ML110040374

Calvert Cliffs - Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to Areva Nuclear Fuel

ML110040374
Person / Time
Site:	Calvert Cliffs
Issue date:	12/30/2010
From:	Gellrich G H Calvert Cliffs, Constellation Energy Group, EDF Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 10-101
Download: ML110040374 (83)

Text

{{#Wiki_filter:George H. Gellrich Vice President Calvert Cliffs Nuclear Power Plant, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.5200 410.495.3500 Fax CENG a joint venture of Constellation -Dn SEnergy, Ot CALVERT CLIFFS NUCLEAR POWER PLANT NRC 10-101 December 30, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

REFERENCES:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (a) Letter from Mr. G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated October 29, 2010, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (b) Letter from Mr. G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated November 19, 2010, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (c) Letter from Mr. T. E. Trepanier (CCNPP) to. Document Control Desk (NRC), dated November, 23, 2009, License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel On August 23 and 24, 2010, the Nuclear Regulatory Commission (NRC) staff conducted an audit of analyses related to the proposed license amendment to support the transition from Westinghouse nuclear fuel to AREVA Advanced CE-14 High Thermal Performance fuel. A number of questions were raised by the NRC staff during the audit. Responses to the questions were provided in References (a) and (b).After review of the responses, the NRC staff requested additional information. This supplemental information was discussed during a followup audit conducted on December 8 and 9, 2010 and some of the requested supplemental information is contained in Attachment (1). This supplemental information does not change the No Significant Hazards determination previously provided in Reference (c). Document Control Desk December 30, 2010 Page 2 Attachment (1) contains information that is proprietary to AREVA, therefore, it is accompanied by an affidavit signed by AREVA, owner of the information (Attachment 2). The affidavit sets forth the basis on which information may be withheld from public disclosure by the Commission, and address, with specificity, the considerations listed in 10 CFR 2.390(b)(4). Accordingly, it is requested that the information that is proprietary to AREVA be withheld from public disclosure. The non-proprietary version of Attachment (1) is included (Attachment 3).Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.Very truly yours, STATE OF MARYLAND COUNTY OF CALVERT: TO WIT: I, George H. Gellrich, being duly sworn, state that I am Vice President -Calvert Cliffs Nuclear Power Plant, LLC (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.Subscribed and sworn before me a Notary yublic iq and for the State of Maryland and County of ,this (O day of Lec t, 2010.(a,-/ /' L fa, I Notarial Seal: I Notary Public Date My Commission Expires: GHG/PSF/bjd Document Control Desk December 30, 2010 Page 3 Attachments: (1) Proprietary Supplement to License Amendment Request: Transition to AREVA Nuclear Fuel (2) AREVA Proprietary Affidavit (3) Non-Proprietary Supplement to License Amendment Request: Transition to AREVA Nuclear Fuel cc: [Without Attachment (1)]D. V. Pickett, NRC W. M. Dean, NRC Resident Inspector, NRC S. Gray, DNR ATTACHMENT (2)AREVA PROPRIETARY AFFIDAVIT /WESTINGHOUSE PROPRIETARY AFFIDAVIT Calvert Cliffs Nuclear Power Plant, LLC December 30, 2010 AFFIDAVIT COMMONWEALTH OF VIRGINIA )) ss.CITY OF LYNCHBURG )1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in the attachment to a letter from G.H. Gellrich (Calvert Cliffs Nuclear Power Plant) to Document Control Desk (NRC)entitled "Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel," numbered NRC 10-101 and referred to herein as"Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

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

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in I accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial'- information." 6. The following criteria are customarily applied by AREVA NP to determine/whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, j or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this 6-day of ______________2010.

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/11 Reg. # 110864 S."KATHLEEN ANN BENNETT 5 4Notary Public] Commonwealth of Virginia oio prA ,21108640.MY commission Expires Aug 31, 20!1 I i ----m l v v v W W W ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Calvert Cliffs Nuclear Power Plant, LLC December 00, 2010 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Calvert Cliffs Nuclear Power Plant, LLC December 30,2010 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Based on a review of References 1 and 2, the Nuclear Regulatory Commission (NRC) staff has identified the need for additional information. The additional questions provided by the NRC staff and our responses are below.Question 1: In response to Question lb, the licensee states that a reactor trip terminates the asymmetric steam generator transient (ASGT) event prior to the development of any asymmetry at the core inlet. Based on this position, the ASGT is modeled using a uniform core inlet flow and temperature distribution. This modeling assumption essentially removes the unique aspects of the ASGT including the asymmetric core inlet temperature distribution and resulting core power tilt. The existing Calvert Cliffs Nuclear Power Plant (CCNPP) licensing basis for the ASGT specifically captures the asymmetric core inlet flow distribution. Without new information, the staff is unable to accept this change to the CCNPP licensing basis. The staff requests that the ASGT be re-analyzed using a justified asymmetric core inlet temperature distribution. CCNPP Response 1: Response to be provided later.Question 2: In response to Questions Ja and 7, the licensee states that "modeling assumptions for flow mixing in the lower plenum do not have a first order effect on the minimum DNBR. " The response also states that inlet flow distributions are "washed out quickly" in an open lattice PWR core. Based on this position, the single Reactor Coolant Pump Locked Rotor event is modeled using a uniform core inlet flow and temperature distribution. This modeling assumption essentially removes the unique aspects of the asymmetric core inlet flow distribution resulting from a coast down from 4-pump to 3-pump conditions. The existing CCNPP licensing basis for the LR event specifically captures the asymmetric core inlet flow distribution. Without new information, the staff is unable to accept this change to the CCNPP licensing basis. The staff requests that the Locked Rotor minimum DNBR be re-calculated using the existing 3-pump limiting assembly inlet flow factor.CCNPP Response 2: In the standard AREVA method for calculating the departure from nucleate boiling (DNB) for the Locked Rotor Transient, the impacted flow region is limited to the assembly with the highest power and the four face adjacent fuel assemblies. For those assemblies, a flow penalty is applied to the mass flux in the XCOBRA-IIIC. To evaluate the effect of a non-symmetrical flow distribution, additional DNB sensitivities were conducted, including: o Varying the flow factor applied to the hot assembly, o Modifying the flow distribution in the core by reducing the flow concentrically to the limiting bundle.From these sensitivities, the following flow inputs and results are presented in Table 2-1: o the standard analytic method, o the flow factor applied to the power limiting assemblies [] ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL o the flow factor applied to the power limiting assemblies reduced from [ ] and an increased impacted region (see core map on Figure 2-1).A map of the core configuration, showing the impacted region for the flow gradient case, is provided as Figure 2-1. The comparison of coolant mass flux and crossflow between the three cases are plotted in Figure 2-2 and Figure 2-3.The results show that at the time of minimum departure from nucleate boiling ratio (DNBR) the trends are the same as those calculated with the standard AREVA method. The analysis positively demonstrates that the flow distributions are 'washed out' by the time of minimum DNBR. Therefore AREVA's modeling assumption on the flow distribution is validated. Table 2-1: Seized Rotor Results 2 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 2-1: Increased Impacted Region -Flow Gradient 3 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 2-2: Mass Flux of Limiting Bundle 4 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 2-3: Cross Flow of Limiting Bundle Ouestion 3: Based on the licensee's letter dated November 19, 2010, the staff has identified the need for additional information. In response to Question 2d, AREVA appears to have run some RODEX-2 cases for the CEA drop transient. The maximum cladding strain is calculated to be 0.78% strain at 16.4 GWd/MTU The CEA drop case was selected because it exhibited the "peak attainable linear heat rate." A peak LHGR is limiting with respect to fuel centerline temperature. However, the maximum change in LHGR is limiting with respect to calculated cladding strain. For each AO0 and accident, please provide the peak calculate cladding strain along with the pre- and post-LHGR. Specify radial (Fr) and axial power peaking (F 7) components. CCNPP Response 3: Response to be provided later.Ouestion 4.a: The response to Question 3 does not provide sufficient information to address the staff's concerns. The staff has prepared this request for follow-up information using two examples: the quasi-steady state control element assembly (CEA) withdrawal error at power (CWAP) and the transient CEA drop events.5 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL a. CWAP The CWAP transient is analytically terminated by the variable high power trip (VHPT). It is asserted that the limiting power ascension occurs with the zero-power transient because the trip ceiling at that point provides for the longest trip delay, and with reactor power greater than 30-percent, the trip setpoint is 10 -percent greater than the reactor power level.The response provided considers various phenomena as separate effects, including perturbations in peaking factor, core response characteristics, fractional power level, and setpoint methodology conservatisms. The response asserts that each of these effects provides sufficient conservatism in the analysis to assure that a hot zero power and a HFP analytic case are bounding of power levels in between. The conservatism is unquantified and is not supported with analytic examples.Please perform a sensitivity analysis of the CWAP transient to demonstrate the effectiveness of the VHPT as afunction of reactor power.It is stated, "The [ I is proportional to the fraction of power, so that at powers below 60% this effect overshadows the increases in local peaking from the axial shape index and CEAs at lower powers. " Demonstrate that this is true. At each power level, consider limiting achievable initial axial shape indices (ASIs) and address the attendant DNBR effects. Also include appropriate consideration of the transient radial power redistribution. CCNPP Response 4.a: Response to be provided later.Question 4.b: b. CEA Drop The CEA drop event is evaluated for both minimum departure from nucleate boiling ratio and for fuel centerline melt. The peak linear heat generation rate is predicted based on a steady-state evaluation of the end state power level, and a linear heat rate calculation factoring in the maximum allowable Technical Specification peaking factor values. The CEA drop event, however, causes a transient change to the peaking factors.While the response to RAI 3 states that setpoint verification calculations account for transient variations in local power distribution, this transient is unmitigated by a reactor trip. The response to RAI 3 also asserts that proportionality in linear heat rate provides adequate margin to SAFDLs to assure that lower-power transients remain bounded by the HFP analyses.Demonstrate that the above discussed phenomenology holds true for the CEA Drop transient. Analyze the hot full-power sensitivities to dropped rod worth, initial and final power level, and limiting power distributions and redistributions at lower power levels. Confirm that the limiting results from these studies are bounded by the return to power and radial peaking augmentation factors obtained from the full-power neutronic analyses [ J.CCNPP Response 4.b: Response to be provided later.6 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Question 5: Withdrawn CCNPP Response 5: N/A Question 6: The staff will request a license condition from CCNPP to limit itself from changing Technical Specification COLR Figures 3.1.6, 3.2.1-2, 3.2.3, or 3.2.5 without prior NRC review and approval until an NRC-accepted, generic or CCNPP-specific basis is developed for analyzing power level-sensitive transients (Control Rod Bank Withdrawal, CEA Drop, and CEA Ejection) at full power conditions only.CCNPP Response 6: Calvert Cliffs proposes the following license condition to address the concerns raised in Question 6: "For Unit I and Unit 2: Core Operating Limits Report Figures 3.1.6, 3.2.3, and 3.2.5 will not be changed without prior NRC approval until an NRC-accepted generic, or CCNPP-specific, basis is developed for analyzing the Control Rod Bank Withdrawal event, the Control Element Assembly Drop, and the Control Element Assembly Ejection (power level-sensitive transients) at full power conditions only." Core Operating Limits Report (COLR) Figure 3.2.1-2 will change for Unit -2 (2011 RFO) and Unit 1 (2012 RFO) as shown in Figure 6-1 which provides both the current Unit 2 Cycle 18 (U2C 18) and Unit 1 Cycle 20 (U1C20) limits as well as the proposed Unit 2 Cycle 19 (U2Cl9) limits. Note that the U2C19 Acceptable Operation Region is significantly more restrictive as compared to U2C18/U1C20. Core Operating Limits Report Figure 3.2.1-2 is only used when operating on excore detectors when the incore detector system is out of service. Peripheral axial shape index is monitored to maintain the plant within the power dependent axial shape index limits defined by the linear heat generation rate limiting condition for operation barn in COLR Figure 3.2.1-2. Core Operating Limits Report Figure 3.2.1-2 must change as part of the change from Westinghouse to AREVA fuel. AREVA approved methodology cannot support the current COLR Figure 3.2.1-2 limits, therefore more restrictive limits must be implemented for cores analyzed using AREVA methodology. For future cycles, COLR Figure 3.2.1-2 will be determined by performing cycle specific setpoint calculations and the axial shape index limits will always be more restrictive than the current U2C 18 and U 1 C20 limits.7 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL a, U.0 z I-C., 1.10 1.05 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20-0.60-0.01 UNACCEPTABLE OPERATION REGION/91, 0.950/ \0.075, ACCEPTABLE %OPERATION REGION-0.264, 0.600 0.248, 0.600 0.950 UNACCEPTABLE OPERATION REGION 2$I I I I I I I I/I ,A.I I I I I I I I I I-U2C19-->-- U2C18-0.264, 0.200 0.248, 0.200 2 I 0 I-0.40-0.20 0.00 0.20 0.40 0.6(PERIPHERAL AXIAL SHAPE INDEX, Yi Figure 6-1: Change to COLR Figure 3.2-1-2 Ouestion 7: Withdrawn CCNPP Response 7: N/A Ouestion 8: The response to Question 3 provides no evidence that CEA ejection events initiated at mid-power conditions along the COL PDIL are bounded HZP and HFP conditions. Please provide CEA ejection cases at several at-power conditions which capture power-dependent parameters such as PDIL LCOs, ASI LCOs, power peaking LCOs, LSSSs, and power measurement uncertainty. For example, one scenario might be allowable initial conditions at 19.9% power (e.g., Ejected rod worth spanning up to Bank 3 60% inserted, most severe AXPD, etc.) while initiated at a higher power corresponding to 19.9%plus power measurement uncertainty (e.g., secondary calorimetric uncertainty) along with a similarly decalibrated high power trip.8 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL CCNPP Response 8: The primary variable determining the severity of this event is the ejected CEA worth. The AREVA methodology for calculating the total enthalpy (cal/g) associated with the ejected CEA worth is very conservative. Conservatively high ejected CEA worths are applied when the cycle-specific values are below a threshold value. To demonstrate the extent of this conservatism, the minimum (i.e., most limiting) threshold values for hot zero power and hot full power are shown in Figure 8-1 along with the calculated cycle-specific values at hot zero power, 20 percent power, and hot full power.The small difference of the ejected CEA worth from hot zero power to 20 percent power as shown in Figure 8-1 is investigated. The power dependence of the power dependent insertion limit determines the initial depth of the ejected CEA and deeper insertion tends to increase the worth. The power dependent insertion "limit curve for Calvert Cliffs does not allow deeper bank insertion for powers lower than 20 percent rated thermal power. To examine the effect of a deeper insertion at hot zero power, the bank position from a linear extension of the power dependent insertion limit curve (see Figure 8-2) is used for the boundary conditions for the ejected CEA calculations. The ejected CEA worth for this case is combined with the data from Figure 8-1 and a line is drawn between the hot zero power and hot full power cases in Figure 8-3. The linear relationship of the ejected worth between the newly defined hot zero power case and existing hot full power case bound the calculated values at 20 percent power and represents an improved relationship to bound the cycle specific behavior. As shown in this figure, the current analysis conservatively bounds the effect of the extension of the power dependent insertion limit on ejected worth. Hence, the analysis remains bounding.The calculated post ejection Fqs are shown in Figure 8-4, which is analogous to Figure 8-3 for the ejected CEA worths. As with ejected CEA worth, the AREVA methodology applies a conservatively high Fq for the cal/g calculation. In this example, the cal/g calculation for the beginning of cycle uses a different threshold value than the end of cycle. The trend with power with hot zero power data at the power dependent insertion limit position yields a reasonable bound for the 20 percent data as does.the data from the extended power dependent insertion limit. Neither is more conservative than the other. All the data is less than the threshold Fqs used in the cal/g calculation. Hence, the analysis remains bounding.Therefore, the hot zero power and hot full. power cal/g analysis sufficiently bounds mid-power conditions. 9 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 600-500-8. 400 _A BOC@20% FP oEOC @20% FP 300 -Cal/g Analysis Q A x BOC HZP-HFP 200 I [] EOC HZP-HFP 100-0 20 40 60 80 100 Power Figure 8-1: Ejected CEA Worth Versus Power 10 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL (1.QUfKW. eo4p $a 35% gVsene F 0800 a 0.700 D. 0.600 S0.500.0.00 0.300.S,.0 Z 0.200 0 0.100 LL ._0% W0% 4M% W% 10% c% X 0% O Nl 8c% lD 2%W % 0% Wn 40 M0 2% 500%43ul St Be Wr 2r W i to ir mw 2 'r w~ w ev w4 zr a-0%A 2% 4M% 0%~ 2% lWO% 0% O% ~0% W0% W0% I WA~117 1W W sý4 ir a' s~s icr esi se ar w%CEA INSERTION ,INCHES CEA WITHDRAWN (ARO is defined in NEOP-23)Figure 8-2: CEA Group Insertion Limits Versus Power 11 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 600_ _500 _E -- BOC Ext PDIL C. 400- ---EOC Ext PDIL ABOC @20% FP 300 *_ EOC @20% FP x BOC @PDIL o EOC @PDIL S200- -cal/g analysis 100 0 20 40 60 80 100 Power Figure 8-3: Ejected CEA Worth Versus Power with Trend 13 12 _11 10 0- BOC Ext PDIL 9 ___ EOC Ext PDIL A BOC @20% FP 8 -m EOC@20%FP U- 7 BOC @PDIL 6

• _ EOC @PDIL 6 1-BOC cal/g analysis 5 --EOC cal/g analysis 4 3 _0 20 40 60 80 100 Power Figure 8-4: Post Ejected Fq versus Power 12 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Question 9: The response to Question 4 appears to be non-responsive to the staff's technical concern.In Question 4, the staff requested that the licensee address pressurization effects of allowable initial conditions other than nominal. In response, the licensee stated that the S-RELAP5 code is being used to analyze for conformance to specified acceptable fuel design limits, and not to analyze reactor coolant pressure boundary integrity.

Because this response does not address the staff's technical concern, we would need to indentify a way to restrict our approval of S-RELAP5 to only those safety analyses that confirm acceptable transient performance relative to the specified acceptable fuel design limits, and require prior, transient-specific NRC review and approval of any use of S-RELAP5 to demonstrate reactor coolant pressure boundary integrity. If the licensee is amenable to proceeding in this fashion, we would close Question 4 as resolved, pending the development of the appropriate license condition. If not, we will need to identify an alternative path forward CCNPP Response 9: Calvert Cliffs proposes the following license condition to address the concerns raised in Question 9: "For Unit 1 and Unit 2: Approval of the use of S-RELAP5 is restricted to only those safety analyses that confirm acceptable transient performance relative to the specified acceptable fuel design limits. Prior transient specific NRC approval is required for transient performance relative to reactor coolant pressure boundary integrity until an NRC-accepted, generic, or CCNPP-specific, basis is developed for the use of S-RELAP5 to demonstrate reactor coolant pressure boundary integrity." Question 10: For the SBLOCA please provide the plots of the key system parameters for the breaks re-analyzed that are provided in the UFSAR. No plots were provided nor were the tables summarizing the timing for the key events provided CCNPP Response 10: Plots of key system parameters from the small break loss-of-coolant accident (LOCA) re-analysis are presented. The set of figures below represents the 0.07 ft 2 , 0.08 ft 2 and 0.09 ft 2 break sizes, consistent with the set presented in the Updated Final Safety Analysis Report (UFSAR) for the small break LOCA analysis.13 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Plots for 0.07 ft 2 Break Figure 10-1: Core Power for 0.07 ft 2 Break 14 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-2: Inner Vessel Pressure for 0.07 ft 2 Break 15 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-3: Break Flow for 0.07 ft 2 Break 16 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-4: Core Inlet Flow Rate for 0.07 ft 2 Break 17 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-5: Hot Assembly Collapsed Liquid Level for 0.07 ft 2 Break 18 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-6: Hot Spot HTC for 0.07 ft 2 Break 19 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-7: Hot Spot Coolant Temperature for 0.07 ft 2 Break 20 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-8: Hot Spot Cladding Temperature for 0.07 ft 2 Break 21 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Plots for 0.08 ft 2 Break Figure 10-9: Core Power for 0.08 ft 2 Break 22 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL K Figure 10-10: Inner Vessel Pressure for 0.08 ft 2 Break 23 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-11: Break Flow for 0.08 ft 2 Break 24 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 7 Figure 10-12: Core Inlet Flow Rate for 0.08 ft 2 Break 25 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-13: Hot Assembly Collapsed Liquid Level for 0.08 ft 2 Break 26 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-14: Hot Spot HTC for 0.08 ft 2 Break 27 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-15: Hot Spot Coolant Temperature for 0.08 ft 2 Break 28 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-16: Hot Spot Cladding Temperature for 0.08 ft 2 Break 29 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Plots for 0.09 ft 2 Break Figure 10-17: Core Power for 0.09 ft 2 Break 30 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENTTO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-18: Inner Vessel Pressure for 0.09 ft 2 Break 31 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-19: Break Flow for 0.09 ft 2 Break 32 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-20: Core Inlet Flow Rate for 0.09 ft 2 Break 33 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-21: Hot Assembly Collapsed Liquid Level for 0.09 ft 2 Break.34 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-22: Hot Spot HTC for 0.09 ft 2 Break 35 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-23: Hot Spot Coolant Temperature for 0.09 ft 2 Break 36 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-24: Hot Spot Cladding Temperature for 0.09 ft 2 Break 37 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-25: Hot Assembly Two-Phase Level for 0.07 ft' Break 38 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-26: Hot Assembly Two-Phase Level for 0.08 ft 2 Break 39 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 10-27: Hot Assembly Two-Phase Level for 0.09 ft 2 Break Ouestion 11: An HPSI flow delivery curve with 5% more flow was used in the AREVA analysis compared to the previous SBLOCA submittal by CE. Please justify this new HPSI delivery curve and demonstrate that it meets the latest surveillance measurement for HPSI pressure and flow. The HPSI curve is adjusted to account for measurement error (approximately 5%) for pressure and flow when the surveillance pressure/flow measurements are taken. Please demonstrate that these errors are accounted for in the HPSI pressure and flows used in the re-analysis. Please provide the HPSI pressure vs flow curve used in the analysis.CCNPP Response 11: The high pressure safety injection (HPSI) pump curve used in the current small break LOCA analysis performed by Westinghouse is the curve presented in UFSAR Revision 38, Figure 14.17-11, multiplied 40 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL by 0.95. That curve is presented in UFSAR Revision 37, Figure 14.17-11. Westinghouse applied the 5 percent reduction in flow to provide analysis margin.The HPSI pump curve used in the AREVA analysis is that presented in UFSAR Revision 38, Figure 14.17-11., The curve used by AREVA is supported by the plant surveillance test data. Margin in the surveillance test procedure acceptance criteria account for uncertainties in the flow measurement test.The pump delivery curve used in the small break LOCA analysis assumes that the break occurs in the cold leg with the highest safety injection flow and that flow is lost through the break. The pump curves are generated by inputting the most adverse non-degraded surveillance test data or vendor pump data into a computer model to obtain a conservative minimum prediction of pump flow. Check valve and relief valve leakages are modeled as outflows in the computer model. The computer generated curves are then degraded first for mechanical degradation and then for frequency degradation. Ouestion 12: It was stated that the AREVA SBLOCA analysis results in multiple loop seals clearing relative to the previously approved CE analysis where only one loop seal clears. Please provide the plots of the liquid levels in the loop seals for the limiting break and the steam mass flow and velocity entering the loop seal from the horizontal portions of the suction legs. Please show that the conditions in the loop seals in the unbroken loop support clearing of this, additional loop. Loop seal clearing phenomena following SBLOCA is very difficult to predict correctly and has historically been poorly predicted by all T/H codes, including the RELAP5 series of codes. As such, loop seal clearing only in the broken loop has been the accepted approach by the NRC staff during the review of evaluation models. Please provide benchmarking of the S-RELAP5 model against loop seal clearing separate effects tests as well as integral experimental data. While it is recognized that AREVA is using an approved RELAP5 model, it is still necessary to demonstrate that the model is performing correctly in all plant specific calculations, with a physically based thermal hydraulic behavior that supports the loop seal clearing behavior. It is not clear that additional loop seals will clear once the broken loop seal has cleared for such small break sizes.Please provide justification for the RELAP5 multiple loop seal clearing following a small break in the discharge leg. As a comparison, please provide the results of the limiting SBLOCA with only the single broken cold leg loop seal cleared.CCNPP Response 12: Response to be provided later.Ouestion 13: An analysis of the severed injection leg is needed. Please provide the results of an analysis of the severed injection line with the degraded injection into the RCS since one of the line spills to containment while the others inject at the much higher RCS pressures. Breaks up to and including approximately 1.0 ft 2 are considered in the small break spectrum even though they are in the transition region.CCNPP Response 13: Response to be provided later.Ouestion 14: Please perform an analysis of hot leg breaks to demonstrate that the limiting break location for the RCP trip timing criteria has been identified. 41 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL CCNPP Response 14: Response to be provided later.Question 15: Withdrawn CCNPP Response 15: N/A Question 16: In response to RAI#24, Constellation described the regulatory basis for removing Surveillance Requirement 3.2.1.1, but did not address the technical basis. During the audit, please be prepared to describe (1) how "peripheral" ASI will be used to confirm "interior" linear heat rate limits (TS 3.2.1), (2) will peripheral ASI (Fd) be combined with another local power surveillance (e.g., Fr) to validate peak LHGR (Fq)?, and (3) how does the removal of Fy surveillance guarantee the same level ofprotection? CCNPP Response 16: AREVA's setpoint methods use the three-dimensional neutronics code, PRISM, to calculate axial shapes and peaking factors. Limiting Fq peaking factors are calculated directly during a series of pre-determined core maneuvers that bound the allowed power versus peripheral axial shape index operating space. Direct calculation of the limiting Fq eliminates the need to make inferences about the amount of planar radial peaking, Fxy, that occurs in a particular plane and to synthesize Fq using F, and Fxy.During normal operations, the incore detector signals are used for continuous monitoring of peak linear heat generation rate (kw/ft) by using the alarm setpoints installed on the plant process computer. When operating using excore detectors (i.e., the incore detector system is out of service), peripheral axial shape index is monitored to maintain the plant within the power dependent axial shape index limits defined by the linear heat generation rate limiting condition for operation barn in COLR Figure 3.2.1-2. The linear heat generation rate limiting condition for operation limits are set to provide positive margin based on the cycle specific setpoint analysis.Question 17: Realistic LBLOCA During the audit, we'll need to speak to individuals familiar with the RLB LOCA analysis. The staff doesn't fully understand CCNPP's disposition and RAI response concerning the single failure selection. ANP-2834(P), Section 1, states that "A conservative loss of a diesel assumption is applied in which LPSI inject into the broken loop and one intact loop and HPSI inject into all four loops." CCNPP response to RAI 2 (letter dated August 9, 2010), as well as Section 3.1 of ANP-2834(P) state that the limiting single failure has been determined to be the loss of one ECCS pumped injection train. The staff needs to confirm that these failures are one in the same.The staff also needs to understand how the limiting single failure for the CE NSSS was determined, since the basis for the RAI response defers to NRC-approved methodology. Poring through EMF-2103, the staff only located sensitivity results on 3-loop W systems. In some cases, the limiting failure would be a 42 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL single LPSI and in others it was a diesel. The staff could not locate a clear, generic disposition for the single failure at anyplace in EMF-2103.What was done under the auspices of EMF-2103 development to ensure that the containment analysis produced a sufficiently conservative prediction that a no failure, max SI spillage case, for a CE NSSS, is bounded by the chosen single failure? The staff will need to see that work.CCNPP Response 17: The wording in ANP-2834(P) (Reference

3) is correct and ANP-2834Q(P) (Reference

4) should have stated "loss of diesel with fully functional containment sprays" when discussing the limiting single failure. The definition for loss of a diesel generator scenario by itself would mean that in addition to loss of one LPSI and one HPSI pump, one train of containment spray would not be available.

The current method models all containment pressure-reducing systems as fully functional. Containment fans start at time zero and containment sprays have a 20 second delay (Table 3-8, Reference 3).The response to RAI #111 for EMF-2103 (Reference 5, Attachment 1 page 185 -189) was based on sensitivities to 3-loop Westinghouse plants. The base case, which produced the most limiting results, is described in the RAI #111 response as the loss of one diesel with full containment spray. Figure 17-1 (recreated from RAI #111, Figure 111.2) shows that for the sample plant analysis (Westinghouse 3-loop, the base case, AREVA ECCS failure assumptions), is 35°F higher in peak clad temperature than a fully consistent loss of one diesel generator case and over 170'F greater than the loss of one LPSI pump case.43 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL E Figure 17-1: Clad Temperature Response from Single Failure Study 44 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL A sensitivity study of the limiting case for Calvert Cliffs (case #5) from the analysis of record was conducted with "maximum" ECCS flow conditions to demonstrate that the minimum ECCS single failure assumption is conservatively bounding.Sensitivity studies were run for the limiting case (Case #5) in both offsite power configurations with a maximum ECCS delivery. The loss of offsite power (LOOP) case for the max ECCS configuration had a peak clad temperature of 1656°F compared to the analysis of record LOOP with minimum ECCS, which had a peak clad temperature of 1670'F. The no loss of offsite power (NOLOOP) case for the maximum ECCS configuration had a peak clad temperature of 1597'F compared to analysis of record NOLOOP case with minimum ECCS, which had a peak clad temperature of 1609'F. This demonstrates that the AREVA single failure assumption produces conservative results. Figures 17-2 through 17-5 show the respective peak clad temperature trace, containment and system pressure, ECCS injection rates, and downcomer level for both the analysis of record and the maximum ECCS sensitivity. Figure 17-3 demonstrates that the maximum ECCS flow does not have a significant impact on the containment pressure up to about 75 seconds (approximately the time that the safety injection tank empties); the maximum ECCS containment pressure overlaps the analysis of record containment pressure.Figure 17-5 gives the downcomer level for both the analysis of record and the maximum ECCS case. It can be seen that the downcomer level in the maximum ECCS case is higher than the analysis of record, consequently providing more driving head for reflooding the reactor core. The higher driving head in the maximum ECCS case is enough to compensate for small differences in containment pressure (Figure 17-3) resulting in a faster post-peak cooldown.The AREVA RLB LOCA application, regardless of the loss of diesel assumption, models all containment pressure-reducing systems and conservatively assumes them to be fully functional. The analysis of record conservatively assumes an on-time start and normal lineups of the containment spray and fan coolers to conservatively reduce containment pressure and increase break flow. For the ECCS injection, the analysis of record also assumes, inconsistently with the containment assumption, a loss of a diesel generator (one train of pumped ECCS is lost), and the ECCS that is available to the RCS, as a consequence of the Combustion Engineering configuration, consists of one LPSI pump that injects into the broken loop and an intact loop, and one HPSI pump that injects into all four loops. In comparison to the analysis of record, the maximum ECCS sensitivity study simulates two trains of pumped injection -HPSI flow to each of the loops is doubled and LPSI flow (a LPSI pump is directed only to two cold legs)to each loop is simulated. The results of the study demonstrate that the analysis of record ECCS configuration is peak clad temperature -limiting and oxidation-limiting. 45 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 17-2: Comparison of PCT Independent of Elevation for Max ECCS versus Min ECCS 46 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 17-3: Comparison of Containment and System Pressure for Max ECCS versus Min ECCS 47 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 17-4: Comparison of ECCS Flows for Max ECCS versus Min ECCS 48 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 17-5: Downcomer Level Question 18: The response to RAI 23. 0 discussed the 7-minute operator action that is credited to secure the reactor coolant pumps. Please address the following:

a. Will operators need to know that there is a time-constraint of 7 minutes associated with this action?b. How have these actions been validated to be feasible and reliable?c. Who was, or will be, involved in the validation?

I 49 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL d. Describe the changes, if any, to the plant-reference simulator and training that are planned to support these actions.e. The response to the RAI states, "The new step is in the Pressure and Inventory Safety Function, normally the second safety function performed by the Reactor Operator, and normally within two minutes of a reactor trip. " Where did the two minute time frame originate and what is its basis? Is this from another procedure that has previously been validated? CCNPP Response 18: a. Yes, the Operators will be given the 7 minute basis. A requirement of 6 minutes is included in the draft EOP-0 bases. Training the operators on the basis will ensure more reliable performance.

b. The 7 minute time frame has been validated during a number of simulator runs to be both feasible and reliable.

These simulator runs were performed by staff Licensed Operators. As shown below, trips of the RCPs occurred in all cases well within 7 minutes.Baseline Run: Plant response, with no operator actions, to a 1 0,000gpm leak (estimated leak rate for 0.1 ft 2 slot break, smallest leak of concern for a peak centerline temperature >2200'F). A smaller leak is limiting for operator actions since it delays their cues for securing an RCP.Time Action 0000:00 RCS leak begins 0000:13 Automatic reactor trip -low RCS pressure 0000:18 SIAS (setpoint 1740 psia RCS pressure) (EOP-0 requires tripping 2 RCPs)0000:23 1600 psia RCS pressure (pressure considered in AREVA calcs -begin 7 minute clock to trip ALL RCPs)0000:29 CIS (2.8 psig containment pressure) (EOP-0 requires tripping ALL RCP's)0000:54 RCP pump curves violated (RCS pressure 1200psia -assumed 532F and 2 RCPs running) (EOP-0 requires tripping ALL RCPs)0000:59 <20'F Subcooling (RCP trip criteria to protect AREVA fuel)0st run: 10,000 gpm RCS leak, performed by Staff SRO #1, no knowledge of purpose of exercise Time Action 0000:00 RCS leak begins 0000:10 Manual Rx Trip due to TMILP pretrips not recovering 0001:00 Subcooling less than 20'F, CIS and RCP curves not met 0001:10 Reactivity Control block steps complete, no alternate actions taken, begin Pressure and Inventory Control block steps 0002:10 RCP's secured on low subcooling and violation of RCP curves, time includes alternate action for verifying SIAS with no anomalies 50 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 2nd run: 10,000 gpm RCS leak, performed by Staff SRO #2, no knowledge of purpose of exercise Time Action 0000:00 RCS leak begins 0000:13 Auto Rx Trip due to TMILP pretrips not recovering 0000:50 Reactivity Control block steps complete, begin Pressure and Inventory Control block steps 0001:00 Subcooling less than 20'F, CIS and RCP curves not met 0001:25 RCP's secured on low subcooling and violation of RCP curves, time includes alternate action for verifying SIAS with no anomalies 3 run: 10,000 gpm RCS leak, performed by shift RO #1-, no knowledge of purpose of exercise Time Action 0000:00 RCS leak begins 0000:13 Auto Rx Trip due to TM/LP pretrips not recovering 0000:40 Reactivity Control block steps complete, begin Pressure and Inventory Control block steps 0001:00 Subcooling less than 20°F,' CIS and RCP curves not met 0001:40 RCP's secured onCIS, time includes alternate action for verifying SIAS and CIS with no anomalies Notes: The operator actions are performed by a Reactor Operator, using EOP-0, "Post Rx Trip Actions".During these simulator runs all required reports and verifications with crew were simulated. For simulator runs 2 and 3, no actions were performed until an automatic reactor trip occurs (this delays the operator's response). The times for the plant response from the baseline run apply to simulator runs 1 through 3, but are not listed for clarity of operator actions.All timelines begin at 0000:00 from the RCS leak initiation. The operator action requirement to meet 7 minutes for tripping all RCPs does not begin until time +0000:23 (RCS pressure <1600psia) in simulator runs 1-3.The requirement to trip RCP's on low RCS subcooling is an additional requirement to an existing step, and therefore not a complex change to operator actions. The RCS subcooling parameter is clearly indicated to the Operators, and the Operators are already sensitive to the subcooling parameter as an indication of compliance with the existing RCP pump curve trip criteria in EOP-0.Upon completion of Reactivity Control, and Pressure and Inventory Control block steps, the Reactor Operator would then move on to the Containment Environment block step, which requires tripping all RCPs on a valid CIS. The CRO performs the Turbine Trip block step first, then moves on to the Vital Auxiliaries block step, which requires tripping all RCPs on a valid CIS. The Reactor Operator is cued three times to trip all RCPs in the Pressure and Inventory Control block step and then again (4 th time) in the Containment Environment block step. The CRO actions back up the Reactor 51 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Operator actions for CIS trip in the Vital Auxiliaries block step action. Note that the Control Room Supervisor is responsible to verify proper responses by Reactor Operator and the CRO.Conclusions: Reactor Operators can reliably demonstrate that all RCPs would be secured in less than seven minutes from the time the RCS pressure drops below 1600 psia. Three separate operators all achieved times of 2 minutes and 10 seconds or less. Therefore this conclusion is repeatable and valid. In addition, there are three separate cues to an operator to trip RCPs in the Pressure and Inventory Control block step. Therefore this action is reliable. Note that the Reactor Operator and the CRO each have additional block steps that require tripping RCPs on CIS, this adds to the reliability of the function.c. See CCNPP Response 18.b.d. Since the action to trip the RCPs exists in EOP-0, and the subcooled indication exists in the simulator, no changes to the simulator are required. The Licensed Operators will be trained prior to the 2011 RFO for the changes to EOP-0 and their bases. Continuing training is provided for EOP-0 ,and would use the latest revision of the procedure.

e. As noted in Response 18.b, the total time to trip the RCPs is about two minutes (or less) from the initial break in the RCS. This is the time referred to in Response 23.0 from Reference

1. The calculated time allowed for an RCS trip is seven minutes and starts when the pressurizer pressure reaches 1600 psia.Question 19: In response to RAI #9, the licensee provided Figures 9-4 and 9-6 illustrating steam flow versus time for a 1. Oft 2 break outside containment and inside containment respectively.

Both breaks are located upstream of the MSIV a. Explain the asymmetric steam flow prior to MSIV closure which is not exhibited in larger breaks.CCNPP Response 19: Figure 19-1 shows the nodalization of the steam lines for the asymmetric steam generator breaks inside containment [] For the asymmetric steam generator breaks outside containment, the configuration is essentially the same, but the break plane is moved from [] In the following discussion, this difference is insignificant, as both break configurations behave similarly. Breaks with an area [For the configuration shown in Figure 19-1, [J. The outlet of the intact or unaffected SG-1 [52 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL I When the break is opened, the flow [] The steam flows for a [are shown in Figure 19-2 and demonstrate this behavior.The choking behavior [[I 1] is shown in Figure 19-3.The increased load from SG-1 causes [] The temperatures of each side seen at the core inlet are shown in Figure 19-4. [I The modeling of the distribution of total kinetics power to the two side' of the core and the [[answer this question, the original modeling shown in [] is based on I To I As expected, [I This behavior is seen in Figure 19-5.l Figure 19-6 shows [The calculation of the kinetics power [The results of the calculation are plotted in Figure 19-7. [I I 53 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL I Figure 19-8 shows the calculated core power based [For these small breaks, [Figure 19-1: Asymmetric Break Steam Line Nodalization 54 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 19-2: Steam Flows for 0.5 ft 2 Asymmetric Break 1 Figure 19-3: Choking Behavior for 0.5 ft 2 Asymmetric Break 55 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 19-4: Core Inlet Temperatures for 0.5 ft 2 Asymmetric Break 7 Figure 19-5: Steam Flows for 0.5 ft Asymmetric Break with Swapped Break Location 56 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 19-6: Core inlet Temperatures for 0.5 ft 2 Asymmetric Break with Swapped Break Location Figure 19-7: Power Split for 0.5 fte Asymmetric Break 57 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 19-8: Total Core Power (Heat-Flux-Based) for 0.5 ft 2 Asymmetric Break Question 20: In response to RAI #10, the licensee described various credited trip functions. Please discuss the harsh environment uncertainty for the ASGT and Containment Pressure High trip functions. CCNPP Response 20: Response to be provided later.Question 21: Withdrawn CCNPP Response 21: N/A Question 22: For the PCT-limiting RLBLOCA case, please provide: a. Corrected and uncorrected radial temperature profile of the hot rod at the time and location ofpeak cladding temperature. b' Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average, and clad surface temperatures. Indicate the end of blowdown, start of refill, and start of reflood on this graph.c. Burnup for the limiting rod CCNPP Response 22: Figure 22-1 responds to Question 22a. The squares show the pellet radial temperature distribution for the most severe case after the centerline temperature has been corrected for the bias to fuel data and the sampled uncertainty. The uncorrected, but uncertainty adjusted, centerline temperature is provided by the triangle. Adjusting the pellet thermal conductivity does not impact the cladding temperature and has only 58 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL a minor impact on the pellet surface temperature. Therefore, the temperature distribution for the uncorrected distribution will follow that for the corrected distribution and converge gradually to the same pellet surface value.In response to Question 22b, the pellet and cladding temperature evolutions are provided in Figure 22-2.The burnup for this case was 25.7 GWdiMTU. This responds to Question 22c.Figure 22-1: Radial Temperature Profile for Hot Rod 59 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 22-2: Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Average Question 23: The issue described in IN 2009-23 invalidates AREVA's generic disposition for analyzing fresh fuel only, which is based on sensitivity studies indicating that mid-second-cycle fuel had a PCT of 80'F lower than the limiting PCT This work needs to be repeated accounting for fuel thermal conductivity degradation. Please provide several cases run at various times-in-life for once-burnt fuel, with information similar to the above list provided; Item 1.c is only necessary for the most limiting second-cycle case analyzed.CCNPP Response 23: The issue described in Information Notice 2009-23 does not invalidate AREVA's approach for analyzing fresh fuel only for RLB LOCA. Table 23-1 compares the fresh AREVA fuel radial peaking (Fr) to the burnt Westinghouse fuel radial peaking for Calvert Cliffs Unit-2 Cycle 19. The Fr for the fresh fuel is significantly higher under nominal hot full power conditions, which can be directly related to fuel centerline temperature. The correction factor for thermal conductivity degradation on RODEX3A is shown in Figure 23-1.60 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Table 23-1: Max Fr Fresh and Burned Fuel Max. Max. Fr Max. Max. Fr Cycle BU Fr () Fr (-) Margin Cycle BU Mx Max.(GWdJMTU) Fr ((F) (MaMargi (fresh) (burnt) (%) (fresh) (burnt) (%)0.15 1.541 1.336 13.3 12 1.524 1.306 14.3 0.5 1.535 1.336 13 12.5 1.523 1.299 14.7 1 1.533 1.337 12.8 13 1.532 1.292 15.7 1.5 1.532 1.341 12.5 13.5 1.537 1.285 16.4 2 1.531 1.343 12.3 14 1.54 1.279 16.9 2.5 1.528 1.346 11.9 14.5 1.54 1.274 17.3 3 1.526 1.348 11.7 15 1.536 1.27 17.3 3.5 1.527 1.35 11.6 15.5 1.529 1.267 17.1 4 1.534 1.351 11.9 16 1.519 1.264 16.8 4.5 1.541 1.351 12.3 16.5 1.509 1.261 16.4 5 1.545 1.35 12.6 17 1.498 1.258 16 5.5 1.545 1.348 12.8 17.5 1.488 1.255 15.7 6 1.544 1.346 12.8 18 1.477 1.252 15.2 6.5 1.54 1.342 12.9 18.5 1.466 1.249 14.8 7 1.536 1.339 12.8 19 1.456 1.245 14.5 7.5 1.532 1.336 12.8 19.5 1.447 1.242 14.2 8 1.527 1.334 12.6 20 1.437 1.239 13.8 8.5 1.523 1.332 12.5 20.5 1.43 1.236 13.6 9 1.518 1.329 12.5 21 1.424 1.233 13.4 9.5 1.518 1.327 12.6 21.5 1.419 1.23 13.3 10 1.522 1.324 13 22 1.413 1.227 13.2 10.5 1.525 1.321 13.4 22.1 1.412 1.227 13.1 11 1.526 1.317 13.7 22.5 1.407 1.225 12.9 11.5 1.526 1.312 14 23.09 1.4 1.221 12.8 61 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 23-1: Correction Factor for RODEX3A Adjustment Once burned fuel, typically, can reach burnups approaching to 50 GWd/MTU at the end of the second cycle. The initial fuel temperature for the realistic LOCA model is determined from the RODEX3A computer code with the fuel temperature bias adjustment shown in Figure 23-1. This bias adjustment is determined from experimental benchmarks for a burnup range that exceeds the allowed fuel operating range and implicitly includes the effect of thermal conductivity degradation. Table 23-1 provides the minimum relative power ratio between fresh AREVA fuel and the once burned Westinghouse fuel.Comparing the beginning of cycle, middle of cycle and end of cycle values for fuel temperature bias and power gives the following result in Table 23-2.Table 23-2: Time in Cycle Comparison Beginning of Cycle Middle of Cycle End of Cycle Fresh fuel burnup (GWd/MTU) 0 15 30 Once-burned fuel burnup (GWd/MTU) 25 37.5 50 Fuel thermal conductivity correction adjustment for once-burned fuel 7.7 6.1 4.8 compared to fresh fuel (%) I I Power reduction for once-burned fuel 13.3 14.0 12.8 compared to fresh fuel (%)Evaluating the effect on centerline fuel temperatures for each of these points in the cycle, the initial centerline fuel temperature for second cycle fuel will always be more than 183'F below that of the corresponding fresh fuel. With the initial fuel stored energy, fuel temperature, and fuel pin power lower, the peak cladding temperature for corresponding time in cycle evaluations will always occur in the fresh 62 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL fuel. This validates that the most severe LOCA conditions occur in fresh fuel for the fuel cycle management employed at Calvert Cliffs.Question 24: In RAI #18, the staff requested an explanation of apparent differences between the UFSAR AOR and the AREVA analysis. The response was insufficient to understand differences in analytical techniques, assumptions, and initial conditions. Please expand this response.a. Provide a comparison of Cycle 18 versus Cycle 19 calculated values for initial ASI, initial Fq, ejected rod worth, post ASI, and post Fqfor the limiting rod configuration at HZP and HFP, and b. Using prior reload cycle calculated parameters (ejected worth, post- Fq, fuel enthalpy), populate the plots of Deposited Enthalpy vs. Rod Worth in XN-NF- 78-44.CCNPP Response 24: Response to be provided later.Question 25: Withdrawn CCNPP Response 25: N/A Question 26: In RAI #14, the staff requested an explanation of the use of BOC and EOC physics parameters. The response (provided in RAI #3) was insufficient tojustify analyzing only these two extreme cases. At EOC conditions, FTC is most negative (turns event around) and Beff is smallest (promotes power excursion). These parameters act to offset each other, so the limiting scenario may be at another BUpoint where FTC is not so negative. This is probably the reason why the UFSAR AOR combines limiting parameters, regardless of exposure dependence. CCNPP Response 26: Response to be provided later.Question 27: With respect to the performance of co-resident Westinghouse fuel assemblies:

a. Please explain how the rod power histories considered in the Westinghouse fuel rod thermal-mechanical performance calculations will be verified for future operating cycles.b. Did the Post-Trip MSLB analysis consider the performance of Westinghouse fuel should the stuck* CEA be located above a Westinghouse assembly?CCNPP Response 27: a. Westinghouse calculations were performed for Unit 2 Cycle 19, the first transition core. Data was generated using Westinghouse codes and methods and using the Unit 2 Cycle 19 core loading pattern and-operating constraints.

The fuel performance evaluation showed that Unit 2 Cycle 19 is bounded 63 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL by the current Westinghouse analysis of record. The same process is to be performed for future cores that contain Westinghouse fuel.b. The most reactive CEA location was modeled in the AREVA analysis. None of the most reactive CEAs are located in Westinghouse fuel. Less reactive CEAs (Bank 5) are inserted into Westinghouse fuel and 2 of a ganged Bank C CEA location is inserted into Westinghouse fuel.The AREVA Steam Line Break analysis contains a stuck CEA sector that includes 22 fuel assemblies, some of which are Westinghouse fuel. The fraction of power generated in the stuck CEA sector is based upon a fit of neutronics data, thus the separation between AREVA and Westinghouse fuel is not explicitly modeled in the S-RELAP5 calculation. The DNB and fuel centerline melt calculation for the Steam Line Break event assessed the Fr penalty at the Steam Line Break conditions and determined that the peak Fr in the Westinghouse fuel remained 9 percent less than that in the AREVA fuel. The AREVA post-trip Steam Line Break analysis did not result in fuel failure in the AREVA fuel. Since no fuel failure occurred in the AREVA fuel and the Fr in the Westinghouse fuel remained more than 9 percent lower than that in the AREVA fuel, no fuel failure is predicted in the Westinghouse fuel.Question 28: The staff has completed FRAPCON-3 benchmark calculations which challenge ARE VA calculations.

a. Using the power histories in Figure 2 of 32-9135500-001, best-estimate FRAPCON-3 calculated rod internal pressures are 650 psia and 600 psia above the "worst-case" AREVA calculations for U02 and UGdO2 fuel rod designs.b. Accounting for modeling uncertainty and worst case manufacturing tolerances, FRAPCON-3 calculated rod internal pressures are 1300 psia and 1000 psia above the "worst-case" ARE VA calculations for U02 and UGdO2 fuel rod designs.CCNPP Response 28: Response to be provided later.Question 29: As part of a recent Fitzpatrick fuel transition, the staff developed penalties on calculated rod internal pressure and limitations on fuel rod burnup to address outdated fuel thermal-mechanical methods. The use of cycle-specific rod power histories in the AREVA RODEX-2 methodology makes this difficult.

CCNPP Response 29: The discussion below provides a better understanding of the AREVA effort to support more recent hot cell data that result in different fuel thermal conductivity parameters. Fuel centerline melt limits are calculated using automation tools which run RODEX2 to calculate the cycle specific fuel centerline melt limit. RODEX2 does not directly model the bumup dependent thermal conductivity. To compensate for this, other models were adjusted to benchmark to the high burnup data that was available when the code was approved. However, more recent test data indicated that those adjustments were not sufficient to offset the bumup dependent degradation in the fuel thermal conductivity. It appears that RODEX2 predicts conservatively high temperatures at low bumups 64 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL[ ], but at higher exposures under predicts pellet temperatures. This may result in negative impacts on fuel performance. The impact of this issue on pressurized water reactor fuel centerline melt temperatures was examined and a penalty factor on RODEX2 fuel centerline melt temperature (as a function of rod average burnups) was developed through a code-to-code comparison with COPERNIC.The fuel centerline melt limits as functions of rod average burnup for the Combustion Engineering reference plant are presented in Figure 29-1 to Figure 29-5. Based on these results, penalties were conservatively selected to bound the data. The penalties are reported as a function of rod average burnup and they are applied to the RODEX-2 results.Figure 29-1: CE Reference Plant -U0 2 Results 65 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 29-2: CE Reference Plant -2 wt% Gadolinium Results Figure 29-3: CE Reference Plant -4 wt% Gadolinium Results 66 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 29-4: CE Reference Plant -6 wt% Gadolinium Results 67 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Figure 29-5: CE Reference Plant -8 wt% Gadolinium Results Question 30: Withdrawn CCNPP Response 30: N/A Ouestion 31: a. AREVA postulates that clad swelling and rupture produces a benefit to PCT, and because of this, the realistic large break loss of coolant accident (RLBLOCA) model does not include a clad swelling and rupture model. Does this conjecture include consideration of test data, which has shown that following fuel rupture, the ballooned region fills with fuel fragments? What analytic studies support this conclusion? How are they applicable to CCNPP? Please also address the potential for co-planar blockage with the fuel relocation evaluation.

b. Since blowdown ruptures can occur at end of life conditions, show that blowdown ruptures do not occur at the end of life for the postulated CCNPP large break LOCA.68 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL CCNPP Response 31: Justification for this position is provided by consideration of the phenomena involved, analysis of the effects, and experimental results. The impact of rupture and ballooning on clad cooling occurs through several rupture- or ballooning-induced cooling mechanisms and three detrimental heating effects: Cooling effects: Increased clad heat transfer surface area at ballooned or ruptured regions Increased velocities within the ballooned and ruptured regions Increased turbulence within the ballooned and ruptured regions Droplet shattering resulting in increased interphase heat transfer and steam de-superheating Decrease in gap heat transfer if the fuel does not strongly relocate Decrease in pellet thermal conductivity if the fuel relocates Formation of local quench fronts in ballooned and ruptured regions 69

-' ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL Heating effects: Flow diversion around ballooned and ruptured regions Clad heating load increase at the ruptured elevation due to fuel relocation Cladding heat load increased due to interior oxidation Experience with Appendix K methodologies has shown that the aggregate of these effects acts to decrease the cladding temperatures when no fuel relocation occurs. This was demonstrated in Appendix B, Section B.2 of RLB LOCA evaluation model topical report (Reference

6) and the response to RAI 28 on the topical (Reference

5) with sensitivity studies on both 3- and 4-loop pressurized water reactors with 15xl 5 and 17x17 fuel designs similar to the 14x14 fuel designed for use at Calvert Cliffs. The studies included increased heat transfer surface area, increased local coolant velocities, a decrease in gap heat transfer, flow diversion, and interior cladding oxidation.

The effects of increased turbulence, droplet shattering, and potential local quenching were not included within the modeling. Decrease in pellet thermal conductivity and a clad heating load increase also were not included since the studies were not meant to address fuel relocation. Even without half of the cooling mechanisms modeled, the cladding temperatures and local oxidations were reduced. This effect has also been observed experimentally in the FEBA (Reference

7) and FLECHT (Reference

8) test series.Under a condition of fuel relocation, wherein the fuel above the ballooned region drops into the ballooned region, it has been postulated that increased decay heat generation will lead to an increase in cladding heat flux resulting in higher cladding temperatures.

Various presentations (e.g., Reference 9 Articles 1 and 12)purport to show the effect. However, these studies have uniformly incorporated extreme assumptions on the conditions of relocation and the resultant heat transfer processes. Few include provisions for rupture-induced cooling mechanisms. Most assume that the cladding expands circularly without being encumbered by the surrounding pins in the fuel assembly. In fact, a free expansion of the fuel rod is only possible up to pin strains in the mid-30 percents. For higher strains the local gap volume no longer increases faster than the clad surface area. Finally, the packing factor of the rubble filling the ballooned region is over predicted. If reasonable, yet conservative, assumptions are made, study results would lead to the expectation that fuel relocation, which is real, does not pose a condition by which the ruptured or ballooned regions will exceed the consequence of the non-ballooned regions of the hot pin.The above conclusion was observed experimentally in the KfK experiments as reported in RAI 131 (Reference 5). In the KfK in-pile tests, fuel relocation into the ballooned area of the fuel rod occurred but did not adversely affect the subsequent clad temperature behavior. To determine when the fuel relocates, two tests were performed with thermocouples located at the top of the pellet stack. One test comprised low burnup fuel, which maintained its pellet geometry after rupture. The other test was of higher burnup fuel which relocated. Relocation, for the test that relocated, was demonstrated by temperatures from the upper thermocouples showing a significant drop, loss of energy source, at the time of fuel rod rupture.For this test, the heatup rate, at the rupture elevation, following the rupture was reduced relative to the heatup rate prior to rupture. This reduction in heatup rate indicates that the peak clad temperature at the time of turnover would be less than what would have been reached if rupture had not occurred, even with the increase in localized decay heat from the pellet rubble residing at the ruptured region. Thus, the KfK experiments demonstrate that analyses which ignore the beneficial effects of swelling and rupture provide conservatively high clad temperature estimates for the ruptured region during reflood even when fuel relocation occurs.70 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL In conclusion, the AREVA RLB LOCA evaluation model does not incorporate a clad ballooning, rupture and fuel relocation model. To support this modeling, the cladding temperature and pin stress evolution for individual cases in the case set has been assessed against rupture criteria appropriate for the cladding being evaluated. No rupture occurred during blowdown or refill for fresh or once- and twice-burned fuel.For rupture during reflood, the cladding temperature for the most severe location on the un-ruptured fuel rod has been demonstrated to conservatively bound the result for any possible rupture location.Question 32: Provide information to illustrate the conservative nature of the single-side only oxidation model and its application to the CCNPP RLBLOCA analysis.CCNPP Response 32: AREVA's NRC-approved realistic large break LOCA (RLB LOCA) evaluation model uses the maximum un-ruptured cladding oxidation as representative or bounding of the oxidation that would have been computed' at a rupture location. The position is supported by three aspects of the performed oxidation calculation.

1. The cladding is initialized with no initial corrosion layer. Because the oxidation rate is inversely proportional to the oxidation layer present, the use of clean cladding at the start of the accident leads to substantially higher reaction rates. For corrosions in the range of the first cycle of M5 cladding, the difference in rate is a minimum of a 50 percent increase and increases during the cycle. The increase applies to both exterior and post-rupture interior oxidation.

2. The cladding temperature even in the presence of fuel relocation is reduced for the ruptured region of the cladding.

In the KfK experiments (Reference

7) cited in the response to Question 31 (above)the temperature drop at rupture was between 50 and 75 K. Since the oxidation rate is exponentially proportional to the cladding temperature, a drop of 50 to 75 K for Calvert Cliffs provides an oxidation rate reduction of 50 percent or more.71 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 3. For ruptured cladding either the cladding interior oxidation rate is reduced by attached pellet fragments, moderate to highly burned fuel, or the cladding temperature decrease at rupture is much more than the 50 to 75 K explained above. In either case, an additional mechanism exists to reduce the local oxidation at the rupture location.In conclusion, insights into the evaluation model oxidation process and those that will evolve after rupture clearly identify differences that will reduce the oxidation at the rupture location to less than that which the evaluation model calculates at un-ruptured locations.

Thus, the RLB LOCA evaluation model approach to reporting local oxidation is appropriate to demonstrate compliance with the local oxidation criterion of 10 CFR 50.46.Ouestion 33: Provide additional information to justify the use of the selected analytic treatment for decay heat uncertainty'in the RLBLOCA model.CCNPP Response 33: For the realistic large break LOCA (RLB LOCA) analysis performed for Calvert Cliffs, the decay heat calculations are based on Reference

14. The standard is applicable to light water reactors containing low enriched uranium as the initial fissile material; all plants to which the RLB LOCA evaluation model is applicable, including Calvert Cliffs, are such plants. The selected approach to simulate fission product decay assures a representative yet conservative treatment.

The evaluation model fission product decay heat uncertainty and the basis for the conservatism of the approach are described below.Bias and Uncertainty The RLB LOCA methodology utilizes the U-235 decay curve from Reference 14 for fully saturated decay chains as the decay for all fission products. The fully saturated chains result from an assumption of infinite operation. The total energy per fission is assumed to be 200 MeV (Reference 14). [Conservatism in the Approach The RLB LOCA methodology utilizes the U-235 decay curve from Reference 14 for fully saturated decay chains. The fully saturated chains result from the assumption of infinite operation. The total energy per fission is assumed to be 200 MeV (Reference

14) whereas a more accurate value for U-235 would be greater than 202 MeV per fission. The use of only U-235 and fully saturated chains provide conservatism in the overall fission product decay heat, independent of uncertainties.

During irradiation, plutonium accumulates such that the ratio of plutonium-to-uranium fission-energy production rate is substantial and increasing. Because the decay energy resulting from plutonium fissions 72 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL is less than that from uranium, the decay energy based on the assumption of fully saturated decay chains, is reduced as the fuel is burned. Thus, as burnup increases, the RLB LOCA decay heat modeling (U-235 only) accrues conservatism. The conservatism applies to all regions of the core according to the mix of burnups represented within each region.The fresh fuel, hot pin and hot assembly, begin operation with no plutonium. Therefore, the reduction in decay heat due to plutonium build-up is not applicable to the low burnup fuel in the initial period of the cycle. However, for fresh fuel, the concentrations of long decay term fission products will not have built up. The lack of long decay term sources comprises a reduction in decay heat rate of several percent over the first 500 days of operation, making the infinite operation assumption conservative while the plutonium concentration is accumulating. A further consideration relative to long and intermediate decay term sources is the plant time duration at the analyzed peaking factor prior to the accident. The RLB LOCA analysis is performed with hot pin peaking sampled linearly between a high expectation for normal operation and the limiting conditions of operation. Although there are many sampled variables in the evaluation, the severe or challenging results will occur for cases that sample the peaking near the limiting condition for operation. When a plant experiences peaking that approaches the limiting condition for operation limits, plant operators will take corrective actions to return the plant to a normal condition. This action severely limits the buildup of long and mid-decay term sources; creating a condition where, in a true best estimate evaluation, decay heat would be based on normal operational peaking for long and mid-decay term sources and limiting condition for operation peaking only for the shorter decay term sources. Again, the assumption of infinite operation at the sampled peaking is conservative by several percent.In conclusion, the choice of infinite operation with pure U-235 fission product decay heat provides a base model that is conservative. Sampling this model based on the uncertainty of the U-235 decay chain provides a realistic treatment subject to the conservatism imbedded in the approach.REFERENCES

1. Letter from Mr. G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated November 19, 2010, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel 2. Letter from Mr. G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated October 29, 2010, Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel 3. ANP 103-2834(P)-000, "Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report," September 2009 4. ANP 103-2834Q(P)-000, "Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report," July 2010 5. AREVA Letter NRC:02:062, December 20, 2002, Responses to a Request for Additional Information on EMF-2103(P)

Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (TAC No. MB2865)6. EMF-2103(P)(A) Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors", April 2003 7. P. Ihle, Heat Transfer in Rod Bundles with Severe Clad Deformations, KfK 3607 B, April 1984 73 ATTACHMENT (3)NON-PROPRIETARY SUPPLEMENT TO LICENSE AMENDMENT REQUEST: TRANSITION TO AREVA NUCLEAR FUEL 8. M.J. Loftus, et al., PWR FLECHT SEASET 163-Rod Bundle Flow Blockage: Task Data Report, No. 13, NUJREG/CR-3314, October 1983 9. NEA/CSNI/R(2004)19, SEGFSM Topical Meeting on LOCA Fuel Issues, Argonne National Laboratory, May 25-26 2004, Published by Organization for Economic Cooperation and Development Nuclear Energy Agency, Isy-les-Moulineaux, France, November 2004 10. NUREG/CR-0103, ORNL/NUREG/TM-200, Multirod Burst Test Program Progress Report for July -December 1997, US Nuclear Regulatory Commission, Washington, DC 11. NUREG/CR-0655, ORNL/NUREG/TM-297, Multirod Burst Test Program Progress Report for July -December 1998, US Nuclear Regulatory Commission, Washington, DC 12. NUREG/CR-1023, ORNL/NUREG/TM-351, Multirod Burst Test Program Progress Report for April -June 1999, US Nuclear Regulatory Commission, Washington, DC 13. K. Wieher and U. Harten, Datenbericht REBEKA-6, KfK 3986, March 1986 14. ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, approved August 29, 1979 74}}