be represented by a RELAX hot channel model with appropriate geometric and thermal hydraulic characteristics. Results from the hot channel analysis provide boundary conditions to the HUXY computer code. The HUXY model includes fuel type specific input such as dimensions and local power peaking for each fuel rod.

The core monitoring system will monitor each fuel assembly in the core. Each assembly is modeled with geometric, thermal hydraulic, neutronic, and CPR correlation input data appropriate for the specific fuel type. Each assembly in the core will be monitored relative to thermal limits that have been explicitly developed for each fuel type.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page A-5 In summary, AREVA methodology is used consistent with NRC approval to perform design and licensing analyses for mixed cores. The cycle design and licensing analyses explicitly consider each fuel type in mixed core configurations. Co-resident fuel input parameters are developed consistent with the methodology, and in general are developed in the same manner as for AREVA fuel. Limits are established for each fuel type and operation within these limits is verified by the monitoring system during operation.

A.2 References A.1.

EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel", Siemens Power Corporation, August 2000.

A.2. ANF-913 (P)(A) Volume 1 Revision 1 and Volume I Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses,"

Advanced Nuclear Fuels Corporation, August 1990.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page B-1 Appendix B Thermal Hydraulic Methods B.1 Hydraulic Characterization The Zuber-Findlay drift flux model (Reference B.1) is utilized in the AREVA nuclear and safety analysis methods for predicting vapor void fraction in the BWR system. The model has a generalized form that may be applied to two phase flow by defining an appropriate correlation for the void concentration parameter, Co, and the drift flux, Vgj. The model parameters account for the radially non-uniform distribution of velocity and density and the local relative velocity between the phases, respectively. This model has received broad acceptance in the nuclear industry and has been successfully applied to a host of different applications, geometries, and fluid conditions through the application of different parameter correlations (Reference B.2).

Two different correlations are utilized at AREVA to describe the drift flux parameters for the analysis of a BWR core. The correlations and treatment of uncertainties are as follows:

The nuclear design, frequency domain stability, nuclear AOO transient and accident analysis methods use the [

] void correlation (Reference B.3) to predict nuclear parameters. Uncertainties are addressed at the overall methodology and application level rather than individually for the individual correlations of each method.

The overall uncertainties are determined statistically by comparing predictions using the methods against measured operating data for the reactors operating throughout the world.

The thermal-hydraulic design, system AOO transient and accident analysis, and loss of coolant accident (only at specified junctions) methods use the Ohkawa-Lahey void correlation (Reference B.4). This correlation is not used in the direct computation of nuclear parameters in any of the methods. Uncertainties are addressed at the overall methodology level through the use of conservative assumptions and biases to assure uncertainties are bounded.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page B-2 The [

] void correlation was developed for application to multi-rod geometries operating at typical BWR operating conditions using multi-rod data and was also validated against simple geometry data available in the public domain. The correlation was defined to be functionally dependent on the mass flux, hydraulic diameter, quality, and fluid properties.

The multi-rod database used in the [

]. As a result, the multi-rod database and prediction uncertainties are not available to AREVA. However, the correlation has been independently validated by AREVA against public domain multi-rod data and proprietary data collected for a prototypical ATRIUM-10 test assembly. Selected results for the ATRIUM-10 test assembly are reported in the public domain in Reference B.5.

The Ohkawa-Lahey void correlation was developed for application in BWR transient calculations. In particular, the correlation was carefully designed to predict the onset of counter current flow limit (CCFL) characteristics during the occurrence of a sudden inlet flow blockage.

The correlation was defined to be functionally dependent on the mass flux, quality, and fluid properties.

Independent validation of the correlation was performed by AREVA at the request of the NRC during the NRC review of the XCOBRA-T code. The NRC staff subsequently reviewed and approved Reference B.6, which compared the code to a selected test from the FRIGG experiments (Reference B.7). More recently the correlation has been independently validated by AREVA against additional public domain multi-rod data and proprietary data collected for a prototypic ATRIUM-1 0 test assembly.

The characteristics of the AREVA multi-rod void fraction validation database are listed in Table B-I.

The FRIGG experiments have been included in the validating database because of the broad industry use of these experiments in benchmarking activities, including TRAC, RETRAN, and S-RELAP5. The experiments include a wide range of pressure, subcooling, and quality from which to validate the general applicability of a void correlation. However, the experiments do not contain features found in modern rod bundles such as part length fuel rods and mixing vane grids. The lack of such features makes the data less useful in validating correlations for modern AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page B-3 fuel designs. Also the reported instrument uncertainty for these tests is provided in Table B-1 based on mockup testing. However, the total uncertainty of the measurements (including power and flow uncertainties) is larger than the indicated values.

Because of its prototypical geometry, the ATRIUM-10 void data collected at KATHY was useful in validating void correlation performance in modern rod bundles that include part length fuel rods, mixing vane grids, and prototypic axial/radial power distributions. Void measurements were made at one of three different elevations in the bundle for each test point: just before the end of the part length fuel rods, midway between the last two spacers, and just before the last spacer.

As shown in Figure B-I, the range of conditions for the ATRIUM-10 void data are valid for typical reactor conditions. This figure compares the equilibrium quality at the plane of measurement for the ATRIUM-10 void data with the exit quality of bundles in the EMF-2158 benchmarks and Brunswick operating at EPU conditions. As seen in the figure, the data at the measurement plane covers nearly the entire range of reactor conditions. However, calculations of the exit quality from the void tests show the overall test conditions actually envelope the reactor conditions. (Note, the ATRIUM-10 data shown in Figure B-1 is not from the same database as illustrated in Table B-I.)

Figure B-2 and Figure B-3 provide comparisons of predicted versus measured void fractions for the AREVA multi-rod void fraction validation database using the [

] correlation.

These figures show the predictions fall within +/-0.05 (predicted - measured) error bands with good reliability and with very little bias. Also, there is no observable trend of uncertainty as a function of void fraction.

Figure B-4 and Figure B-5 provide comparisons of predicted versus measured void fractions for the AREVA multi-rod void fraction validation database using the Ohkawa-Lahey correlation.

In general, the correlation predicts the void data with a scatter of about +/-0.05 (predicted -

measured), but a bias in the prediction is evident for voids between 0.5 and 0.8. The observed under prediction is consistent with the observations made in Reference B.8.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page B-4 In conclusion, validation using the AREVA multi-rod void fraction validation database has shown that both drift flux correlations remain valid for modern fuel designs. Furthermore, there is no observable trend of uncertainty as a function of void fraction. This shows there is no increased uncertainty in the prediction of nuclear parameters at EPU (including an extended power/flow operating map) conditions within the nuclear methods as a result of changes to the population distribution of the nodal void fractions with respect to pre EPU conditions.

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Applicability of AREVA NP BWR Methods to Extended Power

[Unrnfe Conditions ANP-2638NP Revision 0 Paae B-5 Table B-1 AREVA Multi-Rod Void Fraction Validation Database FRIGG-3 FRIGG-2 ATRIUM-10-KATHY (Reference B.7 and (Reference 8.9)

B8 1B.8)

Axial Power Shape uniform uniform

]

Radial Power Peaking uniform mild peaking circular array with 36 circular array with 36 prototypic Bundle Design ATRIUM-IO CHE rods + central thimble rods + central thimble bundle Pressure (psi) 725 725, 1000, and 1260

[

]

Inlet Subcooling (OF) 4.3 to 40.3 4.1 to 54.7

[

]

Mass Flow Rate (Ibm/s)

(calculated from mass flux 14.3 to 31.0 10.1 to 42.5[]

assuming ATRIUM-10 inlet flow area)

Equilibrium Quality at Measurement Plane

-0.036 to 0.203

-0.058 to 0.330 (fraction)

Max Void at Measurement 0.828 0.848 Plane (fraction)

Reported Instrument 0.025 0.016

[

]

Uncertainty (fraction)

Number of Data 27 tests, 174 points 39 tests, 157 points

[

]

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page B-6

[

I Figure B-1 Comparison of the Measured Local Quality for ATRIUM-10 Void Data and Exit Quality for Typical Reactor Conditions I

Figure B-2 Validation of [

] using FRIGG-2 and FRIGG-3 Void Data AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Page B-7 I

I Figure B-3 Validation of [

] using ATRIUM-10 Void Data I

Figure B-4 Validation of Ohkawa-Lahey using FRIGG-2 and FRIGG-3 Void Data AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Page B-8

[

I Figure B-5 Validation of Ohkawa-Lahey using ATRIUM-10 Void Data B.2 References B.1.

N. Zuber and J. A. Findlay, "Average Volumetric Concentration in Two-Phase Flow Systems," J. Heat Transfer, 1965.

B.2. P. Coddington and R. Macian, "A Study Of The Performance Of Void Fraction Correlations Used In The Context Of Drift-Flux Two-Phase Flow Models," Nuclear Engineering and Design, 215, 199-216, June 2002.

B.3.

B.4. K. Ohkawa and R. T. Lahey, Jr., "The Analysis of CCFL Using Drift-Flux Models," Nuclear Engineering and Design, 61, 1980.

B.5. S. Misu et al., "The Comprehensive Methodology for Challenging BWR Fuel Assembly and Core Design used at FANP," proceedings on CD-ROM, PHYSOR 2002, Seoul, Korea, October 7-10, 2002 B.6. XN-NF-84-105(P)(A) Volume 1 Supplement 4, "XCOBRA-T: A Computer Code For BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data," Advanced Nuclear Fuels Corporation, June 1988.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page B-9 B.7. 0. Nylund et al., "Hydrodynamic and Heat Transfer Measurements on A Full-Scale Simulated 36-Rod Marviken Fuel Element with Non-Uniform Radial Heat Flux Distribution," FRIGG-3, R-494/RL-1154, November 1969.

B.8. J. Skaug et al., "FT-36b, Results of void Measurements," FRIGG-PM-15, May 1968.

B.9. 0. Nylund et al., "Hydrodynamic And Heat Transfer Measurements On A Full-Scale Simulated 36-Rod Marviken Fuel Element With Uniform Heat Flux Distribution," FRIGG-2, R-447/RTL-1007, May 1968.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-1 Appendix C Neutronic Methods C.1 Cross Section Representation CASMO-4 performs a multi-group [

] spectrum calculation using a detailed heterogeneous description of the fuel lattice components. Fuel rods, absorber rods, water rods/channels and structural components are modeled explicitly. The library has cross sections for [

]

materials including [

] heavy metals. Depletion is performed with a predictor-corrector approach in each fuel or absorber rod. The two-dimensional transport solution is based upon the [

]. The solution provides pin-by-pin power and exposure distributions, homogeneous multi-group (2) micro-scopic cross sections as well as macro-scopic cross sections. Discontinuity factors are determined from the solution. [

]

gamma transport calculation are performed. The code has the ability to perform [

] calculations with different mesh spacings. Reflector calculations are easily performed.

MICROBURN-B2 performs microscopic fuel depletion on a nodal basis. The neutron diffusion equation is solved with a full two energy group method. A modern nodal method solution using discontinuity factors is used along with a [

]. The flux discontinuity factors are [

]. A multilevel iteration technique is employed for efficiency.

MICROBURN-B2 treats a total of [

] heavy metal nuclides to account for the primary reactivity components. Models for nodal [

] are used to improve the accurate representation of the in-reactor configuration. A full three-dimensional pin power reconstruction method is utilized. TIP (neutron and gamma) and LPRM response models are included to compare calculated and measured instrument responses.

Modern steady state thermal hydraulics models define the flow distribution among the assemblies. [

] based upon CASMO-4 calculations are used for the in-channel fluid conditions as well as in the bypass and water rod regions. Modules for the calculation of CPR, LHGR and MAPLHGR are implemented for direct comparisons to the operating limits.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-2 MICROBURN-B2 determines the nodal macroscopic cross sections by summing the contribution of the various nuclides.

I Z,(p,fl,E,R) = LNjo.(p,fl,E,R)+AZ (p, r, F, R) where:

Z X

= nodal macroscopic cross section AX*

= background nodal macroscopic cross section (D, Zt, Z., Z,)

Ni

= nodal number density of nuclide "i" i

= microscopic cross section of nuclide 'I" I

= total number of explicitly modeled nuclides p

= nodal instantaneous coolant density rI

= nodal spectral history E

= nodal exposure R

= control fraction The functional representations of o-and A, come from 3 void depletion calculations with CASMO-4. Instantaneous branch calculations at alternate conditions of void and control state are also performed. The result is a multi dimensional table of microscopic and macroscopic cross sections that is shown in Figure C-1 and Figure C-2.

At BOL the relationship is fairly simple; the cross section is only a function of void fraction (water density) and the reason for the variation is the change in the spectrum due to the water density variations. At any exposure point, a quadratic fit of the three CASMO-4 data points is used to represent the continuous cross section over instantaneous variation of void or water density.

This fit is shown in Figure C-3 and Figure C-4.

Detailed CASMO-4 calculations confirm that a quadratic fit accurately represents the cross sections as shown in Figure C-5, Figure C-6, and Figure C-7.

With depletion the isotopic changes cause other spectral changes. Cross sections change due to the spectrum changes. Cross sections also change due to self shielding as the concentrations change. These are accounted for by the void (spectral) history and exposure parameters. Exposure variations utilize a piecewise linear interpolation over tabulated values at AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-3

[

] exposure points. The four dimensional representation can be reduced to three dimensions (see Figure C-8) by looking at a single exposure.

Quadratic interpolation is performed in each direction independently for the most accurate representation. Considering the case at 70 GWd/MTU with an instantaneous void fraction of 70% and a historical void fraction of 60%, Figure C-9 and Figure C-10 illustrate the interpolation process. The table values from the library at 0, 40 and 80 % void fractions are used to generate 3 quadratic curves representing the behavior of the cross section as a function of the historical void fraction for each of the tabular instantaneous void fractions (0, 40 and 80 %).

The intersection of the three quadratic lines with the historical void fraction of interest are then used to create another quadratic fit in order to obtain the resultant cross section as shown in Figure C-10.

The results of this process for all isotopes and all cross sections In MICROBURN-B2 were compared for an independent CASMO-4 calculation with continuous operation at 20, 60 and 90% void and are presented in Figure C-11. Branch calculations at 90% void from a 40% void depletion were performed for multiple exposures. The results show very good agreement for the whole exposure range as shown in Figure C-12.

At the peak reactivity point, multiple comparisons were made (Figure C-13) to show the results for various instantaneous void fractions.

Void fraction has been used for the previous illustrations; however MICROBURN-B2 uses water density rather than void fraction in order to account for pressure changes as well as sub-cooled density changes. This transformation does not change the basic behavior as water density is proportional to void fraction. MICROBURN-B2 uses spectral history rathbr than void history in order to account for other spectral influences due to actual core conditions (fuel loading, control AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-4 rod inventory, leakage, etc.) The Doppler feedback due to the fuel temperature is modeled by accumulating the Doppler broadening of microscopic cross sections of each nuclide.

where:

Te11 = Effective Doppler Fuel Temperature Trj = Reference Doppler Fuel Temperature o-= Microscopic Cross Section (fast and thermal absorption) of nuclide "i" Ni

= Density of nuclide "i" The partial derivatives are determined from branch calculations performed with CASMO-4 at various exposures and void fractions for each void history depletion. The tables of cross sections include data for [

] states. The process is the same for

[

] states. Other important feedbacks to nodal cross sections are lattice [

] and instantaneous [

J between lattices of different [

]. These feedbacks are modeled in detail.

The methods used in CASMO-4 are state of the art. The methods used in MICROBURN-B2 are state of the art. The methodology accurately models a wide range of thermal hydraulic conditions including EPU and extended power/flow operating map conditions.

C.2 Applicability of Uncertainties The EMF-2158(P)(A) data was re-evaluated by looking at the deviations between measured and calculated TIP response for each axial level. The standard deviation of these deviations at each axial plane are presented in Figure C-1 5 and demonstrate that there is no significant trend vs.

axial position, which indicates no significant trend vs. void fraction.

To compare core physics models to the gamma scan results, the calculated pin power distribution Is converted into a Ba140 density distribution. A rigorous mathematical process using CASMO-4 pin nuclide inventory and MICROBURN-B2 nodal nuclide inventory is used.

Gamma scan comparisons for 9X9-1 and ATRIUM-10 fuel were presented in the topical report, EMF-2158(P)(A), in section 8.2.2. Figures 8.18 through 8.31 showed very good comparisons AREVA NP Inc

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-5 between the calculated and measured relative Ba-140 density distributions for both radial and axial values.

The Quad Cities assembly gamma scan data was used to determine the correlation coefficient which accounts for the correspondence between the assembly powers of adjacent assemblies.

This correspondence is quantified by a conservative multiplier to the uncertainty in the TIP measurements. In order to conservatively account for this correspondence, the bundle power uncertainty is increased due to the radial TIP uncertainty by a multiplier based on the correlation coefficient. The correlation coefficient is statistically calculated and shown in Figure 9.1 and Figure 9.2 of EMF-2158(P)(A). It indicates a less than perfect correlation between powers of neighboring bundles. The conservative multiplier is calculated as follows:

The calculated TIP uncertainty would normally be expected to be slightly larger than the calculated power uncertainty due to the TIP model. The Quad Cities gamma scan comparison shows the 2-D radial power uncertainty of [

] (see Section 9.6 of EMF-2158(P)(A)). The D-Lattice plant calculated radial TIP uncertainty is

[

]. The data indicates that the calculated TIP uncertainty is indeed larger than the calculated power uncertainty. The use of the correlation coefficient to increase the calculated power uncertainty is a very conservative approach resulting from the statistical treatment. The types of fuel bundles (8x8, 9x9, or I Ox1 0) loaded in the core has no effect on the reality of the physical model which precludes the possibility of the calculated power uncertainty to be larger than the calculated TIP uncertainty. The accuracy of the MICROBURN-B2 models is demonstrated by comparisons between measured and calculated TIP's as well as comparison of calculated and measured Lanthanum-140 activation. The accuracy of the MICROBURN-B2 models was further validated with detailed axial pin by pin gamma scan measurements of 9X9-1 and ATRIUM-1 0 fuel assemblies in the reactor designated as KWU-S. These measurements demonstrated the continued accuracy of the MICROBURN-B2 models with modem fuel AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-6 assemblies. Details of these measurements are provided in Section 8.2.2 of the topical report, EMF-2158(P)(A).

The AREVA SAFLIM2 code is used to calculate the number of expected rods in boiling transition (BT) for a specified value of the SLMCPR (i.e., SLMCPR is an input, not a calculated result). The extremes of the two correlation coefficients from the Quad Cities assembly gamma scan data sets [

] discussed in section C.2 were used for a sensitivity study of the MCPR safety limit. An analysis of the safety limit was performed with SAFLIM2 using an input SLMCPR of 1.08 and the base RPF uncertainty. The number of boiling transition (BT) rods was calculated to be 60 from this analysis. The analysis was repeated in a series of SAFLIM2 calculations using the increased RPF uncertainty and performed by iterating on the input value of SLMCPR. Different values for the SLMCPR input were used until the number of BT rods calculated by SAFLIM2 was the same as the base case (60 rods). A SLMCPR input value of [

] resulted in 60 rods in BT when the increased RPF uncertainty was input.

The difference in SLMCPR input [

] for the two cases that resulted in the same number of BT rods is a measure of the safety limit sensitivity to the increased RPF uncertainty.

The only input parameters that changed between the two SAFLIM2 analyses were the SLMCPR and the RPF uncertainty. For each analysis, 1000 Monte Carlo trials were performed. To minimize statistical variations in the sensitivity study, the same random number seed was used and all bundles were analyzed for both analyses. As discussed above, 60 rods were calculated to be in BT in both analyses.

This sensitivity study was performed to quantify the sensitivity of SLMCPR to an increase in RPF uncertainty and did not follow the standard approach used in SLMCPR licensing analyses.

In standard licensing calculations, the SLMCPR is not input at a precision greater than the hundredths decimal place. As a result, the increased RPF uncertainty would result [

] in SLMCPR licensing analyses depending on how close the case was to the acceptance criterion prior to the increase in RPF uncertainty.

Gamma scanning provides data on the relative gamma flux from the particular spectrum associated with La140 gamma activity. The relative gamma flux corresponds to the relative La140 concentration. Based upon the time of shutdown and the time of the gamma scan the Ba140 relative distribution at the time of shutdown is determined. This Ba140 relative AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-7 distribution is thus correlated to the pin or assembly power during the last few weeks of operation. The data presented in the topical report, EMF-2158(P)(A), includes both pin and assembly Ba140 relative density data. The assembly gamma scan data was taken at Quad Cities after the operation of cycles 2, 3 and 4. Some of this data also included individual pin data. This data was from 7X7 and 8X8 fuel types. Additional fuel pin gamma scan data was taken at the Gundremingen plant for ATRIUM-9 and ATRIUM-10 fuel. This data is also presented in the topical report.

Pin-by-pin Gamma scan data is used for verification of the local peaking factor uncertainty.

Quad Cities measurements presented in the topical report EMF-2158(P)(A) have been re-evaluated to determine any axial dependency. Figure C-16 presents the raw data including measurement uncertainty and demonstrates that there is no axial dependency. The more recent Gamma scans performed by KWU, presented in the topical report EMF-2158(P)(A) and re-arranged by axial level in Table C-1, indicate no axial dependency. Full axial scans were performed on 16 fuel rods. Comparisons to calculated data show excellent agreement at all axial levels. The dip in power associated with spacers, observed in the measured data, Is not modeled in MICROBURN-B2. There is no indication of reduced accuracy at the higher void fractions.

CASMO-4 and MCNP calculations have been performed to compare the fission rate distribution statistics to Table 2-1 of the topical report EMF-2158(P)(A) which is shown in Table C-2. The fission rate differences at various void fractions demonstrate that CASMO-4 calculations have very similar uncertainties relative to the MCNP results for all void fractions. These fission rate differences also meet the criteria of the topical report EMF-2158(P)(A) for all void fractions.

Data presented in these figures and tables demonstrate that the AREVA methodology is capable of accurately predicting reactor conditions for fuel designs operated under the current operating strategies and core conditions.

C.3 Fuel Cycle Comparisons AREVA has reviewed the data presented in EMF-2158(P)(A) with regard to the maximum assembly power (Figure C-17) and maximum exit void fraction (Figure C-18) to determine the range of data previously benchmarked.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-8 Fuel loading patterns and operating control rod patterns are constrained by the minimum critical power ratio (MCPR) limit, which consequently limits the assembly power and exit void fraction regardless of the core power level. Operating data from several recent fuel cycle designs has been evaluated and compared to that in the topical report EMF-2158(P)(A). Maximum assembly power and maximum void fraction are presented in Figure C-19 and Figure C-20.

In order to evaluate some of the details of the void distribution a current design calculation was reviewed in more detail. Figure C-21 and Figure C-22 present the following parameters at the point of the highest exit void fraction (at 9336 MWd/MTU cycle exposure) in cycle core design for a BWR-6 reactor with ATRIUM-1 0 fuel. Another measure of the thermal hydraulic conditions is the population distribution of the void fractions. These are representative figures for a high power density plant.

" Core average void axial profile

" Axial void profile of the peak assembly Histogram of the nodal void fractions in core The actual core designs used for each cycle will have slightly different power distributions and reactivity characteristics than any other cycle. Conclusions from analyses that are dependent on the core design (loading pattern, control rod patterns, fuel types) are re-confirmed as part of the reload licensing analyses performed each cycle. Cycle-specific reload licensing calculations will continue to be performed for all future EPU cycles using NRC approved methodologies consistent with the current processes.

Brunswick operating under EPU conditions (Figure C-23 and Figure C-24) can be compared to the equivalent data of the topical report EMF-2158(P)(A). Comparison of Figure C-17 vs.

Figure C-23 and Figure C-18 vs. Figure C-24 shows that EPU operation in the standard powerlflow map is within the range of the original methodology approval for assembly power and exit void fraction. From a neutronic perspective, moderator density (void fraction) and exposure cause the greatest variation in cross sections. NRC-approved exposure limits for ATRIUM-10 fuel evaluated with AREVA methods are unchanged for EPU conditions. Reactor conditions for Brunswick with power uprate are not significantly different from that of current experience and are bounded by the experience for the important parameters.

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Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-9 The axial profile of the power and void fraction of the limiting assembly and core average values are presented in Figure C-25 for a Brunswick EPU cycle design. These profiles demonstrate that the core average void fraction and the maximum assembly power void fractions are bounded by the topical report data and are consistent with recent experience on other reactors.

Figure C-26 presents a histogram of the void fraction for EPU conditions. This histogram was taken at the point of maximum exit void fraction expected during the cycle. The distribution of voids is shifted slightly toward the 70 -80 % void fraction levels. The population of nodes experiencing 85 -90% voids is still small.

The neutronic and thermal hydraulic conditions predicted for the EPU operation are bounded by the data provided in the topical report EMF-2158(P)(A) so the isotopic validation continues to be applicable to EPU operation.

The AREVA methodology [

] the reactivity coefficients that are used in the transient analysis. Conservatisms in the methodology are used to produce conservative results that bound the uncertainties in the reactivity coefficients. Data presented in these referenced figures indicate that there are no significant differences between EPU and non-EPU conditions that have an impact on the reactivity coefficients.

C.3.1 Bypass Voiding The core bypass water is modeled in the AREVA steady-state core simulator, transient simulator, LOCA and stability codes as [

The steady-state core simulator, MICROBURN-B2, explicitly models the assembly specific flow paths through the lower tie-plate flow holes and the channel seals in addition to a [

] through the core support plate. The numerical solution for the individual flow paths is computed based on a general parallel channel hydraulic solution that imposes a constant pressure drop across the core fuel assemblies and the bypass region. This solution scheme incorporates [

] that is dependent on the [

I].

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-1 0 The MICROBURN-B2 state-point specific solution for bypass flow rate and [

is then used as initial conditions in the transient and LOCA analyses. When the reactor operates on high rod-lines at low flow conditions, the in-channel pressure drop decreases to a point where a solid column of water cannot be supported in the bypass region, and voiding occurs in the core bypass. For these conditions (in the region of core stability concerns) the neutronic feedback of bypass voiding [

I The level of bypass boiling for a given state-point is a direct result of the hydraulic solution. The potential for boiling increases as the power/flow ratio increases or the inlet sub-cooling decreases. While the licensing methodology utilizes a [

] to estimate the potential for localized bypass boiling. This [

] to specifically determine a bounding local void distribution in the core. The model is conservative in that it [

]. The capability of this model to predict localized bypass boiling is demonstrated in Figure C-27 for a hypothetical case where the inlet sub-cooling was artificially decreased to induce bypass boiling.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-11 Bypass voiding is of greatest concern for stability analysis due to its direct impact on the fuel channel flow rates and the axial power distributions. The reduced density head in the core bypass due to boiling results in a higher bypass flow rate and consequently a lower hot channel flow rate. This lower hot channel flow rate and a more bottom-peaked power distribution (due to lower reactivity in the top of the core due to boiling in the bypass region) destabilize the core through higher channel decay ratios. AREVA stability methods directly model these phenomena to assure that the core stability is accurately predicted.

CASMO-4 has the capability to specify the density of the moderator in the bypass and in-channel water rods, [

Bypass voiding is not encountered during full power, steady-state EPU operation for Brunswick so there is no impact on steady-state analyses. For transient conditions it is conservative to ignore the density changes as additional voiding aids in shutting down the power generation.

For Brunswick, a 100% power/ 100% flow statepoint (120% of the original licensed thermal power) was assessed. Even with the conservative multi-channel model, there was no localized bypass boiling at the EPU power level. This assessment assures that the limiting transients at the uprated thermal power are not adversely affected by bypass boiling. As the flow is reduced along the 100% power line, the decrease in flow is compensated by increased sub-cooling which more than compensates for the decrease in flow. When flow is further reduced along the highest rod line, boiling in the bypass is calculated to begin. This is in the area of stability concerns where the boiling in the bypass is modeled explicitly. For normal operation at 100%

power boiling in the bypass is not expected to occur, so there is no impact on the lattice local peaking or the LPRM response.

C.3.2 Fuel Assembly Design No fuel design modifications have been made for EPU operation, neither mechanical nor thermal hydraulic. The maximum allowed enrichment level of any fuel pellet is 4.95 wt% U-235.

A description of fuel enrichments on both a lattice basis and an assembly basis for the first reload of ATRIUM-10 fuel in Brunswick is provided in Table C-3.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page C-12 All new and spent fuel at Brunswick is stored in the Spent Fuel Storage Pool (SFSP) and in accordance with Technical Specification 4.3.1.1 must maintain a subcritical multiplication factor (keff) of less than 0.95 when flooded with non-borated water. A SFSP criticality analysis has been performed for Brunswick that confirms that this requirement is met for ATRIUM-1 0 fuel designs. This analysis applies to both units which have the same high density storage rack configuration, as detailed in section 9.3 of the FSAR. A reload specific evaluation is performed to verify that the specific bundle designs being loaded remain bounded by the criticality analysis.

C.4 References C.1 XN-NF-85-67(P)(A) Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, September 1986.

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Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-13 Table C-1 KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A)

I Table C-2 Comparison of CASMO-4 and MCNP results for ATRIUM-10 Design

[

I AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-14 Table C-3 Fuel Enrichment Description for the Initial Brunswick EPU ATRIUM -10 Fuel Cycle Design

[

NOTES:

1.

If this limit is not met, a CASMO in-rack k-inf calculation can be performed. If k-inf is less than the base analysis, then fuel may be stored.

I AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Pape C-15 Al 0B-4340L-15070 U235 Thermal Absorption 2ro T~ 250 E

2 U

230 0 VI I00 VH

""W-04 V1100 VH

', 08Vll00VH 00VI/04 VH

-.- 04 VII 04 VH -0 8Vi/04 VH 00 V11

/ 8VH

-04VI/08VH 0.8 V110.8 VtI 2101 2OO 190 180 0

10 20 30 40 50 60 70 80 Exposure (GWdMTU)

Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches Figure C-1 A1OB-4340L-15G70 U235 Fast Absorption Z

0 o Vii0 0 VH 04 VI /0 0 VH I 08VI 00 VHi

--*--0 0 Vii 04 VH

-*--0 4 VI 0 4 VH -08 Vi10 4 VH

-,0 0 V1 10 8Vt

-04VII 08V 0.8 Vi 10.8 ViH 10 20 30 40 50 60 70 80 Exposure (GWdIMTU)

Figure C-2 Microscopic Fast Cross Section of U-235 from Base Depletion and Branches AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power UDrate Conditions ANP-2638NP Revision 0 Page C-16 BOL A1OB-4340L-15G70 U235 Thermal Cross Sections 210.

Absorption 190 180 170 160 I

Quadratic Fit

CASMO-4 Data SCASMO-4 Data 0uadraSio F~it

'--Fission F

.s.

0 10 20 30 40 50 60 Void Fraction (%)

70 80 90 100 Figure C-3 Microscopic Thermal Cross Section of U-235 at Beginning of Life BOL AIOB-4340L-1 5G70 U235 Fast Cross Sections 1320 120

'Absorption 11.0 100 2

go a"

.A-Fission 7.0

CASMO-4 Data Quadratc Fit A CASM0-4 Data

-Quadratic Fit 0

10 20 30 40 50 60 70 80 90 100 Void Fraction (%)

Figure C-4 Microscopic Fast Cross Section of U-235 at Beginning of Life AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paqe C-17 BOL A10B-4245L-14G70 U-235 Microscopic Cross Sections (Thermal) 6 S

200 195 190 185 180.

175 170D 165-160 155 1.

150-0 01 02 0.3 0.4 05 06 07 0.8 09 Void Fraction S ig-A2 (CASMO-4)

Quadratic Fit (0.40.80) 9 Sig-F2 (CASMO-4)

O Quadratic Fit (0.40,80)

Figure C-5 Microscopic Thermal Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions BOL A1OB-4245L-14G70 U-235 Microscopic Cross Sections (Fast)

.0 2

U a

2 I

13 U 1 1.0 100 90.

80 70

Sig-AI (CASMO-4)

Quadratic Fit (0.40,80)

Sig-Fl (CASMO-4)

Quadratic Fit (0,40.80) 0 01 02 03 04 05 06 07 08 09 Void Fraction I

Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paqe C-18 BOL A1OB-4254L.14G70 Macroscopic Diffusion Coefficients 25000

- ¶.0000 0 5D 00000 9-D-1 (CASMO-4)

-0usdra-ic (0-40-80) n D-2 (CASMO-4)

Ouaddakc (0-40-80) 0 0.1 02 03 04 05 06 0.7 08 0.9 Vold Fraction I

Figure C-7 Macroscopic Diffusion Coefficient (Fast and Thermal)

Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-19 Cross Section (barns)

Instantaneous Void Fraction (%)

20 10 0*,. 2 :

0..o 0U C.O

_L, *oz

.5 Iistorical Void Fraction U

q0-

"C "

>o 0

Figure C-8 Microscopic Thermal Cross Section of U-235 at 70 GWdlMTU AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-20 AlOB-4340L-15G70 U235 Thermal Absorption at 70 GWd/MTU

00 Inst Void (CASMO-4) 0 00 Inst Void (quad)

04 InstVoid (CASMO-4) 04 Inst Void (quad)

A 0 8 Inst Void (CASMO-4

-08 Inst Void (quad) -Lookup at 0.60 ViH 0

10 20 30 40 50 60 70 80 90 100 Historical Void Fraction Figure C-9 Quadratic Interpolation Illustration of Microscopic Thermal Cross Section of U-235 AlOB-4340L-15G70 U235 Thermal Absorption at 70 GWdIMTU A Inta*',ccti points

-0 6 Void History (quad) 0 7 Inst Void 0 - Resultant Value 0

10 20 30 40 50 60 70 80 90 100 Instantaneous Void Fraction Figure C-10 Illustration of Final Quadratic Interpolation for Microscopic Thermal Cross Section of U-235 AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-21 I

Figure C-ll Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9

[

I Figure C-12 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMQ-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-22

[

I Figure C-13

[

Delta k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at 0.4 Historical Void Fraction and 0.9 Instantaneous Void Fraction I

Figure C-14 Comparison of Interpolation Process Using Void Fractions of 0.0, 0.4 and 0.8 and Void Fractions of 0.0, 0.45 and 0.9 AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Panqe C-23 I

Figure C-15 I

I EMF-2158(P)(A) TIP Statistics by Axial Level I

Figure 0-16 Quad Cities Unit I Pin by Pin Gamma Scan Results AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-24

[

I Figure C-17 I

Maximum Assembly Power in Topical Report EMF-2158(P)(A)

I Figure C-18 Maximum Exit Void Fraction in Topical Report EMF-2158(P)(A)

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Paae C-25 I

Figure C-19 Maximum Assembly Power Observed from Recent Operating Experience I

I Figure C-20 Void Fractions Observed from Recent Operating Experience AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power U)rate Conditions ANP-2638NP Revision 0 Paqe C-26 I

I Figure C-21 Axial Power and Void Profile Observed from Recent Design Experience

[

Figure C-22 Nodal Void Fraction Histogram Observed from Recent Design Experience I

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uorate Conditions ANP-2638NP Revision 0 Pacie C-27 FiueC2 aiu Asml oe na EUBusikDsg I

Figure C-23 Maximum Assembly Power in an EPU Brunswick Design Figure C-24 Maximum Exit Void Fraction in an EPU Brunswick Design I

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-28 I

Figure C-25 Brunswick EPU Design Axial Profile of Power and Void Fraction

[

I Figure C-26 Brunswick EPU Design Nodal Void Fraction Histogram AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page C-29 EDIT OF VOID FRACTION IN BYPASS CHANNEL

[

IN UNITS OF %

I Figure C-27 Hypothetical MICROBURN-B2 Multi-Channel Average Bypass. Void Distribution AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page D-1 Appendix D Stability Methods D.1 DIVOM Analysis The DIVOM (_elta over Initial CPR Versus Oscillation Magnitude) for Option III plants was initially defined as a constant regional oscillation mode slope of 0.45 in Licensing Topical Report NEDO-32465-A. An earlier 10 CFR part 21 report (MFN-01-046 dated August 31, 2001) determined that this generic slope may not bound all current core and fuel designs. Closure of this part 21 was accomplished with the implementation of plant/cycle-specific DIVOM calculations as discussed in BWR Owners Group communications BWROG-03047 and -03048.

The basis of these calculations is the BWR Owner's Group guideline, GE-NE-0000-0031-6498-RO transmitted by OG04-0153-260 "Plant-Specific Regional Mode DIVOM Procedure Guideline" dated June 15, 2004. This guideline was later revised in OG05-0136-260 on June 2, 2005. All DIVOM calculations that were performed are consistent with the latest guideline.

It is important to understand that variations in DIVOM response on a cycle specific basis does not impact safety margin. The purpose of the Option III system is to protect the Technical Specification MCPR Safety Limit if a diverging oscillation event were to be experienced. A cycle specific setpoint calculation is performed to ensure that cycle specific variations in parameters that have the potential to impact the protection of the MCPR Safety Limit are explicitly included.

The part 21 documented in MFN-01-046 and its associated closure essentially redefined DIVOM as one of these parameters that need to be calculated on a cycle-specific basis.

An increase in DIVOM slope value for a specific operating cycle does not mean that the core is less stable or more likely to experience an oscillation. A larger DIVOM slope value does mean that the MCPR response for a specified oscillation magnitude is larger. However, since the larger cycle-specific DIVOM value is explicitly included in the setpoint calculation, no degradation in the margin of safety occurs.

D.2 ATWS/Instability The industry assessments of ATWS rules relative to BWR core thermal-hydraulic stability presented in NEDO-32047-A (Reference D.1) and the mitigation of BWR core thermal-hydraulic AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page D-2 instabilities in ATWS (NEDO-32164) were performed for the BWROG using GNF proprietary codes and methods. These assessments provide strong evidence that the proposed plant operating guideline change to mitigate the instabilities is effective and reduce the consequences of the event. The NRC recognized in its evaluation of (Reference D.2) that these calculations exhibited a high degree of uncertainty, but considered that even if the calculated power oscillation amplitude were in error by an order of magnitude, the conclusions would remain applicable. Therefore, the concern with respect to the core power distributions for EPU cycles loaded with ATRIUM-10 fuel must be assessed relative to this licensing conclusion (within a factor of 10). Since the dynamics of large amplitude limit cycles are only weakly fuel type dependent, due to the enrichment limits and hydraulic compatibility, the analysis of instabilities in the presence of ATWS conditions has been considered a plant analysis. This view is reinforced by the fact that the only effective mitigating action was a plant operating procedure and not a fuel design or cycle design change. Regardless, a brief discussion of the ATWS/Instability issue has been prepared based on a first principles understanding of the instability phenomena, to provide clarification of the phenomena associated with the underlying instability.

D.2.1 Background The limiting ATWS transient is a turbine trip with 100% bypass which brings the core to natural circulation at a high rod line accompanied by high inlet subcooling. Calculations by GE using TRACG code (Reference D.1) indicate that oscillations can grow to large amplitudes and the global mode is more limiting than the regional mode oscillations. Short prompt-critical pulses were also observed resulting in irregular oscillation peak sequences. The origin of the irregular nature of the peaks was not understood at the time of the review and was attributed to numerical sensitivities leading to the NRC accepting the results with large uncertainties or rather on a qualitative basis. In using this original approved analysis as a discussion basis, it is recognized that the following are the important phenomena:

Maximum amplitude of the oscillation (limit cycle)

Average power increase due to the oscillation Limiting mode Role of the prompt-critical pulses AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paqe D-3 Role of void-reactivity coefficient Based on the phenomenological discussion of the above effects, the consequences of EPU operation and ATRIUM-10 fuel loading will be put in perspective regarding the potential for fuel damage as a consequence of the limiting ATWS transient with instability.

D.2.2 Review of the Important Phenomena D.2.2.1-Global Oscillation Limit Cycle Amplitude

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I AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paae D-4 I

I D.2.2.2-Average Power Increase due to Oscillations AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power UDrate Conditions ANP-2638NP Revision 0 Paae D-5 I

I D.2.2.3-Limiting Mode D.2.2.4-The Role of the Prompt-Critical Pulses I

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Paae D-6 I

D.2.2.5-The Role of Void-Reactivity Coefficient

[

D.2.3 Assessing the Change in Fuel Damage Potential The assessment of the fuel damage potential is made with regard to variations related to the ATRIUM-10 fuel type and EPU.

AREVA NP Inc.

Applicability of AREVA NP BWR ANP-2638NP Methods to Extended Power Revision 0 Uprate Conditions Page D-7 D.2.3.1-Effect of the ATRIUM-10 Fuel Type The fuel types manufactured by different vendors vary in their respective mechanical design and geometry. However, the variations in the essential operational characteristics such as pressure drop and void-reactivity coefficient are relatively small as necessitated by the need for compatibility in mixed core situations. These variations are not sufficiently large as to affect the nature of a qualitative analysis such as ATWS with instability. No other significant ATWS/Instability phenomena can be related directly to fuel properties.

D.2.3.2-Effect of EPU Core Loading

[

AREVA NP Inc.

Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions ANP-2638NP Revision 0 Page D-8

[

I D.3 References D.1. NEDO-32047-A, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability,"

GE Nuclear Energy, June 1995.

D.2. W. Wolf et al., "BWR Stability Analysis with the BNL Engineering Plant Analyzer,"

NUREG/CR-5816, October 1992.

D.3. Y. M. Farawila and D. W. Pruitt, "A Study of Nonlinear Oscillations and Limit Cycles in Boiling Water Reactors - I: The Global Mode," Nuclear Science and Engineering: 154, 302-315 (2006)

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v t e Letter Sequence Other
TAC:MD4061, (Approved, Closed) TAC:MD4062, (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Questions Answers Results Approval, Approval Other: BSEP 07-0053, Areva Np, Inc., Topical Report, ANP-2637, Revision 1, Boiling Water Reactor Licensing Methodology Compendium, BSEP 07-0067, BSEP 07-0067 Enclosure 9; Areva Report ANP-2624(NP), Revision 0, Brunswick, Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM-1O Fuel, Dated June 2007, BSEP 07-0075, Areva Report ANP-2638NP, Revision 0, Applicability of Areva Np BWR Methods to Extended Power Uprate Conditions, Enclosure 6, BSEP 07-0102, ANP-2661 (Np), Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUM-10 Fuel, ML072070311, ML072070318, ML072180372, ML072340562, ML072570210, ML072840219
MONTHYEAR2007-07-18 [Table View]

BSEP 07-0075, Areva Report ANP-2638NP, Revision 0, Applicability of Areva Np BWR Methods to Extended Power Uprate Conditions, Enclosure 6

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