ML032890044

Stations - Partial Response to RAI Regarding the Use of Mixed Oxide Lead Fuel Assemblies

ML032890044
Person / Time
Site:	Mcguire, Catawba, McGuire, 07003098
Issue date:	10/03/2003
From:	Tuckman M Duke Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-RFPFR
Download: ML032890044 (45)

Text

_ Duke MICHAEL S. TUCKMAN OrPowere Executive Vice President A Duke Energy Company Nuclear Generation Duke Power P.O. Box 1006 Charlotte, NC 28201-1006 Mailing Address:

526 South Church Street Charlotte, NC 28202 704 382 2200 October 3, 2003 704 382 4360 fax U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Catawba Nuclear Station Units 1 and 2; Docket Nos. 50-413, 50-414 McGuire Nuclear Station Units 1 and 2; Docket Nos. 50-369, 50-370 Partial Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies

Reference:

NRC Letter dated July 25, 2003, Request for Additional Information Re: Mixed Oxide Lead Fuel Assemblies (TAC Nos. MB7863, 7864, 7865, 7866)

Please find attached a partial response to the Nuclear Regulatory Commission Staffs Request for Additional Information (RAI) transmitted by the referenced letter.

By letter dated February 27, 2003 Duke Energy submitted an application to amend the licenses of McGuire and Catawba to allow the use of four mixed oxide fuel lead assemblies. In a subsequent submittal dated September 23, 2003, Duke notified the Nuclear Regulatory" Commission of the decision to focus the lead assembly program on Catawba only. As part of the review of this application the Nuclear Regulatory Commission staff in a letter dated July 25, 2003 requested that Duke provide additional information related to the application. To facilitate the NRC Staff's review, Duke is providing responses to 23 of the 47 Reactor Systems questions contained in the RAI at this time. The remaining responses will be submitted by the committed date of October 31, 2003.

Inquiries on this matter should be directed to G. A. Copp at (704) 373-5620.

Very truly yours, 1.S.W M. S. Tuckman Attachment f\ 0 (l www. duke-energy. corn

-t U.S. Nuclear Regulatory Commission October 3, 2003 Page 2 Oath and Affirmation M. S. Tuckman affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

M.S. Tuckman Subscribed and sworn to before me on this g day of Oc f , 2.A oo3 mug C Notary Public My Commission expires: MICHAEL T. CASH Notary Public Lincoln County, North Carolina Commission Expires January 22, 2008 gun 7';* 2-o Date Seal ..

T U.S. Nuclear Regulatory Commission October 3, 2003 Page 3 cc: with Attachment R. E. Martin, NRC Senior Project Manager (08G9)

U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2738 L. A. Reyes, Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 J. B. Brady NRC Sr. Resident Inspector McGuire Nuclear Station E. F. Guthrie NRC Sr. Resident Inspector Catawba Nuclear Station

I U.S. Nuclear Regulatory Commission October 3, 2003 Page 4 bcc: with Attachment P. M. Abraham - ECO8I D. E. Bortz- EC08G L. F. Vaughn - PBO5E J. L. Eller - EC09A S. P. Nesbit - EC09A KS. Canady- EC08H M.S. Scott - EC08H R.M. Gribble - EC08F C. J. Thomas - MGOlRC M.T. Cash - ECO50 G. D. Gilbert - CNOlRC R. H. Clark - Duke Cogema Stone & Webster A. W. Cottingham - Winston & Strawn G. A. Meyer - Framatome ANP P. T. Rhoads - Department of Energy D. J. Spellman - Oak Ridge National Laboratory NRIA File/ELL - ECO50 McGuire Master File - MGO1DM Catawba Master File 801.01 - CN04DM Catawba RGC Date File (J.M Ferguson - CNOlSA)

MOX File - 1607.3202

Attachment Partial Response to NRC Staff Request for Additional Information dated July 25, 2003 Reactor Systems

1. With respect to section 3.1, provide a full description of the post-irradiation examination planned to verify the mechanical properties of the lead test assemblies (LTAs) following irradiation. Describe the test methods planned and the acceptance criteria for each test as well as the frequency of each test. Additionally, please provide the same information for the hot cell examinations that will be performed following irradiation.

Response

Post-irradiation examination (PIE) of the lead assemblies will include both poolside and hot cell examinations. A description of the examinations, including test methods, acceptance criteria, and frequency of testing, is found in Section 8.5 of Reference Q1-1. This topical report is currently under NRC review.

The purpose of the hot-cell examinations is to collect data to confirm the applicability of fuel performance models for fuel burnups greater than the limit (50,000 MWd/MThm maximum rod average) that is proposed for batch application. Therefore, specific acceptance criteria have not been set for the hot-cell examinations.

Reference Q1-1. BAW-10238(NP) Revision 1,MOX Fuel Design Report, Framatome ANP, May 2003.

3. Provide the specific bumup limit that is being requested for the LTAs.

Response

The requested bumup limit for the lead assemblies is 60,000 MWd/MThm for the maximum rod (axially averaged). This is greater than the 50,000 MWdIMThm maximum rod average that is proposed for batch application. Additional information on these limits is contained in Section 8.4 of Reference Q3-1.

Reference Q3-1. BAW-10238(NP) Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.

4. Section 3.5.1.2 of the submittal states that the reduction in total plutonium concentration ensures that the macroscopic plutonium effects on fuel performance are bounded. Please define the subject macroscopic plutonium effects.

Response

The macroscopic effects referred to are changes in the following parameters:

• Fuel thermal conductivity

• Pellet thermal expansion

• Pellet thermal creep

• Fission gas release

• Fuel densification and swelling

• Helium gas accumulation and release

• Pellet radial power profile

• Fuel melting point Each of these parameters is affected by the amount of PuO 2 in the MOX fuel pellets. By using approximately 40% less PuO 2 in weapons grade MOX fuel relative to reactor grade MOX fuel, the effects of changes to these fuel performance parameters relative to low enriched uranium (LEU) fuel will be bounded by the European experience with reactor grade MOX fuel.

These effects are also listed in Section 2.1 of Reference Q4-1, the MOX Fuel Design Report, which has been submitted to the NRC for review. Further discussion of these macroscopic effects is provided in Sections 2 and 3 of that document. The pellet radial power profile will be addressed further in the response to Reactor Systems RAI Question 7.

Reference Q4-1. BAW-10238(P), Revision 1,MOX Fuel Design Report, Framatome ANP, May 2003.

5. Provide the appropriate regulatory criteria used for the parameters discussed in section 3.5.1.

Response

Section 3.5.1 provides a description of MOX fuel and fuel rod design features, and is consistent with Reference Q5-1, which is currently under NRC review. Similarly, Section 3.5.2 describes the MOX fuel assembly mechanical design features. These sections provide the necessary background for the design evaluation, which is presented in Section 3.5.3 and discusses how the design satisfies the design criteria. The parameters described in Sections 3.5.1 and 3.5.2 support a fuel design that satisfies the acceptance criteria in Section 4.2 of Reference Q5-2 that pertain to pressurized water reactor fuel assembly design.

References Q5-1. BAW-10238(P), Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.

Q5-2. NUREG-0800, U. S. Nuclear Regulatory Commission StandardReview Plan, Revision 2, July 1981.

6. Provide a statistical analysis showing that the distribution of fissile material for the weapons-grade (WG) mixed oxide (MOX) fuel is the same as the distribution of fissile material for reactor-grade (RG) MOX fuel.

Response

The requested statistical analysis is not practical until after the fuel is manufactured.

However, it is not necessary to perform such an analysis in order to show that the distribution 2

of fissile material is approximately the same for both types of MOX fuels when using the same manufacturing process.

As discussed in Section 2.3 of Reference Q6-1, the MIvAS process ensures homogeneity of the pellet. The plutonium-rich agglomerates are finely dispersed in the U0 2 matrix, and this distribution is not a function of isotopic content, only the final desired plutonium content.

With regard to plutonium-rich agglomerate size, the relevant portions of the European (Reactor Grade) and American (Weapons Grade) Fuel Pellet Specifications are identical.

This is also discussed in Section 2.3. Therefore, the distribution of plutonium-rich agglomerates is the same in reactor-grade and weapons-grade fuels.

As discussed in Section 3.3 of Reference Q6-1, the fissile plutonium contents of the agglomerates in RG and WG MOX fuel are approximately 22.5% and 19.2%, respectively.

The fissile uranium content of the agglomerates is very small and is approximately 0.2% for both RG and WG MOX fuel, assuming a 0.25 weight percent tails uranium is used for feed powder.

Since the distribution of plutonium-rich agglomerates is the same in RG and WG MOX fuels, and the fissile plutonium content of the agglomerates is approximately equal, the distribution of fissile material will also be approximately the same for both types of MOX fuel of equivalent reactivity. However, note that it is common practice to adjust the master mix to U0 2 ratio in order to achieve the desired final plutonium content of a given rod. Each MOX fuel assembly is composed of three different plutonium content rods. For example, the three plutonium concentrations could be 2.40, 3.35, and 4.94 w/o plutonium for corner, periphery, and interior pins, respectively (see response to Reactor Systems RAI Question 18, Figure 18-2). The adjustment for these different pins, for which there is plenty of experience, dramatically changes the fissile material distribution within the pellet and far outweighs the small difference between RG and WG MOX fuel.

Reference Q6-1. BAW-10238(P), Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.

8. In section 3.5.3, the statement is made that the Mark-BW/MOX1 fuel assembly design meets all applicable criteria to maintain safe plant operation. Specify all of the regulatory criteria being referred to in this statement.

Response

A detailed analysis of the Mark-BW/MOX1 fuel assembly design is presented in Reference Q8-1. The design satisfies the acceptance criteria in Section 4.2 of Reference Q8-2 that pertain to pressurized water reactor fuel assembly design. The acceptance criteria in Reference Q8-2, in turn, align with the relevant requirements (regulatory criteria) in 10 CFR 50.46, General Design Criteria 10, 27, and 35, and 10 CFR 100.

The criteria in Reference Q8-2, Section 4.2.ll.A, that are not addressed are as follows:

I.A.1(h): This criterion applies to control rods rather than fuel.

II.A.2(c): Fretting is treated in the response to criteria in paragraph II.A.I(c).

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II.A.2(f): This criterion applies only to boiling water reactor fuel.

II.A.2(i): Mechanical fracturing is treated in the response to criteria in paragraph II.A.3(e).

ll.A.3(c): More stringent criteria are already applied in paragraph ll.A.3(a).

Reference Q8-1. BAW-10238(NP) Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.

Q8-2. NUREG-0800, U. S. Nuclear Regulatory Commission StandardReview Plan, Revision 2, July 1981.

9. Section 3.5.3 refers to fuel rod analyses which follow previously approved methods. Please state the methods being referred to.

Response

Approved methods used for analysis of MOX fuel are listed below. A more detailed discussion of the methods used for analysis of MOX fuel is provided in Reference Q9-1.

The fuel rod cladding stress and cladding fatigue were analyzed using approved methodologies for M5S'cladding that are described in Reference Q9-2. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.

Cladding corrosion was analyzed using the COPERNIC code with models approved for predicting M5T'cladding oxide thickness as described in Reference Q9-2. This methodology is applicable to both LEU and MOX fuel because it is independent of the pellet type.

Axial growth calculations for fuel rods and fuel assemblies also used the methodology that was approved in Reference Q9-2.

The hydraulic forces for assembly liftoff calculations were determined using the NRC approved LYNXT code described in Reference Q9-3. This methodology is applicable to both LEU and MOX fuel because it is independent of the pellet type.

Fuel rod end-of-life internal pressure and centerline fuel melt temperatures were evaluated according to the methods and criteria presented in Reference Q9-4 for MOX fuel rods.

Approval of Reference Q9-4 is pending. Revision 0 of this topical report has been approved for use with LEU fuel. A consistent methodology is used to describe the variations in properties between LEU and MOX fuel.

The fuel rod was analyzed for creep collapse using approved methods described in Reference Q9-5. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.

Overheating of cladding was analyzed with the approved BWU-N and BWU-Z critical heat flux correlations, which are described in References Q9-6 and Q9-7. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.

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Fuel enthalpy during a rapid insertion of reactivity (control rod ejection) is calculated using the SIMULATE-3K MOX computer code as described in Section 3.7.2.4 of Attachment 3 to the Duke Power MOX fuel lead assembly license amendment request dated February 27, 2003. SIMULATE-3K MOX is not yet an approved methodology for MOX fuel, although the similar code SIMULATE-3K has been approved for application to rod ejection analyses in low enriched uranium fuel cores (Reference Q9-8).

LOCA calculations for cladding rupture, cladding embrittlement, and fuel rod ballooning used approved methods described Reference Q9-2. These methods are applicable to both LEU and MOX fuel because they are independent of the pellet type.

The axial and horizontal faulted analysis methodologies (for external forces) are consistent with the approved methodologies described in References Q9-9 and Q9-10, respectively.

These methods are applicable to both LEU and MOX fuel because they are independent of the pellet type. Application of these methods to the Advanced Mark-BW fuel assembly design is discussed Reference Q9-11.

References Q9-1. BAW- 10238(NP) Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.

Q9-2. BAW-10227P-A Revision 0, Evaluationof Advanced Cladding and Structural Material(M5r') in PWR Reactor Fuel, February 2000.

Q9-3. BAW-10156-A Revision 1, LYNXT: Core Transient Thermal-HydraulicProgram, August 1993.

Q9-4. BAW-10231P Revision 2, COPERNIC Fuel Rod Design Computer Code, July 2000.

Q9-5. BAW-10084-A Revision 3, Programto Determine In-ReactorPerformanceof BWFC Fuel Cladding Creep Collapse, July 1995.

Q9-6. BAW-10199P-A Revision 0, The BWU CriticalHeat Flux Correlations,December 1994.

Q9-7. BAW-10199P-A Addendum 2, Application of the BWU-Z CHF Correlationto the Mark-BW1 7 FuelDesign with Mid-Span Mixing Grids, June 2002.

Q9-8. DPC-NE-2009-P-A, Duke Power Company Westinghouse Fuel Transition Topical Report, September 1999.

Q9-9. BAW-10133P-A Revision 0, Mark-C FuelAssembly LOCA-Seismic Analysis, June 1986.

Q9-10. BAW-10133P-A Revision 1, Addendum 1 and Addendum 2, Mark-C FuelAssembly LOCA-Seismic Analyses, October 2000.

Q9-1 1. BAW-10239(NP) Revision 0, Advanced Mark-BW FuelAssembly Mechanical Design Topical Report, March 2002.

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10. Section 3.5.3 states that the Mark-BW/MOX1 design preserves the original plant licensing bases for all reactor internal components. Specify the components and the bases for them that are being referred to in this statement. Also explain how the MARK-BW/MOX1 design preserves them.

Response

The reactor internal components that are affected by the fuel assembly design are the upper and lower core plates. The licensing basis is preserved through ensuring that the criteria established in Reference Q10-1 are met. These criteria were developed from the acceptance criteria of Reference Q10-2. The axial growth and assembly liftoff criteria in Reference Q10-1 must be met during the assembly lifetime. In addition, the fuel assembly must meet the following criteria for externally applied forces:

a) Operating basis earthquake: Allow continued safe operation of the fuel assembly following an event by ensuring the fuel assembly components do not violate their dimensional requirements.

b) Safe shutdown earthquake: Ensure safe shutdown of the reactor by maintaining the overall structural integrity of the fuel assemblies, control rod insertability, and a coolable geometry within the deformation limits consistent with the emergency core cooling system and safety analysis.

c) LOCA or safe shutdown earthquake plus LOCA: Ensure safe shutdown of the reactor by maintaining the overall structural integrity of the fuel assemblies and a coolable geometry within deformation limits consistent with the emergency core cooling system and safety analysis.

The Mark-BW/MOX1 design preserves the design basis because the growth models, nominal dimensions, and tolerances for the holddown springs, nozzles, and guide thimble assemblies are the same as those given in Reference Q10-3.

References Q1O-1. BAW-10238(NP) Revision 1,MOX Fuel Design Report, Framatome ANP, May 2003.

Q10-2. NUREG-0800, U. S. Nuclear Regulatory Commission StandardReview Plan, Revision 2, July 1981.

Q10-3. BAW-10239(NP) Revision 0, Advanced Mark-BWFuelAssembly Mechanical Design Topical Report, March 2002.

11. Please respond to the following items related to section 3.6.1, so that the Nuclear Regulatory Commission (NRC) staff may evaluate the conclusion that the vessel fluence increase is limited:

(A) Identify the location of the LTAs.

(B) Identify the existing peak fluence azimuthal location of the vessel.

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Response (A)

The MOX fuel lead assembly license amendment request is not tied to a specific unit and cycle. Current plans are to insert the lead assemblies in Catawba Unit 1 Cycle 16 (ClC16),

which will start up in late spring 2005. The C1C16 core design will not be finalized until the first quarter 2004. The location of the lead assemblies will be influenced by numerous design considerations, and it is not feasible to commit to the location at this time. Duke has made a commitment (Reference Ql 1-1)to the Nuclear Regulatory Commission to place at least two of the four lead assemblies in core locations that contain incore flux mapping instrumentation. Core location C-08 (and its symmetrical locations) is a potentially desirable location because it is the only core location that will (i) allow full instrumentation for all four assemblies and (ii) maintain an eighth core symmetric arrangement. Peripheral core locations would not be desirable because the lead assemblies would not be subjected to sufficient fuel duty to achieve burnups representative of MOX fuel batch implementation.

Duke has also made a commitment (Reference Q I1-1) not to place the lead assembly in a core location with a control rod during the first cycle of use.

Figure QI 1-1 depicts the Catawba one-quarter core geometry. The locations shaded in gray and red show typical fresh fuel placement in recent Catawba fuel cycles. The desirable lead assembly location C-08 (and its symmetrical location H-13) is shown in red.

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Figure Q1 1-1 Representative MOX Fuel Lead Assembly Core Design H G F E D C B A 8r 9

10 11 12 13 14 15 I I I I Response (B)

For Catawba Units 1 and 2, the peak fast neutron fluence at the pressure vessel cladding/base metal interface is at the 300 azimuth relative to the core cardinal axis.

Catawba cores are designed with eighth-core symmetry. Accordingly, relative to the diagram shown in Figure Q I1-1, the core cardinal axis would be through row eight.

Figures Q1 1-2 and Q1 1-3 show the relative fast neutron flux differences between a core with four MOX fuel lead assemblies and an all LEU core at beginning of cycle and end of cycle, respectively (This information was not specifically requested but it shows the negligible impact of four MOX fuel lead assemblies on core exterior fast flux, which control vessel fluence).

Reference Q11-1. Tuckman, M.S., June 26, 2003, Letter to U.S. Nuclear Regulatory Commission, Physics Testing Program in Support of Topical Report DPC-NE- 1005P Nuclear Design Methodology Using CASMO-4/ SIMULA TE-3 MOX, August 2001.

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Figure Q1 1-2 Fast Neutron Flux Comparison All LEU Core vs Core with Four MOX Fuel Assemblies (Beginning of Cycle)

Flux above 0.625 Mev HFP, 4 EFPD, ARO, Nominal Conditions H G F E D C B A 3.619 3.561 3.652 3.654 3.688 3.709 2.970 1.429 8 3.607 3.551 3.649 3.670 3.761 3.948 3.050 1.452 40.3 -0.3 -0.1 0.4 2.0 6.4 2.7 1.6

~ ~ 3.565

_~ 3.597 ~~.2

_1.629___ 3_0 ___ _

3.626 3.604 3.442 3.121 1.629 9 3.587 3.560 3.635 3.644 3.515 3.170 1.646

-0.3 a - -0.1 a - 0.2 a - 1.1 a - 2.1 a 1.6 a - 1.0 1.0 3.635 0.2 1.1 3.644 - 1.646 3.587 3.555 3.588 3.582 3.260 1.652 10 3.581 3.552 3.593 3.594 3.271 1.657

-0.2

_ -0.1 0.1 -_0.3 0.3 0.3 3.584 3.680 3.664 3.055 1.220 11 3.571 3.664 3.649 3.044 1.217

-OA -0.4 -0.4 -0.4

_ -0.2 3.975 3.678 2.240 12 3.947 3.650 2.223

-0.7 -0.8 -0.8 3.006 1.325 No MOX 13 2.979 1.312 4MOX

-0.9 -1 .0 %relative diff Figure Q1 1-3 Fast Neutron Flux Comparison All LEU Core vs Core with Four MOX Fuel Assemblies (End of Cycle)

Flux above 0.625 Mev HFP, 495 EFPD, ARO, Nominal Conditions H G F E D C B A 3.857 3.983 3.679 3.434 3.534 13.6671 3.072 11.651 a 3.872 3.997 3.692 3.438 3.514 3.647 3.047 1.648 0.4 0.4 0.4 0.1 -0.6 -0.5 -0.8 -0.2

~ ~ A&J A.1 A8ASA3.4 A^- 87_

1.AAA 3.811A~ 3.916 .4 3.644 3.8u4 3.540 3.410 1.876 9 3.825 3.929 3.652 3.798 3.520 3.4u00 1.877 0.4 0.3 0.2 402 40.6 40.3 0.1

- , - S - 0.2 S - -0.2 S - -0.6 S - -0.3 S - 0.1 3.827 3.968 3.782 3.909 3.548 1.872 10 3.841 3.981 3.790 3.916 3.555 1.878 0.4 0.3 0.2 0.2 0.2 0.3 3.830 3.958 3.605 3.052 1.345 11 3.844 3.973 3.619 3.065 1.352 0.4 0.4 0.4 0.4 0.5 3.790 3.524 2.196 12 3.807 3.541 2.208 0.4 0.5 0.5 2.872 1.356 No MOX 13 2.887 1.364 4 MOX 0.5 0.6 %relative dIff Note: Figures indicate fast neutron flux scaled by 1xi 0.14, in units of neutronslcm 2/sec 9

12. Provide the appropriate regulatory criteria to be satisfied by the information in section 3.7, i.e., how this section meets the general design criteria specified in the Standard Review Plan.

Response

Section 3.7 contains the safety analysis of three distinct subject areas; loss of coolant accidents (LOCA), non-LOCA accidents, and radiological consequences. The appropriate regulatory criteria for each of these topics are summarized in Tables Q12-1 through Q12-3.

LOCA Criteria The LOCA acceptance criteria of 10CFR 50.46 (b) were established for light water reactors fueled with U0 2 pellets within cylindrical Zircaloy cladding. The MOX fuel lead assemblies have M5'cladding and mixed oxide fuel pellets. The applicability of the 10CFR 50.46 criteria to the MOX fuel lead assemblies is established in Table Q12-1.

Non-LOCA Criteria The criteria used to evaluate the non-LOCA transients/accidents in the Updated Final Safety Analysis Report are summarized in Table Q12-2 and except for rod ejection accident criteria are the same criteria used for analysis of non-LOCA transients/accidents in LEU fuel cores.

Provisional Rod Ejection Accident Criteria The current acceptance criteria for a rod ejection accident (REA) at Catawba are described in Section 4.1.2 of Reference Q12-1. These criteria are based on Section 15.4.8 of Reference Q12-2 and are summarized below.

1. The radially averaged fuel pellet enthalpy shall not exceed 280 cal/gm at any location.

2. Doses must be "well within" the 10 CFR 100 dose limits of 25 rem whole-body and 300 rem to the thyroid, where "well within" is interpreted as less than 25% of those values.

3. The peak Reactor Coolant System pressure must be within Service Limit C as defined by the ASME Code, which is 3000 psia (120% of the 2500 psia design pressure).

With the exception of the enthalpy limit of 280 cal/gm, those criteria are equally valid for mixed oxide (MOX) fuel as for low enriched uranium (LEU) fuel during a REA. The dose acceptance criteria relate to the radiological consequences to the public, not the fuel type.

The primary system pressure acceptance criterion relates to the integrity of the pressure boundary, not the fuel type.

The enthalpy limit was established to ensure coolability of the core after a REA and to preclude the energetic dispersal of fuel particles into the coolant (Reference Q12-3). The current pressurized water reactor regulatory acceptance criterion of 280 cal/gm is based primarily on experiments such as SPERT that were conducted by the Atomic Energy Commission. More recent REA experiments conducted at the Cabri facility in France, among others, suggest that a lower enthalpy limit may be appropriate, particularly for high burnup irradiated fuel. The Electric Power Research Institute (EPRI) has used the more recent experimental data, coupled with cladding failure predictions using the Critical Strain Energy Density (CSED) approach, to develop proposed REA enthalpy limits as a function of 10

burnup. The work is documented in EPRI's Topical Report on Reactivity Initiated Accident:

Bases for RIA Fuel and Core Coolability Criteria" (Reference Q12-4), which as been submitted to the Nuclear Regulatory Commission (NRC) and is currently under review.

Four MOX fuel rods have been tested under simulated REA conditions as part of the Cabri test program. Of those tests, three experienced no cladding failure with peak enthalpies of 138, 203, and 90 cal/gm. However, the Rep Na-7 test saw a cladding failure with fuel dispersal at an enthalpy of 120 cal/gm. The Rep Na-7 rod had a burnup of 55 GWdIMThm and a cladding oxidation layer of 50 microns (Reference Q12-4, Table 2-1). Based on the results of that test, it has been postulated that differences in fuel pellet microstructure between MOX and LEU fuel may make MOX fuel more susceptible to disruptive cladding failure at lower fuel pellet enthalpy values.

Accordingly, for the MOX fuel lead assemblies, Duke proposes to use a radial average peak fuel enthalpy limit that is substantially more conservative than the current acceptance criterion for LEU fuel in Reference Q12-2. Duke proposes to use a value of 100 cal/gm at all burnups as the acceptance criterion for MOX fuel rods experiencing a power excursion from hot zero power (HZP). This criterion is considered to be appropriate and conservative, for the reasons provided below.

1. The value is significantly lower than enthalpies at which disruptive failure has been experienced in any MOX fuel REA tests.

2. The value is significantly lower than the Fuel Rod Failure Threshold curve for LEU fuel as proposed by EPRI (Reference Q12-4, Figure S-1).

3. MOX fuel rods will be clad in M51'. Fuel rod corrosion is considered to be a contributing factor to cladding failure under REA conditions. M5T has demonstrated extremely low corrosion relative to Zircaloy-4, the cladding material that was used in all MOX fuel REA tests (see Figure 6.1 of Reference Q12-5).

4. MOX fuel lead assembly rod burnup will be limited to less than 60 GWd/MThm.

5. Applying the criterion only to accidents from HZP excludes accidents initiating from hot full power with a high initial enthalpy (reflective of full power) but no rapid energy deposition in the fuel pellet.

Duke will use the SIMULATE-3K MOX computer code to perform three-dimensional reactor kinetics calculations of licensing basis REAs for all cores containing MOX fuel lead assemblies. Duke will verify that the peak enthalpy in all MOX fuel lead assembly rods remains below the 100 cal/gm acceptance criterion during postulated REAs. SIMULATE-3K MOX, described in Section 2.4 of Reference Q12-6, is an extension of SMULATE-3K.

Application of SIMULATE-3K for REAs at Catawba has been reviewed and approved by the NRC (Reference Q12-7) for cores containing LEU fuel. Analyses of representative cores containing MOX fuel lead assemblies are summarized in Section 3.7.2.4 of Reference Q12-8 and further detail will be provided in the response to Reactor Systems RAI Question 33.

The above criteria are conservative provisional criteria for the MOX fuel lead assembly program. To support the batch use of MOX fuel, Duke intends to propose alternative REA acceptance criteria. Duke plans to document the batch use MOX fuel REA acceptance 11

criteria and REA analytical methodology in a MOX fuel safety analysis topical report and submit the report to the NRC for review in 2004.

References Q12-1. DPC-NE-3001-PA, MultidimensionalReactor Transients and Safety Analysis Physics ParametersMethodology, Duke Power Company, December 2000.

Q12-2. NUREG-0800, U. S. Nuclear Regulatory Commission StandardReview Plan, Revision 2, July 1981.

Q12-3. Meyer, R. O., McCardell, R. K., Chung, H. M. Diamond, D. J. and Scott, H. H., A Regulatory Assessment of Test Datafor Reactivity-InsertionAccidents, Nuclear Safety, Volume 37, No. 4, October-December 1996.

Q12-4. EPRI Technical Report 1002865, Topical Report on Reactivity InitiatedAccident:

Basesfor RLA Fuel and Core Coolability Criteria,June 2002 (currently under NRC review).

Q12-5. BAW-10238(P), Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003 (currently under NRC review).

Q12-6. DPC-NE-1005P, Duke PowerNuclear Design Methodology Using CASM04ISIMULATE-3 MOX, August 2001 (currently under NRC review).

Q12-7. DPC-NE-2009-P-A, Revision 2, Duke Power Company Westinghouse Fuel TransitionReport, December 2002.

Q12-8. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.

Radiological Dose Criteria General radiological criteria are provided in 10CFR 20, 10CFR 50 Appendix A, 10CFR 50.67 and 10CFR 100. These are not published as uranium specific criteria, but have been consistently applied to reactor applications by the nuclear industry. Some of these regulations also apply to other applications, such as nuclear medicine. The applicable acceptance criteria in 10CFR are determined by the purpose or scenario for which the consequences must be calculated, rather than by the source term or specific isotopes involved.

The purpose of modeling the event and projecting consequences is to protect the health and safety of the public. To that end, there must be a standard for comparison to draw a definitive conclusion as to the impact upon the public. In order to compare the biological effects from the various isotopes which are produced in nuclear applications and industries, the concept of dose equivalent (or committed dose equivalent) was adopted.

Usually expressed in Rems or Sieverts, these units provide a comparison of biological effects by accounting for the energy deposition and the relative biological effectiveness from radiation emitted by isotopes.

Since dose is a measure of the cumulative biological effect of the emitted particles and rays regardless of the isotope of their origin, there is no need to specify specific dose acceptance criteria for a reactor using MOX fuel. Furthermore, the criteria which are 12

currently in regulations for the protection of the health and safety of the public and control room operators can be applied for the same purpose and application that they currently are being applied within a plant's licensing basis. The dose acceptance criteria in 10 CFR can be applied in the same manner as applied for LEU fuel. Standard Review Plan guidance can continue to be applied in accordance with a plant's licensing basis as it has been for LEU fuel. The specific regulatory dose criteria used to analyze MOX fuel events are summarized in Table Q12-3.

13

Table Q12-1 Applicability of 10CFR 50.46 Criteria to MOX Fuel Lead Assemblies

~ ~~~'AppIcadbliity tofMO Fuel Lead Assemhblies.-'-

50.46(b).

This criterion concerns the performance of the fuel pin cladding material during LOCA and is, therefore, primarily related to cladding properties. The MOX lead assembly fuel rods will be constructed using Framatome ANP's M5w cladding.

The 2200 OF criterion has been approved by the NRC as applicable to M5T cladding in granting the licensing of replacement fuel for several light water Peak Clad reactors over the last few years. The basis for approval is experimental evidence Temperature that M5'behavior during LOCA conditions isequivalent to or superior to Zircaloy

<2200 F and isdocumented in BAW-1 0227P-A, "Evaluation of Advanced Cladding and Structural Material (M5w) in PWR Reactor Fuel," February 2000."

This temperature criterion has no dependence on the fuel pellet design or makeup and is equally applicable for use with either U02 or MOX fuel pellets.

This criterion is fully applicable to the MOX fuel lead assemblies.

This criterion concerns the performance of the fuel pin cladding material during LOCA and is,therefore, primarily related to cladding properties. The MOX lead assembly fuel rods will be constructed using Framatome ANP's M5TM cladding, The 17 percent criterion has been approved by the NRC as applicable to M5 cladding in granting the licensing of replacement fuel for several light water reactors over the last few years. The basis for approval isexperimental evidence 17% Local that MTm behavior during LOCA conditions is equivalent to or superior to Zircaloy 17% Local Oxidation and is documented Structural in BAW-1 Material (M5'm) 0227P-A, in PWR "Evaluation Reactor Fuel," February 2000."Cladding and of Advanced The oxidation limit criterion controls the amount of hydrogen available to develop zirconium hydrides which increase the brittleness of the cladding in the post-accident environment. The criterion is not affected by the type of fuel pellet.

This criterion isfully applicable to the MOX fuel lead assemblies.

This criterion assures acceptable conditions within the reactor building and is 1%Core- unrelated to the core fuel and cladding so long as the hydrogen produced per wide percent cladding reacted Is unchanged. Because the reaction for both M5 and and oxygen, the hydrogen produced per reaction Oxidation Zircaloy is percent isthe same zirconium between for both materials. The criterion is unaffected by the use of M5' cladding and is fully applicable to the MOX fuel lead assemblies.

Core This criterion controls the geometry of the core following a LOCA. As a criterion, it Amenable to achieves its purpose regardless of the cladding material or the fuel pellet makeup.

Cooling It is fully applicable to the MOX fuel lead assemblies.

14

• tApplicability to MOX Fuel Lead Assemblies S- i '.

This criterion controls the availability of long-term cooling systems and core conditions. As a criterion, it achieves its purpose regardless of the cladding material or the fuel pellet makeup. It Isfully applicable to the MOX fuel lead assemblies.

15

Table 01 2-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies TrnslientlAccident -  ;-ett Criteria* . - X"7;jg

.Descriptlon -

62.1.3 LOCA Mass and Energy a Containment design margin is maintained.

Release and Containment

• Environmental qualification of the safety related equipment Inside Dressure/Temperature Response containment Is not compromised.

.2.1 A Secondary System Pipe . Containment design margin is maintained.

Ruptures and Containment

• Environmental qualification of the safety related equipment Inside Dressure/Temperature Response containment Is not compromised.

15.1.1 Feedwater System ialfunctions that Result in a

• Bounded by excessive increase In secondary steam flow analysis Reduction in Feedwater in Section 15.1.2 and same criteria apply.

Temperature

• Peak RCS pressure remains below 110% of the design limit 15.1.2 Feedwater System (<2750 psia)

Malfunction Causing an Increase

• Fuel cladding integrity shall be maintained by ensuring that the n Feedwater Flow calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110°%of the design limit 15.1.3 Excessive Increase in (<2750 psla) e5.1.3rEcSteam IFlow Fuel cladding Integrity shall be maintained by ensuring that the Secondary Steam Flow calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110% of the design limit 15.1.4 Inadvertent Opening of a (<2750 psia)

Steam Generator Relief or Safety . Fuel cladding Integrity shall be maintained by ensuring that the alve calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110% of the design limit

(<2750 psia)

• The potential for core damage Is evaluated on the basis that it is acceptable If the minimum DNBR remains above the 95/95 DNBR

15. 1.5 Steam System Pipin limit based on an acceptable DNBR correlation. If the DNBR falls Failure below these values, fuel failure must be assumed for all rods that do not meet these criteria. Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain In place and Intact with no loss of core cooling capability.

• Offsite doses calculated shall not exceed the guidelines of 10CFR100.

16

Table Q1 2-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies TransienVtAccidnt -

D: pDescription A e a Criteria s 152.1 Steam Pressure Regulator

• Not applicable, there are no pressure regulators In the McGuire or Ifunction or Failure That Results Catawba plants whose failure or malfunction could cause a steam n Decreasing Steam Flow flow transient.

15.2.2 Loss of External Load a Bounded by turbine trip analysis in Section 15.2.3 and same criteria apply.

• Peak RCS pressure remains below 110% of the design limit

(<2750 psia) 5.2.3 Turbine Trip

• Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

15.2.4 Inadvertent Closure of

• Bounded by turbine trip analysis in Section 15.2.3 and same Main Steam Isolation Valves criteria apply.

15.2.5 Loss of Condenser . Bounded by turbine trip analysis In Section 15.2.3 and same Causing a Turbine Trip critera apply.

• Peak RCS pressure remains below 110% of the design limit 15.2.6Loss f NonEmergncy * (<2750 psia) ower to the Station Auxiiaries Fuel cladding integrity shall be maintained by ensuring that the P t e t uacalculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110% of the design limit 15.2.7 Loss of Normal Feedwater (<2750 psla) 1low o Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110 % of the design Umit

(<2750 psia) for low probability events.

15.2.8 Feedwater System Pipe . Fuel cladding Integrity shall be maintained by ensuring that the Break calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• No hot leg boiling occurs.

• Peak RCS pressure remains below 110°%of the design limit 15.3.1 Partial Loss of Forced (<2750 psla)

Reactor Coolant Flow a Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

17

Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Trnin/AccidebntA cp eCiei

• Peak RCS pressure remains below 110% of the design limit 15.3.2 Complete Loss of Forced (<2750 psIa)

Reactor Coolant loow

• Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Peak RCS pressure remains below 110% of the design limit

(<2750 psia) 15.3.3 Reactor Coolant Pump

• Any fuel damage calculated to occur must be of sufficiently limited haft Seizure (Locked Rotor) extent that the core will remain in place and Intact with no loss of core cooling capability.

a Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10CFR1 00 guidelines.

15.3.4 Reactor Coolant Pump . Bounded by reactor coolant pump shaft seizure analysis In haft Break Section 15.3.3 and same criteria apply.

• Peak RCS pressure remains below 110% of the design limit 15.4.1 Uncontrolled Rod Cluster (<2750 psia)

Control Assembly Bank

• Fuel cladding Integrity shall be maintained by ensuring that the Withdrawal From a Subcritical or calculated DNB ratio remains above the 95/95 DNBR limit based Low Power Startup Condition on an acceptable DNBR correlation.

a Fuel centerline temperatures do not exceed the melting point

• Peak RCS pressure remains below 110% of the design limit 15.4.2 Uncontrolled Red Cluster (<2750 psia) 1ontrol Assembly Bank

• Fuel cladding Integrity shall be maintained by ensuring that the ithdrawal at Power calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

• Fuel centerline temperatures do not exceed the melting point.

• Peak RCS pressure remains below 110% of the design limit 5.4.3 Red Cluster Control (< 2750 psia)

Fuel cladding Integrity shall be maintained by ensuring that the e R , sodClusteratCon(Systrol

• sembly Misoperation (System calculated DNB ratio remains above the 95/95 DNBR limit based Drutonp rOeaorErr on an acceptable DNBR correlation.

nod Drop . Fuel centerline temperatures do not exceed the melting point a Peak RCS pressure remains below 110% of the design limit

(<2750 psia) 15.4.3 Rod Cluster Control

• Fuel cladding Integrity shall be maintained by ensuring that the sembly Misoperation (System calculated DNB ratio remains above the 95/95 DNBR limit based Malfunction or Operator Error) - on an acceptable DNBR correlation.

Single Rod Withdrawal . Fuel centerline temperatures do not exceed the melting point

• Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10CFR100 guidelines.

18

Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies

-Translent/Aicddent ampSta

A666-banq4Cr teria

'.aDe's'riptionin -y - A en ra

• Peak RCS pressure remains below 110% of the design limit 15.4A Startup of an Inactive (<2750 psia)

Reactor Coolant Pump at an

• Fuel cladding Integrity shall be maintained by ensuring that the Incorrect Temperature calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

15.4.6 Chemical and Volume

• Peak RCS pressure remains below 110% of the design limit Control System Malfunction that (<2750 psia)

Results in a Decrease In Boron

• Fuel cladding Integrity shall be maintained by ensuring that the Concentration in the Reactor calculated DNB ratio remains above the 95/95 DNBR limit based Coolant on an acceptable DNBR correlation.

15.4.7 Inadvertent Loading and . Any activity release must be such that the calculated doses at the peration of a Fuel Assembly In site boundary are a small fraction of the 10CFR1 00 guidelines.

an Improper Position

. Peak RCS pressure remains below 120% of design for very low probability events (< 3000 psla).

• Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based 15.4.8 Spectrum of Rod Cluster on an acceptable DNBR correlation.

Control Assembly Ejection a Any fuel damage calculated to occur must be of sufficiently limited Accidents extent that the core will remain In place and Intact with no loss of core cooling capability.

a The fission product release to the environment Is well within the established dose acceptance criteria of 10CFR1 00.

• See provisional cal/gm acceptance criteria attached.

a Peak RCS pressure remains below 110% of the design limit 15.5.1 Inadvertent Operation of (<2750 psia)

Emergency Core Cooling System

• Fuel cladding Integrity shall be maintained by ensuring that the During Power Operation calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

Chemical and Volume T5.5.2 Bounded by Inadvertent operation of emergency core cooling ontrol System Malfunction system durnng power operation analysis In Section 15.5.1 and Increases Reactor Coolant sm rtraapy Inventory sm rtraapy

• Peak RCS pressure remains below 110% of the design limit 15.6.1 Inadvertent Opening of a (<2750 psia)

Pressurizer Safety or Relief Valve . Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.

15.6.2 Break In Instrument Une or er Lines From Reactor Coolant

• Any activity release must be such that the calculated doses at the ressure Boundary That Penetrate site boundary are a small fraction of the 10CFR100 guidelines.

ontainment 19

Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies

-Transient/Accident - Acceptance ri . Critelia .

Descrpto

• Fuel cladding Integrity shall be maintained by ensuring that the 15.6.3 Steam Generator Tube calculated DNB ratio remains above the 95/95 DNBR limit based allure on an acceptable DNBR correlation.

• Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10CFR1 00 guidelines.

20

Table 012-3 Regulatory Dose Criteria For Accidents with MOX Fuel Lead Assemblies "Accident Cla'ssc Source Tr ReferencedSue NiTerm tidReference~-;

-!--g--;~  ;; -J^Off site Doses1 (EAanLZ_________

LOCA300 Rem Thyrold 1 CFR100.11 25 Rem TEDE RG 1.183 LOCA 25 em W~r 25 1CFR13001 ED 1eOCFR50.67 25 WB Rem ~SRP1 15.6.5 App. A ______

Steam Generator Tube Rupture 300 Rem Thyroid 1CR1- .1 25 Rem TEDE RG 1.183 with fuel failure or pre-incident Iodine spike 25 Rem WB SRP 15.6.3 SRG 15.196R118 Steam Generator Tube Rupture 30 Rem Thyroid 1 CFR100.11/ 2.5 Rem TEDEFG 1.183 with concurrent Iodine spike 2.5 Rem WB SRP 15.6.3 25RmICFRSO.67 SRP 15.196R118 Main Steam Line Break 300 Rem Thyroid RG 1.195 25 RemTEDEG 1.183 with fuel failure or pre-incident Iodine spike 25 Rem WB SRP 15.1.5 A 2RmE SRP 15.195 Ap Main Steam Une Break 30 Rem Thyroid 1OF1.195 2.5 Rem TEDE RG 1.183 with concurrent Iodine spike 2.5 Rem WB SRP 15.1.5 App. A2_5_Rem__ 1_183 Locked Rotor Accident 30 2.5Rem RemThyroid WI RG SRP1.19583 15.3.32.ReTEER118 Rod Ejection Accident 75 6.3 Rem Rem Thyroid W§ 2 RG SRP1.19583 15.4.8 App A6.ReTEEG118 Fuel Handling Accident 75 Rem Thyroid 2 RFG 1.19583 6.3 Rem W- SRP 15.7.4 6 Rem T R 1.183 50 Rem Throld3 RG 1.195 All 5 Rem WBO -OCFR50/ 5 Rem TEDE 10CFRI1 .67 50 Rem skin GDC 19 WB= Whole body, RG=Regulatory Guide, SRP= Standard Review Plan 2

Where a conflict exists between SRP and RG 1.195 on the whole body dose limit for aparticular accident, the more current guidance is shown.

3 RG 1.195 specifically states that this criterion may be used in lieu of the one in the SRP.

21

13. To allow the NRC staff to perform confirmatory analysis, please provide both the McGuire and Catawba loss-of-coolant accident (LOCA) input decks for the low enriched uranium (LEU) as well as the MOX fuel rods. Provide the decks in an electronic format, including nodalization diagrams.

Response

Framatome ANP is submitting the input decks to the NRC Staff in compact disc format by separate letter. The compact disc includes two RELAP5/MOD2-B&W input decks in UNIX format as follows:

r5moxnrc.in - Input deck for MOX fuel pins, power peaked at 10.3 ft.

r5uo2nrc.in - Input deck for LEU fuel pins, power peaked at 10.3 ft.

These are blowdown input decks used in the deterministic evaluations of MOX and LEU fuel pins reported in the license amendment request. The deterministic MOX fuel calculations comprise the licensing basis for the MOX fuel lead assemblies. Deterministic LEU fuel calculations were included to address the relative LOCA performance between MOX and LEU fuel.

Figures Q13-1 and Q13-2 are node diagrams for the decks. Figure Q13-1 shows the loop node arrangement while Figure Q13-2 shows the reactor vessel node arrangement. Figure 3-5 of Attachment 3 to Reference Q13-1 provides some additional detail specific to the core region.

REELAP5/MOD2-B&W is a derivative of the INEL code RELAP5IMOD2. Many changes were made to the INEL code to create the approved Framatome ANP deterministic LOCA code. Because the input for these changes may not be recognizable by other versions of RELAP5, the following list of related input card images is provided to assist the NRC staff.

Card 190: EM Choking Model Specification Card (Activates Framatome ANP specific choked flow break modeling.)

Card 192: EM Critical Flow Transition Data (Activates Framatome ANP specific critical flow break modeling.)

Card 195: Interface Heat Transfer Weighting (Activates Framatome ANP specific interface heat transfer weighting.)

Cards 10000020-10000029: Heat Structure Cards (Activate Framatome ANP specific filtered flow model - 10CFR50.46 Appendix K requirement.)

Cards 10000S80-1000OS99: Reflood Grid and Wall Heat Transfer Factor Data (Activate Framatome ANP specific grid model for droplet breakup and convective heat transfer due to grids.)

22

Cards ICCCG801-ICCCG899: Left Boundary Heat Structure Cards Cards lCCCG901-1CCCG999: Right Boundary Heat Structure Cards (Activate the Framatome ANP specific EM heat transfer package.)

Cards 19997000-19999999: EM Pin Model Specification (Activate Framatome ANP specific EM core package providing for dynamic fuel-clad gap conductance and fuel rod swell and rupture. Also provide the M5rm cladding properties.)

Reference Q13-1. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.

23

Figure 013-1 Loop Noding Diagram TSV mmI 690AX

.1 I

Triple Loop Single Loop 24

Figure Q1 3-2 Core Noding Diagram 364 IUpper Head l

El 360 l U'pper Plenum 358 300 CV 170 CV 280 1352-5 1445-1 345-1 435 345 443 343 442 342 350-3 341 346-3 440 -w-304 439 436 339 438 338 437 337 436 336 435 335 434 334 19 433 33 0

.60 332  :

itb 350-2 (5 432 346-2 c

I be b10 430 429 428 330 329 Ii 427 327 419 425 325 306 424 423 323 322 350-1 421 321 346-1 319 419 Rrr 417 317

,.-T 31-3* T14- 1315-1 I 314-2 308 1 314 L Lower Plenum 312 i Jo , I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

1 310 25

14. Provide the reference to the best estimate LOCA model noted in section 3.7.1.7.

Response

Based on RAI Questions 14, 15, and 16 it appears that some clarification is needed with respect to the LOCA analysis performed for the MOX fuel lead assemblies and how this analysis is used to support the lead assembly cores. In summary, the licensing basis for the resident Westinghouse RFA fuel remains the best estimate large break LOCA analysis performed by Westinghouse. Framatome ANP Appendix K analyses demonstrate that changing the fuel pellet material to MOX has no significant impact on peak cladding temperature following a large break LOCA. Framatome ANP Appendix K analyses provide peaking limits that ensure the peak cladding temperature for MOX fuel rods following a large break LOCA remain within the regulatory limit. The following discussion provides a further description of the analysis performed for the resident fuel assemblies as well as the MOX fuel lead assemblies.

Resident Fuel The resident fuel in MOX fuel lead assembly cores will be the robust fuel assembly (RFA) design that is supplied by Westinghouse. The large break LOCA analysis that supports this fuel design is the Westinghouse best estimate method described in Reference Q14-1. The analysis is based on the WCOBRAfTRAC method and includes detailed treatment of the uncertainties associated with the computer models and the inputs related with plant operation.

As part of the analysis, Westinghouse performed sensitivity studies to address transition or mixed core effects. This was necessary because the RFA fuel was initially introduced into cores containing Framatome ANP Mark-BW design fuel. The conclusion of the mixed core sensitivities was that the presence of the Mark-BW fuel assemblies had an insignificant impact on the calculated results. Westinghouse also performed small break LOCA calculations for McGuire and Catawba using the NOTRUMP methodology as described in Reference Q14-2. A mixed core penalty of 100F was assessed and applied to the small break LOCA results to accommodate the presence of the Mark-BW fuel assemblies. Given that the MOX fuel lead assemblies are more similar hydraulically to the RFA fuel than the Mark-BW design fuel, the mixed core penalty developed for the Mark-BW fuel assemblies bounds the MOX fuel lead assemblies. Therefore, the Westinghouse LOCA analyses for the resident RFA fuel remain valid in the presence of four MOX fuel lead assemblies.

MOX Fuel Lead Assemblies To address the MOX fuel lead assemblies, Framatome ANP performed deterministic large break LOCA calculations consistent with the requirements of 10 CFR 50 Appendix K In order to model accurately the effect of changing the fuel pellet material to MOX, Framatome ANP made modifications to their deterministic large break LOCA method as described in Reference Q14-3. These modifications are described in Section 3.7.1.2 of Attachment 3 to Reference Q14-4. Next, Framatome ANP performed large break LOCA calculations for a MOX fuel lead assembly as well as a Framatome ANP LEU fuel assembly, with both analyses assuming the hydraulic characteristics of the Advanced Mark-BW fuel assembly design. This sensitivity study was performed to assess the impact of the change in fuel rod parameters (MOX vs. LEU) on the calculated results. As discussed in Section 3.7.1.3 of Attachment 3 to Reference Q14-4, this sensitivity study showed that there is essentially no 26

difference between the LOCA results for the MOX fuel and the LEU fuel (APCT of 370F).

The Framatome ANP MOX fuel lead assembly results were also compared to the Westinghouse best estimate results to illustrate the similarity of the results. Given the differences in the two analytical methods, a direct comparison of the results is not completely valid. However, the comparison illustrates that the MOX fuel lead assembly with the lower peaking assumptions yields lower peak cladding temperature results (APCT of -380 F).

Following submittal of the MOX fuel license amendment request, Framatome ANP completed additional cases to investigate the impact of steam generator type, time in life, and axial power shape. Two different steam generator designs were examined: Westinghouse Model D5 steam generators (Catawba Unit 2), with a 10% tube plugging assumption; and BWI steam generators (Catawba Unit 1), with 5% tube plugging. The study concluded that the Model D5 steam generators with the 10% tube plugging assumption are limiting with respect to the Framatome ANP deterministic large break LOCA analysis. The D5 case provided the base case input for the other sensitivities cases.

Framatome ANP performed time in life sensitivities to assess the large break LOCA results as the stored energy in the fuel rod varies with cycle burnup. At burnups greater than 30 GWd/MThm, a KBU factor is applied to limit the PCT for these cases. The KBU factor reduces the FQ (total peaking factor) as well as the FAh (enthalpy rise factor or radial peaking factor).

Furthermore, using the limiting burnup case which uses a KBU of 1.0, i.e., the 30 GWdIMThm case, Framatome ANP evaluated power peaks at different elevations. The purpose of these sensitivities was to establish LOCA limits as a function of core height. At elevations above the 8 foot elevation a Kz factor was applied. The Kz factor reduces the FQ as well as the axial peaking factor (Fz).

A summary of the sensitivity cases is provided in Table Q14-1. The resulting LOCA peaking requirements for the MOX fuel lead assemblies are shown in Figure Q14-1. These peaking requirements will assure that the MOX fuel will comply with the regulatory limits for LOCA as provided in the response to Reactor System RAI Question 12.

MOX Fuel Lead Assembly Licensing Basis The licensing of the MOX lead assemblies will be based on analysis to determine the relative accident performance between the MOX and resident LEU assemblies because of the different fission source materials. As presented in the license amendment request, large break LOCA calculations, using the Framatome ANP deterministic LOCA evaluation model, have been performed for both LEU and MOX assemblies. The LEU calculations applied the evaluation model as approved by NRC. The MOX calculations applied the evaluation model with specified alterations, described in the LAR, necessary to simulate MOX fuel. The comparison of these two calculations demonstrated the expected result: that there is essentially no difference in the large break LOCA performance between fuel, of comparable design, using MOX pellets and fuel using LEU pellets. An evaluation of the small break LOCA, provided in the LAR, also determined that there would be no differences in the calculated results between the MOX and LEU fuel assemblies. Therefore, the assessment of the Catawba LOCA performance for the cores with four MOX lead assemblies is that LOCA performance is not altered. This result, in combination with a reduction in the allowed 27

peaking factor for the MOX fuel pins, provides the licensing basis for the MOX fuel lead assemblies assuring that all of the criteria of 10CFR50.46 are met.

References Q14-1. WCAP-12945P-A, Volume 1 Revision 2 and Volumes 2-5 Revision 1, Code QualificationDocument for Best-EstimateLoss of CoolantAnalysis, March 1998.

Q14-2. WCAP-100564P-A, Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code, August 1985.

Q14-3. BAW-10168P-A, Revision 3, RSG LOCA - BWNTLoss-of-CoolantAccident EvaluationModelfor RecirculatingSteam GeneratorPlants, December 1996.

Q14-4. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.

28

Table Q14-1 Summary of MOX Fuel Lead Assembly Large Break LOCA Sensitivity Cases Model D5 SGs with 10% Tube Plugging

.TT¶'XLT-=-.: Eeiti6-W; - - - , f;a-

>(GWd/Thm)> 0 il-zJ ___ au; -Fq - PCT(by H- ,2 BOL 6.8556 1.0 1.0 1.6 1.500 2.4 1919.2 20 6.8556 1.0 1.0 1.6 1.500 2.4 1943.6 30 6.8556 1.0 1.0 1.6 1.500 2.4 1948.8 50 6.8556 0.867 1.0 1.387 1.500 2.08 1824.4 60 6.8556 0.8 1.0 1.280 1.500 1.92 1787.6 30 4.7001 1.0 1.0 1.6 1.500 2.4 1815.0 30 8.5656 1.0 0.993 1.6 1.490 2.383 1964.0 30 10.2756 1.0 0.972 1.6 1.458 2.332 2019.5 Figure Q14-1 MOX Fuel Lead Assembly Total Core Peaking Factor 2.5 2.4 2.3

2.2 E

X t 2.1 E

8 S2 1.9 1.8 1.7 0 10 20 30 40 50 60 Burnup (GWd/MThm) 29

15. Provide the uncertainty analysis that was performed for the LEU and MOX LTA demonstrating that the 95/95 peak cladding temperature has been calculated for the core.

The response is expected to include a complete discussion of the statistical methodology used.

Response

The MOX fuel and LEU fuel LOCA analyses that support the use of the MOX fuel lead assemblies are deterministic calculations and therefore no uncertainty analysis was performed. See the response to Reactor Systems RAI Question 14 for additional explanation.

16. Section 3.7.1 states that the LOCA model used for the LEU fuel is a best estimate model.

Provide the Phenomena Identification and Ranking Table for the LOCA analyses performed with the best estimate model and reference the best estimate model used for the analysis.

Response

The Phenomena Identification and Ranking Table (PIRT) used in the Westinghouse best-estimate LBLOCA analysis is contained in Reference Q16-1. Since this method was not used to directly support the MOX fuel lead assemblies, this PIRT is not applicable to the MOX fuel lead assembly analysis. See the response to Reactor Systems RAI Question 14 for additional explanation.

Reference Q16-1. WCAP-12945P-A, Volume 1 Revision 2 and Volumes 2-5 Revision 1, Code QualificationDocumentfor Best-Estimate Loss of CoolantAnalysis, March 1998.

17. Provide the experimental data base used to assess the biases and to determine the uncertainties in the fuel rod behavior for the MOX LTA.

Response

The database is provided in Chapter 3 of the COPERNIC topical report (Reference Q17-1).

Additionally, at the NRC's request, several MOX fuel rods from the Halden experiments were analyzed with COPERNIC to end-of-life burnups in the range of 50 to 64 GWdIMThm.

Reference Q17-1. BAW-10231P Revision 2, COPERNIC Fuel Rod Design Computer Code, July 2000.

18. In sub-section 3.7.1.1.1, nothing is mentioned about the MOX/LEU interface behavior.

Provide a qualitative and quantitative discussion regarding the neutron flux behavior at the interface of the MOX and LEU fuel assemblies.

Response

Duke used the CASMO-4 computer code to model pin cell neutron flux and power at the intersection of four quarter-assembly lattices. These "colorsets" provide detailed two dimensional neutronic calculations that account for interface effects between dissimilar fuel assemblies. MOX fuel assemblies and LEU fuel assemblies of equivalent lifetime reactivity were placed in a checker board arrangement to simulate face adjacent MOX and LEU fuel assemblies (see Figure Q18-1). Two cases were considered- one using MOX fuel with weapons grade (WG) plutonium isotopics, and another using MOX fuel with reactor grade (RG) plutonium isotopics. The lead assemblies will be WG MOX fuel; the RG MOX fuel 30

cases are included for illustration purposes. Each case is examined at two burnup conditions:

one with all fresh fuel, and the other with 20 GWdIMThm burnup on each assembly (typical of the end of one cycle of operation).

The MOX fuel assemblies used plutonium concentration zoning as shown in Figure Q18-2. This is standard practice in European pressurized water reactors that use MOX fuel.

Low plutonium concentrations in the corner and peripheral pins serve to flatten the power distribution in the MOX fuel assemblies. The LEU fuel assemblies were a uniform lattice of 4.27 weight percent (w/o) 235U. The MOX and LEU fuel share the same dimensional characteristics (e.g., pellet diameter, cladding inner and outer diameter, and lattice pitch).

Figures Q18-3 through Q18-6 show how neutron flux and power change in a single row of pin cells traversing the MOXILEU fuel assembly interface. Pin cell locations -9 (minus nine) thru -1 (minus one) are LEU fuel pins. Pin cell locations 1 thru 9 are MOX fuel pins. The zero location on the x-axis of each graph corresponds to the center of the inter-assembly gap.

Fast flux (> .625 MeV), thermal flux (< 0.625 MeV), and power are each normalized to the colorset average value.

Figures Q18-3 and Q18-4 show the results for a WG MOX/LEU colorset at burnups of 0 GWd/MThm and 20,000 GWd/MThm, respectively. Figures Q18-5 and Q18-6 show the corresponding results for a RG MOXILEU colorset. The results are qualitatively similar between MOX fuel types and between burnups, with key points discussed below.

a) The fast flux is nearly uniform across the assemblies. Consistent with the results of the core simulations in the response to Reactor Systems RAI Question 11, the fast flux in MOX fuel is only slightly higher than the fast flux in LEU fuel.

b) There is a steep thermal flux gradient between the assemblies. The flux gradient results from the fact that the thermal neutron absorption cross-section in plutonium is larger than in uranium.

c) The thermal flux gradient is more pronounced for the RG MOX fuel case than for the WG MOX fuel case. The higher overall plutonium concentration in the RG MOX fuel assembly (7.07 w/o) depresses the thermal flux more than the WG MOX fuel assembly with only 4.37 w/o plutonium. Therefore, the thermal flux gradient from the LEU to the MOX fuel assembly is steeper in the RG MOX fuel case. It is not unexpected to find that WG MOX fuel behavior falls between that of LEU fuel and RG MOX fuel; this is consistent with studies of other neutronic characteristics of MOX fuel (Reference Q18-1).

d) Most importantly, the plutonium concentration zoning in the MOX fuel assemblies is effective in producing a relatively flat power profile across the LEU and MOX fuel assemblies. This conclusion is valid for both WG MOX fuel and RG MOX fuel. The edge and corner pins MOX pins see a much higher thermal flux than the interior fuel pins, but the lower plutonium concentration in the edge and corner pins makes the fission rate and power about the same as the interior pins. The response to a LOCA is driven by the pin power profile, not by the pin flux profile. Therefore, the neutron flux behavior at the interface of these MOX and LEU fuel assemblies should have no significant impact on the cladding temperature response following a loss of coolant accident.

31

Reference Q18-1. BAW-10238(P), Revision 1, MOXFuel Design Report, Section 3.1, Framatome ANP, May 2003.

32

Figure Q18-1 MOX Fuel and LEU Fuel Colorsets Reactor Grade MOX Figure Q18-2 MOX Fuel Assembly Zoning Pin Number RG w/o Pu WG w/o Pu 1 176 8.00 4.94 2 76 5.42 3.35 3 12 3.89 2.40 Total i Avg 264 7.07 4.37 33

Figure Q18-3 Weapons Grade MOX/LEU Fuel Colorset (0 GWd/MThm) 2.00 -

1.75 _

a)

0) 1.50.

% 1.25 U)

.° 1.00 0

C.)

_2 0 0.75 -

.! 0.50 a!

0.25

-~ ~~~ . ~ 1 WC. iyw Ig,S, 0.00 k] ~~~~~~~~~I I I

-10 -8 -6 -4 -2 0 2 4 6 8 10 Pin Cells Figure Q18-4 Weapons Grade MOXILEU Fuel Colorset (20 GWd/MThm) 2.00 1.75 al)

Lv 1.50 0~ ~ y x.00 - f0 1.25 0)

I-

.20 0. 75 . 2 ,

1.00 <~~ ~ ~ ~ ~ ~ ~ ~ :\~~~i, Cells.F;.sS...:DS:)

-10 -80 -6 -4 ;' -2 0'S 0 C el 2 4 6 8 10 34

Figure Q18-5 Reactor Grade MOX/LEU Fuel Colorset (OGWd/MThm) 2.00 1.75 E 1.50 . N 1.25 o 1.00 0.75

' 0.50 0.25 0.00

-10 -8 -6 -4 -2 0 2 4 6 8 10 Pin Cells Figure Q18-6 Reactor Grade MOX/LEU Fuel Colorset (20 GWd/MThm) 35 C (V

, '1 i.-

19. Section 3.7.1.1.1 discusses a variety of neutronic parameters. Provide additional detail about the differences between LEU and MOX parameters. Please use graphs, data, or any other visual representations to help clarify the impact of these parameter differences.

Response

Information on differences between LEU and MOX fuel neutronic parameters was provided in Tables 3-7 through 3-10 and Figure 3-2 in Attachment 3 to Reference Q19-1. Additional information is provided in the responses to Reactor Systems RAI Questions 11, 18, 20, and 23.

Reference Q19-1. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.

20. Also, in the second paragraph of sub-section 3.7.1.1.1, the change in the delayed neutron fraction is discussed at beginning of life. However, the behavior of the delayed neutron fraction at middle of life and end of life is not addressed. Provide a discussion on the delayed neutron factor change with respect to burnup.

Response

Table Q20-1 and Figure Q20-1 show the variation in core average effective delayed neutron fraction (0,ff) with cycle burnup and MOX fuel loading. The introduction of four MOX fuel lead assemblies results in a minor decrease in beginning of cycle (BOC) Peff. The impact on end of cycle (EOC) Jeff is even smaller. The 40% MOX fuel core sees a more significant reduction in uff at BOC, but the change in neff with burnup is much smaller than the corresponding change in a core containing all low enriched uranium (LEU). The result is that the 40% MOX fuel core has more uniform kinetics over the cycle.

36

Table Q20-1 Impact of MOX Fuel Loading and Burnup on Effective Delayed Neutron Fraction (fPeff)

MOX Lead Batch MOX EFPD LEU Core' Assembly Core2 Core3 4 6.216E-03 6.089E-03 4.827E-03 12 6.180E-03 6.060E-03 4.829E-03 25 6.125E-03 6.015E-03 4.825E-03 50 6.026E-03 5.929E-03 4.81 OE-03 100 5.852E-03 5.772E-03 4.768E-03 150 5.704E-03 5.634E-03 4.726E-03 200 5.579E-03 5.516E-03 4.686E-03 250 5.471 E-03 5.413E-03 4.650E-03 300 5.377E-03 5.322E-03 4.615E-03 350 5.292E-03 5.240E-03 4.583E-03 400 5.216E-03 5.166E-03 4.551 E-03 450 5.146E-03 5.098E-03 4.521 E-03 490 5.086E-03 5.041 E-03 4.502E-03 1 No MOX fuel assemblies 2 4 MOX fuel assemblies 3 76 MOX fuel assemblies (-40% MOX fuel core)

Figure Q20-1 Impact of MOX Fuel Loading and Burnup on Effective Delayed Neutron Fraction (Peff) 0.0065 0.0060 I

wI 0.0055 I 0.0050 0.0045 0.0040 0 100 200 300 400 500 Cycle Exposure, EFPD 37

,' r* ,,

23. All operating plants must have sufficient shutdown margin at the beginning and throughout a fuel cycle. Provide the predicted shutdown margin in graphical or tabulated form for the cores including the four MOX lead test assemblies. Show that the predicted shutdown margin will meet the Technical Specification limit for each of the plants that may load the LTAs.

Response

Results of quantitative Mode 1 shutdown margin calculations are presented in Table Q23-1 for a representative core consisting of 189 LEU fuel assemblies and four unirradiated MOX fuel lead assemblies. The results are compared against the same core with 193 LEU fuel assemblies. Shutdown margins are shown at beginning, middle, and end of cycle.

The calculated shutdown margins include standard conservatisms. Operation at the control rod insertion limit is assumed. Power defect is increased and control rod worth is decreased to account for uncertainty in those values. It is assumed that the highest worth control rod remains stuck out of the core, and the worth of that rod is conservatively increased.

The resulting shutdown margins are substantially greater than the current Catawba Mode 1 minimum shutdown margin operating limit of 1300 pcm. Furthermore, there are no significant adverse impacts on calculated shutdown margin due to the presence of four MOX fuel lead assemblies. The differences observed are within normal fuel cycle design variations.

As part of the standard Catawba reload design process, Duke calculates control rod worths and verifies adequate shutdown margin. This same standard reload design process will be applied to the design of the actual MOX fuel lead assembly cores, thereby ensuring that shutdown margin limits are met.

Table Q23-1 Shutdown Margins Impact of Four Fresh MOX Fuel Lead Assemblies at Catawba r> ~-TType of Gore '§--CacltdSudwn clabShtdo "Ma- I (pnm Mrgin.

___________.____-____:_-_-_____ 4 EFPD . 250 EFPD E PD E490 All LEU Fuel 3237 2498 2020 4 Fresh MOX Fuel Lead Assemblies 3255 2465 1988 38

- w -s

28. In sub-section 3.7.1.6, the subject of mixed cores is discussed. In the middle of the paragraph it is stated that the MOX LTA pressure drop is less than four percent lower than the pressure drop for a resident Westinghouse fuel assembly at design flow rates. Please provide additional detail on the cause of this pressure drop difference, how it was calculated, and the impact including the consequences of this pressure drop. Also, please provide the design flow rate used for this analysis.

Response

The pressure drop difference between the resident Westinghouse Robust Fuel Assembly (RFA) fuel and the MOX fuel lead assemblies is due to mechanical design differences in the grids and the top and bottom nozzles of the fuel assemblies. Even though the rod geometry, pitch, and axial grid locations are the same, unique design differences in the grids and nozzles themselves cause differences in hydraulic resistance. This overall difference was calculated by evaluating full core RFA and full core MOX models with the VIPRE-01 thermal-hydraulic code and comparing the overall calculated Ap. The code represents these hydraulic differences by means of vendor-provided form loss coefficients for each grid design, top, and bottom nozzles. The design flow rate for these evaluations was the current Technical Specification minimum flow rate of 390,000 gpm.

The impact of this difference in pressure drop is flow redistribution between fuel types in a mixed core environment. This redistribution varies with axial elevation in the core as a direct effect of the difference in local grid form loss coefficients. The consequences of this pressure drop difference result in the need to account for this flow redistribution in the analyses of fuel assembly lift, departure from nucleate boiling ratio (DNBR) in steady state and transient analyses, and fuel assembly performance issues such as maximum allowable crossflow. Flow redistribution is accounted for in these analyses by modeling the hydraulic differences directly in a conservative representation of the mixed core fuel assembly geometry.

29. The staff presumes that a mixed core analysis will be performed to account for the use of four MOX LTAs in the core. Therefore, provide the mixed core penalty that was calculated.

If a mixed core calculation was not performed, provide a technical justification for not performing the analysis.

Response

The mixed core MOX fuel lead assembly DNBR penalty is explicitly calculated for the entire range of conditions analyzed in a reload cycle. With the currently licensed Duke Power analysis methodology, maximum allowable radial peaking limits are calculated for a range of axial peak locations and magnitudes as described in DPC-NE-2004P-A. This family of peaking limits is repeated for the various sets of reactor statepoints (power level, pressure, temperature, and flow) analyzed to support cycle reload analyses. This entire set of limits is used to represent the limiting fuel assembly in the core.

To model the mixed core, a bounding model of a single high powered MOX fuel assembly at the center of the core surrounded by a remaining core of resident Westinghouse RFA fuel assemblies was used to calculate the explicit peaking limits. This model contained the 39

correct geometry and local form loss coefficients to represent both fuel types (see response to Reactor Systems RAI Question 28). Therefore, the mixed core peaking penalty is calculated for each unique set of conditions and the appropriate conservative limits will be applied to the lead assembly core positions in the specific cycle reload analyses. This penalty magnitude varies as a function of the axial power distribution, with the overall average penalty equal to 3% in radial peaking or 10% in DNBR, relative to a full core of Westinghouse RFA fuel assemblies.

31. Section 3.7.2.2 discusses the thermal-hydraulic differences for the different co-resident fuel types. Please provide the limits analyzed for the different fuel types, including a discussion on how they were obtained.

Response

The co-resident fuel types, Westinghouse RFA and Mark BwIMOXI, will be analyzed with their respective critical heat flux (CHF) correlations and limits. The RFA fuel will be analyzed with the Westinghouse WRB-2M CHF correlation with a design limit deparature from nucleate boiling ratio (DNBR) value (95/95) of 1.14 for deterministic analyses and a design limit DNBR value of 1.30 for analyses where uncertainties are combined statistically.

The Mark BW/MOX1 fuel will be analyzed with the Framatome BWU-Z CHF correlation with a design limit DNBR value (95/95) of 1.19 for deterministic analyses and a design limit DNBR value of 1.36 for analyses where uncertainties are combined statistically.

Derivation of these limits is described in References Q31-1, Q31-2, and Q31-2.

References Q31-1. DPC-NE-2005P-A, Revision 3 (Appendix E), Duke Power Company Thermal-Hydraulic StatisticalCore Design Methodology, September 2001.

Q31-2. DPC-NE-2009P-A, Revision 2, Duke Power Company Westinghouse Fuel TransitionReport, December 2002.

Q31-3. WCAP-15025P-A, Modified WRB-2 Correlation, WRB-2M, for PredictingCritical Heat Flux in 17x1 7 Rod Bundles with Modified LPD Mixing Vane Grids, February 1998.

35. Provide information on how the peaking factor limits used for MOX fuel were developed.

Response

During the August 13, 2003 telephone conference among Duke, Framatome ANP, and the NRC staff, the staff clarified that the specific peaking factor limits of interest are LOCA peaking limits as provided in Table 3-6 of the February 27, 2003 License Amendment Request. This response was developed accordingly.

The MOX fuel assemblies are limited to operation within the allowed peaking presented in Table 3-6 of the license amendment request. The allowed peaking was established using the same approach as for any fuel design. A target allowable peaking (Limiting Condition for Operation) was determined considering plant operability and the required fuel service.

The acceptability of that target peaking was then confirmed through appropriate analysis 40

f -. ik.-

(design, safety, and LOCA). The considerations included in establishing the MOX fuel assembly target limit for normal operation were:

a) MOX fuel lead assemblies are demonstration lead assemblies and it is therefore appropriate to control them to somewhat reduced limits b) Recognition that demonstration lead assemblies should be operated with fairly high duty in order to promote the discovery of issues that should be addressed prior to batch implementation, c) The insertion of four demonstration lead assemblies should not impose an operational hardship on the irradiating plant.

As discussed in Section 3.7.1.4 of Reference Q35-1, additional LOCA studies were performed after the submittal of the license amendment request to further refine the LOCA peaking limits for the MOX fuel lead assemblies. These studies included:

a) Time-in-life (bumup) sensitivity to determine any burnup dependent limitation on the fuel, and b) Axial power distribution sensitivity to confirm the Kz curve.

The results of these studies are discussed in the response to Reactor Systems RAI Question 14.

Reference Q35-1. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.

41