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{{Adams
#REDIRECT [[RS-16-018, Response to Request for Additional Information Regarding Request for License Amendment Regarding Transition to Areva Fuel]]
| number = ML16028A303
| issue date = 01/28/2016
| title = Dresden and Quad Cities - Response to Request for Additional Information Regarding Request for License Amendment Regarding Transition to Areva Fuel
| author name = Simpson P R
| author affiliation = Exelon Generation Co, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRO
| docket = 05000237, 05000249, 05000254, 05000265
| license number = DPR-019, DPR-025, DPR-029, DPR-030
| contact person =
| case reference number = RS-16-018, TAC MF5736, TAC MF5737, TAC MF5738, TAC MF5739
| document report number = ANP-3463NP, Rev. 0
| package number = ML16028A302
| document type = Legal-Affidavit, Letter, Response to Request for Additional Information (RAI), Topical Report
| page count = 76
| project = TAC:MF5736, TAC:MF5737, TAC:MF5738, TAC:MF5739
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:4301 W1P r c R'l it W 111 c1 ville GO'> >"-Exelon Generation G.30 G')l 2000 Uflc , PROPRIETARY INFORMATION-WITHHOLD UNDER 10 CFR 2.390 RS-16-018 10 CFR 50.90 January 28, 2016 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001
 
==Subject:==
Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRG Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRG Docket Nos. 50-254 and 50-265 Response to Request for Additional Information Regarding Request for License Amendment Regarding Transition to AREVA Fuel
 
==References:==
( 1) Letter from Patrick A. Simpson (Exelon Generation Company, LLC) to U.S. NRG, "Request for License Amendment Regarding Transition to AREVA Fuel," dated February 6, 2015 (2) Letter from U.S. NRG to Bryan Hanson (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 -Request for Additional Information Related to the License Amendment Request to Transition to AREVA Fuel (CAC Nos. MF5736, MF5737, MF5738, and MF5739)," dated January 19, 2016 In Reference 1, Exelon Generation Company, LLC (EGG) requested an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS) Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. The proposed change supports the transition from Westinghouse SVEA-96 Optima2 (Optima2) fuel to AREVA ATRIUM 10XM fuel at DNPS and QCNPS. Specifically, EGG proposes to revise Technical Specification (TS) 5.6.5, "Core Operating Limits Report (COLA)," paragraph b, to delete no longer used methodologies and to add the AREVA analysis methodologies to the list of approved methods to be used in determining the core operating limits in the COLA. Also, in support of the planned transition to AREVA ATRIUM 10XM fuel, EGG proposes to revise DNPS and QCNPS TS 3.2.3, "Linear Heat Generation Rate (LHGR)," and TS 3.7.7, "The Main Turbine Bypass System." Attachment 2 transmitted herewith contains Proprietary Information.
When separated from attachments, this document is decontrolled.
January 28, 2016 U. S. Nuclear Regulatory Commission Page2 In Reference 2, the NRC requested that EGC provide additional information to support their review of the subject amendment request (i.e., Reference 1). The requested information is provided in the attachments to this letter. The following attachments are included in support of this RAI response:
Attachment 1: Response to Request for Additional Information Attachment 2: AREVA, Inc. Document No. ANP-3463P, "Response to RAl's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel" (Proprietary)
Attachment 3: AREVA, Inc. Affidavit Requesting Proprietary Report be Withheld from Public Disclosure Attachment 4: AREVA, Inc. Document No. ANP-3463NP, "Response to RAl's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel" (Non-Proprietary)
Attachment 2 contains information proprietary to AREVA, Inc. As Attachment 2 contains information proprietary to AREVA, Inc., this document is supported by an affidavit (i.e., Attachment
: 3) signed by AREVA, Inc., the owner of the information.
The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information which is proprietary to AREVA, Inc. be withheld from public disclosure.
A proprietary version of Attachment 2 is provided in Attachment
: 4. In a January 20, 2016 teleconference between Russell Haskell (U.S. NRC) and Timothy Byam (EGC), it was agreed that EGC would provide the requested information to the NRC on or before January 29, 2016. In accordance with 1 O CFR 50.91 (b), EGC is notifying the State of Illinois of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.
EGC has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in Attachment 1 of Reference
: 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.
In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2818.
Attachment 2 transmitted herewith contains Proprietary Information.
When separated from attachments, this document is decontrolled.
January 28, 2016 U. S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of January 2016. Respectfully, Patrick R. Simpson Manager -Licensing Attachments:
: 1. Response to Request for Additional Information
: 2. AREVA, Inc. Document No. ANP-3463P, "Response to RAl's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel" (Proprietary)
: 3. AREVA, Inc. Affidavit Requesting Proprietary Report be Withheld from Public Disclosure
: 4. AREVA, Inc. Document No. ANP-3463NP, "Response to RAl's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel" (Non-Proprietary) cc: Regional Administrator-NRC Region Ill NRC Senior Resident Inspector-Dresden Nuclear Power Station NRC Senior Resident Inspector
-Quad Cities Nuclear Power Station Illinois Emergency Management Agency -Division of Nuclear Safety Attachment 2 transmitted herewith contains Proprietary Information.
When separated from attachments, this document is decontrolled.
ATTACHMENT 1 Response to Request for Additional Information Page 1 of 11 NRC RAI 1 In Attachment 1 to the submittal dated February 6, 2015 (Agencywide Document Access Management System (ADAMS) Accession No. ML15043A489), the licensee states that:
 
-the application of the AREVA methodologies to a representative transition core design of ATRIUM 10XM and OPTIMA2 fuel (i.e., QCNPS Units 1 and 2) is sufficient to demonstrate the applicability of the methodologies for AREVA ATRIUM 10XM fuel at both stations (i.e. DNPS
 
and QCNPS).
 
Provide a description demonstrating that the non-cycle specific accident analyses are applicable to the transition cores for each unit and provide a justification that the representative cycle values for thermal limits and transients are characteristic of typical values expected for the actual transition cores for each unit.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 2 Tables 3.5 through 3.8 of Attachment 9 of the submittal appears to indicate that the critical power ratio (CPR) for OPTIMA2 fuel is consistently lower than that of ATRIUM 10XM fuel.
For the transition core, address whether the OPTIMA2 fuel will be the limiting fuel from the standpoint of operating limit maximum CPR values calculated for the anticipated operational occurrences (AOOs).
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 3 Page 2-1 of Attachment 9 to the submittal, appears to indicate that under rated conditions the ATRIUM 10XM assembly has a lower flow compared to an OPTIMA2 assembly, at the same power. However, tables 3.5 through 3.8 in Attachment 9 appear to show that the CPR for OPTIMA2 fuel remains consistently lower than that of ATRIUM 10XM fuel. 
 
Given the identified flow for the ATRIUM 10XM assembly, address why the data appears to indicate CPR values for OPTIMA2 fuel are lower than that of the ATRIUM 10XM fuel.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 4 Page 4-5 of Attachment 5 to the supplement dated September 1, 2015 (ADAMS Accession No.
ML15251A381) discusses that the low pressure coolant injection (LPCI) and low pressure core spray (LPCS) systems inherently have leakage flows which do not reach the intended injection or target location. Due to LPCI/LPCS flow leakage; Provide the following information:
ATTACHMENT 1 Response to Request for Additional Information Page 2 of 11
: a. The percentage (%) of assumed leakage rate for each the LPCI and LPCS system flows; 
: b. Address whether the assumed leakage rate for each the LPCI and LPCS system flows is conservative and bounds the actual leakage through each system; and,  c. Provide the basis for the LPCI and LPCS system flow leakage.
Response The following data provide percentages of the LPCI and LPCS leakage flow rates used in the QCNPS Loss of Coolant Accident (LOCA) analysis (Attachment 5 to the submittal dated September 1, 2015). The limiting of the leakages for QCNPS Units 1 and 2 is conservatively used and deducted from the Emergency Core Cooling System (ECCS) flow assumed in the analysis. As such, the leakage flow rate is conservative since it assumes a bounding value of leakage flow for each unit along with some additional leakage. Section 4.6 of the LOCA break spectrum analysis report (Attachment 5 to the submittal dated September 1, 2015) contains discussion on the approach in using these leakages. 
 
Table 4.1 Core Spray Leakage Outside Shroud (per loop)
Percent of Core Spray Rated Flow Rate (4500 gpm at 90 psid)
Core Spray (CS) Quad Cities (U1/U2)  Basis Assessed Leakage Flow 5.6/1.6 Bounding leakages for the identified piping flaws are assessed following BWR Vessel and Internals Project. Bounding design leakages are assessed and accounted for in the analyses. Assumed 7.8/3.8 Bounding leakage value is used in the analysis.
 
ATTACHMENT 1 Response to Request for Additional Information Page 3 of 11 Table 4.2 LPCI Leakage (per loop) Percent of LPCI Rated Flow Rate (9000 gpm at 20 psid)
LPCI Quad Cities (U1/U2)  Basis Assessed Leakage Flow 9.6/6.5 Bounding leakages for the identified piping flaws are assessed following BWR Vessel and Internals Project. Bounding design leakages are assessed and accounted for in the analyses. Assumed 11.5/8.7 Bounding leakage value is used in the analysis.
Jet Pump Slip Joints 2.5/2.5 A bounding design leakage is assessed and accounted for in the long-term cooling evaluation. This is based on 2/3 core height flooding.
NRC RAI 5 On page 1-1 of Attachment 5 to the supplement dated September 1, 2015, states: 
 
[t]he break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUM 10XM fuel at beginning-of-life (BOL) conditions. The actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels.
Assuming that there are appreciable geometrical differences between the two fuel designs causing different thermal-hydraulic conditions to exist within the assemblies in a transition core, and that the actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels. Discuss the following:
: a. The assumption made in the loss-of-coolant-accident (LOCA) analysis that a core composed entirely of ATRIUM 10XM will conservatively bound the peak clad temperature (PCT) reported for the limiting break analysis, and,  b. The conservatism of the limiting PCT of 2127 degrees Fahrenheit (°F), and whether this is bounding for either of the fuel types, i.e., ATRIUM 10XM or OPTIMA2.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 6 The transition from OPTIMA2 to ATRIUM 10XM fuel on LOCA analysis is described in Attachment 5 to the supplement dated September 1, 2015. 
: a. Provide a comparison of the values of PCT, limiting break-size (small or large-break LOCA), and break location for the current licensing-basis LOCA to that of the predicted values for the transition core.
ATTACHMENT 1 Response to Request for Additional Information Page 4 of 11
: b. Address whether there are any significant changes in the parameters specified above in 6.a between the current licensing basis (CLB) and the representative transition core. If so, explain.
Response a. Exelon will maintain separate LOCA analyses for the fuel types residing in the transition cores at DNPS and QCNPS. The table below shows key comparisons between the QCNPS LOCA break spectrum analysis results. The licensing basis PCT for each fuel type is confirmed based on the cycle-specific MAPLHGR evaluation (see response to RAI 5.b above). Table 6.1 Analysis  (Fuel Type)
PCT (°F) Limiting Break Size  Break Location Single Failure Westinghouse (OPTIMA2) 2150  Large - 1.0 Double-Ended Guillotine  Recirculation pump suction LPCI - injection valve AREVA  (ATRIUM-10XM) 2127  Small - 0.13 ft 2 (split) Recirculation Discharge Pipe HPCI  b. The ECCS parameters input to the LOCA analyses for the fuel types in question are the same. The differences shown in the above table are due mainly to the licensed methodology for each fuel vendor.
 
NRC RAI 7 In Table 6.1 of Attachment 5 to the supplement dated September 1, 2015, the limiting PCT was reported as 2127 °F. In Table 2.1 of Attachment 14 of the submittal dated February 6, 2015, the PCT was reported as 2138 °F. 
 
Clarify why the higher PCT value of 2138 °F is not the limiting PCT.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 8 Table 2.1 of Attachment 12 of the submittal dated February 6, 2015, provides a summary of the disposition of events for ATRIUM 10XM Fuel Introduction at QCNPS. 
 
Discuss whether any significant changes are observed in the transients and accidents analyzed, including potentially limiting events, with ATRIUM 10XM fuel in the core as opposed to OPTIMA2 fuel in the core. If so, discuss the specific events and reasons for the change.
Response See Attachment 2 for this response as provided by AREVA, Inc.
ATTACHMENT 1 Response to Request for Additional Information Page 5 of 11 NRC RAI 9 Page 7-1 of Attachment 12 to the submittal dated February 6, 2015, indicates that the limiting American Society of Mechanical Engineers (ASME) over pressurization event was the feedwater controller failure with turbine bypass valves out-of-service and high neutron flux scram. For most of the boiling-water reactors, typically main steam isolation valve (MSIV) closure with high neutron flux scram is the limiting ASME over pressurization event.
: a. Address why the MSIV closure with high neutron flux scram is not the limiting over pressurization event at DNPS/QCNPS, and
: b. Discuss any significant changes in the results of the limiting ASME over pressurization analysis due to transition from OPTIMA2 to ATRIUM 10XM fuel.
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 10 Describe the methodology used to perform the anticipated transient without scram analysis, its applicability to mixed core of ATRIUM 10XM and OPTIMA2 fuels, and whether the computer code employed was NRC-approved.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
NRC RAI 11 Page 4-2 of Attachment 12 to the submittal dated February 6, 2015, states that QCNPS has implemented the Boiling Water Reactor Owners Group (BWROG) Long Term Stability Solution Option III (oscillation power range monitor (OPRM)). It was further stated that in cases where the OPRM system is declared inoperable backup stability protection (BSP) is provided in accordance with OG02-0119-260, "Backup Stability Protection for Inoperable Option III Solution," General Electric Nuclear Energy, July 17, 2002. 
 
Discuss the long term stability solution option and BSP for DNPS.
Response For the AREVA ATRIUM 10XM, the DIVOM cycle-specific analysis will be performed using the long-term stability Option III licensing methodology to establish or confirm the OPRM setpoints.
This analysis follows the NRC approved licensing methodology documented in NEDO-32465-A.
Furthermore, the BSP analysis will be performed to establish or confirm the plant specific exclusion boundary for operation in the event t he OPRM is out of service. The BSP analysis follows the same approach as prescribed in the BWROG guidelines documented in OG 02-0119-260. 
 
NRC RAI 12 Chapter 4.2 of NUREG-0800, "Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," states that the fuel system safety review ATTACHMENT 1 Response to Request for Additional Information Page 6 of 11 should provide assurance that the fuel system is not damaged as a result of normal operation and AOOs. 
 
Appendix A to Chapter 4.2 indicates that earth quakes and postulated pipe breaks in reactor coolant system would result in external forces on the fuel assembly. It further states that fuel system coolability should be maintained and that the damage should not be so severe as to prevent control rod insertion when required during these low probability accidents.
to the supplement dated September 1, 2015 (ADAMS Accession No.
ML15251A381), summarizes the evaluation of fuel handling loads and structural evaluation of faulted conditions to confirm the structural integrity of the fuel assembly components. Regarding the structural integrity of the fuel assembly components under normal and faulted conditions, provide the following information:
: a. Details of the stress calculations performed to confirm the design margin to establish a baseline for adding accident loads.
: b. Details of the analysis and testing that determined the maximum axial handling of the load by the fuel assembly without yielding.
: c. Details of the rod bow analysis for ATRIUM 10XM fuel design and its impact on thermal margin assessment.
: d. Discuss the results of the rod bow analysis for the OPTIMA2 fuel design for each core and the impact of rod bow on thermal margin assessment.
Response a. See Attachment 2 for this response as provided by AREVA, Inc. b. See Attachment 2 for this response as provided by AREVA, Inc. c. See Attachment 2 for this response as provided by AREVA, Inc. d. Compliance with Chapter 4.2 of NUGREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," for OPTIMA2 fuel is accomplished by developing fuel operating limits to provide assurance that the fuel system is not damaged as a result of normal operation and AOOs. For OPTIMA2 fuel, these limits are developed in accordance with approved Westinghouse methodology. 
 
The detailed evaluation of OPTIMA2 fuel rod bow using Westinghouse methods can be found in Westinghouse Topical Report WCAP-15942-P, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENP-287," submitted to the NRC for review on October 29, 2004 (Accession No. ML043090165). This topical report was subsequently approved by the NRC on February 2, 2006 (Accession No. ML060110269). To comply with the fuel rod bow design criterion from WCAP-15942-P, the OPTIMA2 fuel design utilizes materials, manufacturing process controls, as well as assembly mechanical design features to minimize the potential for rod bow. The specific design features are listed in Section 4.3.10 of WCAP-15942-P-A.
In a letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Additional Information Supporting Request for License Amendment Regarding Transition to ATTACHMENT 1 Response to Request for Additional Information Page 7 of 11 Westinghouse Fuel," dated January 26, 2006, EGC documented that the Westinghouse methods for determining thermal mechanical operating limits (TMOL) for OPTIMA2 fuel at DNPS and QCNPS comply with fuel design bases, including fuel rod bow. Westinghouse provides the TMOLs for the OPTIMA2 fuel in terms of LHGR operating limits that should not be exceeded during plant operation. During the fuel transition, the co-resident OPTIMA2 fuel LHGR limits and LOCA APLHGR limits will continue to be monitored by the limits previously provided by Westinghouse.
 
The development of the AREVA critical power correlation for OPTIMA2 fuel was performed in accordance with EMF-2245(P)(A), Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, dated August 2000, and is detailed in Section 4 of Attachment 1 to the supplement dated September 1, 2015. This method uses CPR data provided by Westinghouse for OPTIMA2 fuel over a wide range of conditions, and implicitly includes the effect of rod bow detailed in WCAP-15942-P.
 
Therefore, the results of the rod bow analysis for the OPTIMA2 design were included in the generic Westinghouse Mechanical Design Methods (WCAP-15942-P), the associated Westinghouse methodologies, and the DNPS and QCNPS License Amendment Request for OPTIMA2 fuel. The impact of rod bow on OPTIMA2 thermal margin assessment is included in the LHGR and APLHGR limits developed by Westinghouse. These limits are independent of co-resident AREVA fuel, and will continue to be used for monitoring OPTIMA2 fuel during the transition. The AREVA critical power correlation for OPTIMA2 fuel was developed in accordance with approved methods and is based on CPR data supplied by Westinghouse that implicitly includes the effect of rod bow.
 
NRC RAI 13 Section 3.2.3 of Attachment 10 to the supplement dated September 1, 2015, addresses the overheating of fuel pellets. In relation to this section:
: a. Explain how radial depression of the thermal neutron flux is accounted for in defining the local volumetric heat generation rate.
: b. Explain the term "neutronic fuel assembly types." 
: c. Explain how linear heat generation rate (LHGR) uncertainties are calculated using the stated methodology.
: d. Explain how fuel rod power histories are created using the stated methodology. Specifically discuss an equilibrium core design and the upcoming cycle.
: e. Explain why transients are randomly selected using the stated methodology.
: f. Discuss the methodology used for power measurement and operational uncertainties. Address whether this is a deviation from the RODEX4 methodology.
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
ATTACHMENT 1 Response to Request for Additional Information Page 8 of 11 NRC RAI 14 RAI 14 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 15 Provide the dimensional data associated with the resident OPTIMA2 fuel and include fuel channel dimensions, thickness, and length.
Response Dimensional data associated with resident OPTIMA2 fuel can be found in Westinghouse Topical Report WCAP-15942-P, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENP-287," submitted to the NRC for review on October 29, 2004 (Accession No. ML043090165). This topical report was subsequently approved by the NRC on February 2, 2006 (Accession No. ML060110269). WCAP-15942-P-A contains fuel channel dimensions, channel thickness, bundle length, and many additional OPTIMA2 fuel dimensions.
 
NRC RAI 16 With the introduction of ATRIUM 10XM fuel to the DNPS and QCNPS units, discuss for each unit how the licensee has accounted for sufficient clearance for the control rods and in-core instrumentation.
 
Response See Attachment 2 for this response as provided by AREVA, Inc.
NRC RAI 17 With the introduction of ATRIUM 10XM fuel to the DNPS and QC units, discuss for each unit the impacts of channel bow (as a function of burnup) on the clearances for control rods, in-core instrumentation, and adjacent assemblies.
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 18 RAI 18 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 19 RAI 19 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 20 RAI 20 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
ATTACHMENT 1 Response to Request for Additional Information Page 9 of 11 NRC RAI 21 RAI 21 contains proprietary information.
See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 22 Page 1-1 of Attachment 11 to the supplement dated September 1, 2015, Version 2, of MICROBURN-B2 (MB2) is listed as part of the cycle design analysis. 
: a. Address whether Version 2 of MICROBURN-B2 is the same version listed in "Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation (dated October 1999).
: b. Discuss the methodology used in the following MB2 modeling features;
: 1. Explicit control blade modeling. 2. Explicit neutronic treatment of spacer grids. 3. Explicit thermal-hydraulic modeling of water rod flow.
Response See Attachment 2 for this response as provided by AREVA, Inc.
NRC RAI 23 Section 7.1 of Attachment 8 to the supplement dated September 1, 2015, states that for each transition cycle, shutdown margin is computed by performing restart solutions based on a shuffled core from a short window previous cycle condition.
Discuss the impact on the shutdown margin for future cycles if the plants were to operate for either nominal or long cycles.
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 24 Section 7.2 of Attachment 8 to the supplement dated September 1, 2015, addresses LHGR monitoring of advanced fuel designs. From this discussion:
: a. Explain the explicit local power range monitor (LPRM) model,  b. Explain the in-core monitoring system and how the model is used to account for perturbations to the local peaking factors of the rods surrounding the LPRM, and
: c. Explain how the rod power biases due to the presence of LPRM detectors are accounted for in the monitoring of LHGRs.
 
ATTACHMENT 1 Response to Request for Additional Information Page 10 of 11 Response See Attachment 2 for this response as provided by AREVA, Inc.
 
NRC RAI 25 RAI 25 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 26 RAI 26 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
 
NRC RAI 27 Appendix C.2 of Attachment 8 to the supplement dated September 1, 2015, states the correspondence between the assembly powers of adjacent assemblies is quantified by a conservative multiplier as listed on Page C-5. Additionally, this multiplier is based on the correlation coefficient that is statistically calculated and shown in Figure 9.1 and 9.2 of EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (ADAMS Accession No. ML003698495).
: a. Provide the basis of the calculations used to derive the conservative multiplier shown on page C-5. 
: b. In Section 8.2 of EMF-2158(P)(A), the report states a combination of uranium oxide (UO 2) and plutonium oxide (PuO
: 2) bundles are used.
For each unit address the application from a measurement with UO 2 and PuO 2 to each core containing only UO 2 fuel. Describe the process used in this analysis.
Response See Attachment 2 for this response as provided by AREVA, Inc.
NRC RAI 28 Appendix D of Attachment 8 to the supplement dated September 1, 2015, describes the 11/2 group diffusion equation (page D-4).
Address whether the first term of the equation should be 1f1 and the  in the second term should be 2 to reflect fission rates for groups 1 and 2.
Response See Attachment 2 for this response as provided by AREVA, Inc.
NRC RAI 29 RAI 29 contains proprietary information. See Attachment 2 for this RAI and response as provided by AREVA, Inc.
ATTACHMENT 1 Response to Request for Additional Information Page 11 of 11 NRC RAI 30 Appendix F of Attachment 8 to the supplement dated September 1, 2015, summarizes the impact and treatment of fuel thermal conductivity degradation (TCD) with fuel burnup for licensing safety analyses such as AOOs, LOCA analyses.
: a. Address whether TCD was applied to the models provided in letters dated July 14, 2009 and April 27, 2012 (ADAMS Accession Nos. ML092010157 and ML121220377, respectively). 
: b. For each unit, discuss how AREVA intends to implement TCD models.
Response See Attachment 2 for this response as provided by AREVA, Inc.
 
ATTACHMENT 2 AREVA, Inc. Document No. ANP-3463P  "Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel"  (Proprietary)
 
ATTACHMENT 3 AREVA, Inc. Affidavit Requesting Proprietary Report be Withheld from Public Disclosure
 
STATE OF WASHINGTON COUNTY OF BENTON SS. AFFIDAVIT
: 1. My name is Alan B. Meginnis.
I am Manager, Product Licensing, for AREVA Inc. and as such I am authorized to execute this Affidavit.
: 2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.
I am familiar with the policies established by AREVA to ensure the proper application of these criteria.
: 3. I am familiar with the AREVA information contained in the report ANP-3463P, Revision 0, "Response to RAl's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel," dated January 2016 and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA 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 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 accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 1 O CFR 2.390(a)(4)
'Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA'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, 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. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA. The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above. 7. In accordance with AREVA'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 only as required and under suitable agreement providing for nondisclosure and limited use of the information.
: 8. AREVA 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 lG!A..'f
'2016. Mary Anne Heilman NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 6/6/2016 
 
ATTACHMENT 4 AREVA, Inc. Document No. ANP-3463NP  "Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel"  (Non-Proprietary)
ANP-3463 N P Revision 0 Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel January 201 6 AREVA Inc.
  (c) 2016 AREVA Inc.
 
ANP-3463 N P  Revision 0      Copyright © 2016 AREVA Inc. All Rights Reserved
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page i  Nature of Changes Item Section(s) or Page(s) Description and Justification 1 All Initial Issue
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page ii  Contents  Introduction
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1-1 1.0 RAIs and Responses
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....2-1 2.0 Tables Table 14.1  First Transition Core Hydraulic Channel Grouping and Radial Power Distribution
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2-18 Table 18.1 SPCB and ACE Range of Applicability Comparison for Quad Cities / Dresden ................................
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2-22    Figures Figure 12.1 Predicted Gap Closure versus exposure for ATRIUM 10XM and ATRIUM-10A designs
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2-13 Figure 12.2 MCPR Penalty versus Fuel Rod Gap Closure
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2-13  Figure 20.1 Location of Assemblies with Fast Fluence Gradients Outside of Channel Measurement Database
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....................................... 2-28 Figure 20.2  Core Limiting MFLCPR Comparison, Assemblies with Channels Outside Measurement Database
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....................................... 2-29 Figure 20.3  Core Limiting MFLCPR Comparison, Face Adjacent OPTIMA2
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2-29 Figure 20.4  Core Limiting MFLCPR Comparison, Face Adjacent ATRIUM 10XM
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2-30  Figure 22.1  Effect of Heated Perimeter on Spacer Cross Section Multiplier
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2-35 Figure 22.2  Effect of Spacer Volume on Spacer Cross Section Multiplier
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2-36 Figure 22.3  Effect of Void and Heated Perimeter on Spacer Cross Section Multiplier
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2-37 Figure 24.1  Effect of LPRM Modeling on the Bundle GEH01 Axial level 1 Pin Gamma Scan Comparison for Quad Cities Unit 1 EOC 2
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.. 2-44 
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 1-1  Introduction 1.0In Reference 1, Exelon requested an amendment to the Renewed Facility Operating Licenses for Dresden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear Power Station Units 1 and 2. The request was supplemented by Reference
: 2. The amendment, if approved, would allow for a transition to the AREVA ATRIUM 10XM fuel design at all four units. The amendment would also allow the implementation of AREVA licensing analysis methods.
The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing the license amendment request and has determined that additional information is needed to complete the review (Reference 3). While all the Requests for Additional Information (RAI) are presented, only the responses prepared by AREVA are included in this document. The responses that will be provided by Exelon are identified accordingly.
References
: 1. Letter, PR Simpson (Exelon) to USNRC, "Request for License Amendment Regarding Transition to AREVA Fuel," RS 008, February 6, 2015 (Accession Number ML15055A154).
: 2. Letter, PR Simpson (Exelon) to USNRC, "Supplemental Information in Support of Request for License Amendment Regarding Transition to AREVA Fuel," RS 237, September 1, 2015. (Accession Number ML15251A381).
: 3. Letter, EA Brown (USNRC) to BC Hanson (Exelon),"Dresden Nuclear Power Station , Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 - Request for Additional Information Related to the License Amendment Request to Transition to AREVA Fuel (CAC Nos. MF5736, MF5737, MF5738, and MF5739)", January 19, 2016.
: 4. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co
-Resident Fuel, Siemens Power Corporation, August 2000.
: 5. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses , Advanced Nuclear Fuels Corporation, August 1990.
: 6. Letter, S. Richards (NRC) to J. F. Mallay (FANP), "Siemens Power Corporation RE: Request for Concurrence on Safety Evaluation Report Clarifications (TAC No. MA6160)," May 31, 2000.
: 7. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO
-4/MICROBURN
-B2 , Siemens Power Corporation, October 1999.
: 8. ASME Boiler and Pressure Vessel Code, Section III, Division 1, American Society of Mechanical Engineers.
: 9. ANP-3305P Revision 1, Mechanical Design Report for Quad Cities and Dresden ATRIUM 10XM Fuel Assemblies, AREVA, August 2015.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 1-2  10. XN-NF-75-32(P)(A) Supplements 1 through 4, Computational Procedure for Evaluating Fuel Rod Bowing, Exxon Nuclear Company, October 1983. (Base document not approved.)
: 11. XN-NF-82-06(P)(A) Supplement 1 Revision 2, Qualification of Exxon Nuclear Fuel for Extended Burnup, Supplement 1, Extended Burnup Qualification of ENC 9x9 BWR Fuel , Advanced Nuclear Fuels Corporation, May 1988.
: 12. ANP-3287(P) Revision 1, Quad Cities Units 1 and 2 Thermal
-Hydraulic Design Report for ATRIUM 10XM Fuel Assemblies, AREVA, November 2014.
: 13. EMF-2994(P), Revision 6,  RODEX4: Thermal
-Mechanical Fuel Rod Performance Code, Theory Manual , AREVA NP, February 2012. 14. BAW-10247PA, Revision 0, Realistic Thermal
-Mechanical Fuel Rod Methodology for Boiling Water Reactors
, AREVA NP, February 2008. 15. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009. 16. ANP-102 98P-A Revision 1, ACE / ATRIUM 10XM Critical Power Correlation, AREVA, March 2014.
: 17. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
: 18. General Electric Co., "Quad Cities Nuclear and Fuel Performance Measurements." EPRI NP-3568, Electric Power Research Institute, July 1984
.
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-1  RAIs and Responses 2.0 RAI-1 2.1In Attachment 1 to the submittal dated February 6, 2015 (Agencywide Document Access Management System (ADAMS) Accession No. ML15043A489)
, the licensee states that:
  -the application of the AREVA methodologies to a representative transition core design of ATRIUM 10XM and OPTIMA2 fuel (i.e., QCNPS Units 1 and 2) is sufficient to demonstrate the applicability of the methodologies for AREVA ATRIUM 10XM fuel at both stations (i.e.
, DNPS and QCNPS).
Provide a description demonstrating that the non
-cycle specific accident analyses are applicable to the transition cores for each unit and provide a justification that the representative cycle values for thermal limits and transients are characteristic of typical values expected for the actual transition cores for each unit.
AREVA Response There are two non
-cycle specific accident analyses that are applicable to the transition cores and a full core of ATRIUM 10XM fuel; the loss
-of-coolant-accident (LOCA) break spectrum analysis and the fuel handling accident (FHA).
The Quad Cities ATRIUM 10XM LOCA break spectrum analysis is presented in Attachment 5 to the supplement dated September 1, 2015. The analysis is applicable for Quad Cities Units 1 and 2 for the transition and full ATRIUM 10XM cores.
Justification of the applicability of the break spectrum analysis for the transition core configurations is provided in the response to RAI-5. The follow
-on heatup analysis that evaluates the fuel response to a LOCA event is performed for each new nuclear fuel design loaded in the core to confirm the applicability of the MAPLHGR limits. A separate LOCA analysis will be performed for the Dresden units using the same methodology. The results are expected to be similar as the plants have very similar ECCS systems. While the capacities of ECCS systems are somewhat different causing some differences in the results, the application of the methodology will be the same.
The fuel handling accident (FHA) is applicable for both plants as the mechanical designs of the ATRIUM 10XM and co
-resident fuels are the same at both plants. In addition, the source term for the ATRIUM 10XM fuel is the same for both plants. The dose analysis assumed a source term based on the fuel types and rod failures for each core. AREVA has performed a fuel handling accident analysis for the ATRIUM 10XM fuel design and determined that a maximum of 162 fuel rods fail. This result bounds both a full core of ATRIUM 10XM and a transition core with both ATRIUM 10XM and OPTIMA2 fuel.
This number of failed rods is less than the 176 failed rods in the GE14 10x10 FHA analysis, and is bounded by dose from licensing basis failure of 111 fuel rods in the GE7x7 fuel. Therefore the current FHA alternate source term analysis remains applicable for the introduction of ATR IUM 10XM fuel design. As long as the mechanical design stays the same, the FHA analysis remains applicable for the transition cores and full ATRIUM 10XM cores.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-2  The representative cycle values for thermal limits and transients are characteristic of typical values for the actual transition cores for each unit because The representative cycle fuel and core design are typical of transition cores composed of fresh ATRIUM 10XM fuel and previously irradiated OPTIMA2 fuel, since typical energy requirements were provided in the core design specification.
Quad Cities and Dresden are both similar size BWR/3 plants with 724 assemblies. While there are some minor differences in some of the plant parameters, the response to the various events is expected to be similar.
Some cycle to cycle variation in the results is expected due to varying design requirements, but the changes in transient results would be relatively small. Changes to equipment and/or parameters could also impact the thermal limits and transient results.
The system response to the transients for the second reload of ATRIUM 10XM fuel is expected to be similar to the first transition core.
[            ]  RAI-2 2.2Tables 3.5 through 3.8 of Attachment 9 of the submittal appears to indicate that the critical power ratio (CPR) for OPTIMA2 fuel is consistently lower than that of ATRIUM 10XM fuel.
For the transition core, address whether the OPTIMA2 fuel will be the limiting fuel from the standpoint of operating limit maximum CPR values calculated for the anticipated operational occurrences (AOOs).
AREVA Response The main purpose of Attachment 9 is to show that the AREVA ATRIUM 10XM fuel is hydraulically compatible with the previously loaded Westinghouse OPTIMA2 fuel. This is accomplished by evaluating the impact on core pressure drop and flow distribution among the fuel assemblies. The analysis results can also provide an indication of how certain parameters -
including CPR for a given fuel design - change as the core transitions to a full core of ATRIUM 10XM fuel. The CPR results presented in Tables 3.7 and 3.8 in Attachment 9 show that as the core transitions to a full core of ATRIUM 10XM fuel, the CPR for the OPTIMA2 and AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-3  ATRIUM 10XM fuel
[  ]. The CPR results are not intended to be licensing values or to assess the relative CPR performance of the two fuel designs.
RAI-3 2.3Page 2-1 of Attachment 9 to the submittal appears to indicate that under rated conditions the ATRIUM 10XM assembly has a lower flow compared to an OPTIMA2 assembly, at the same power. However, Tables 3.5 through 3.8 in Attachment 9 appear to show that the CPR for OPTIMA2 fuel remains consistently lower than that of ATRIUM 10XM fuel.
Given the identified flow for the ATRIUM 10XM assembly, address why the data appears to indicate CPR values for OPTIMA2 fuel are lower than that of the ATRIUM 10XM fuel.
AREVA Response While assembly flow is an important parameter in the critical power performance of a given fuel design, it is not the only parameter that impacts the CPR.
[        ]  RAI-4 2.4Page 4-5 of Attachment 5 to the supplement dated September 1, 2015 (ADAMS Accession No. ML15251A381), discusses that the low pressure coolant injection (LPCI) and low pressure core spray (LPCS) systems inherently have leakage flows which do not reach the intended injection or target location. Due to LPCI/LPCS flow leakage; Provide the following information:
: a. The percentage (%) of assumed leakage rate for each the LPCI and LPCS system flows;
: b. Address whether the assumed leakage rate for each the LPCI and LPCS system flows is conservative and bounds the actual leakage through each system; and,  c. Provide the basis for the LPCI and LPCS system flow leakage.
AREVA Response Exelon to provide a response separate from this document.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-4  RAI-5 2.5On page 1-1 of Attachment 5 to the supplement dated September 1, 2015, state s:  [t]he break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUM 10XM fuel at beginning
-of-life conditions. The actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels.
 
Assuming that there are appreciable geometrical differences between the two fuel designs causing different thermal
-hydraulic conditions to exist within the assemblies in a transition core and that the actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels. Discuss the following:
: a. The assumption made in the lo ss-of-coolant accident (LOCA) analysis that a core composed entirely of ATRIUM 10XM will conservatively bound the peak clad temperature (PCT) reported for the limiting break analysis, and, 
: b. The conservatism of the limiting PCT of 2127 degrees Fahrenheit (° F), and whether this is bounding for either of the fuel types, i.e., ATRIUM 10XM or OPTIMA2.
AREVA Response
: a. As presented in Figure 4.1 of Attachment 5, the LOCA break spectrum analysis includes a system calculation and a hot assembly analysis. 
[            ] The hot assembly analysis uses the system calculation results as boundary conditions to determine the hot channel response to the loss of coolant accident.  [  ] b. The purpose of the analyses presented in Attachment 5 is to determine the characteristics of the limiting break, i.e., the break location, break type, break size, limiting ECCS single failure, and the limiting axial power shape. The highest peak clad temperature (2127
°F for the Quad Cities analysis) is used as a measure to determine the limiting break. Even though the characteristics of the limiting break do not change AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-5  with exposure or nuclear fuel design, the PCT calculated for any given break is dependent on the exposure and local power peaking of the fuel assembly. As a result, heatup analyses are performed and reported in the follow
-on LOCA-ECCS MAPLHGR report, ANP
-3356P (Attachment 14 to the submittal dated February 6, 2015) to determine the licensing PCT for the ATRIUM 10XM fuel. The PCT presented in the LOCA-ECCS MAPLHGR report is the licensing PCT for the ATRIUM 10XM fuel. The heatup analyses are performed for each new ATRIUM 10XM nuclear fuel design loaded in the core.
The PCT for the previous OPTIMA2 LOCA evaluation remains applicable during the transition to a full core of ATRIUM 10XM fuel.
RAI-6 2.6The transition from OPTIMA2 to ATRIUM 10XM fuel on LOCA analysis is described in Attachment 5 to the supplement dated September 1, 2015.
: a. Provide a comparison of the values of PCT, limiting break
-size (small or large
-break LOCA), and break location for the current licensing
-basis LOCA to that of the predicted values for the transition core.
: b. Address whether there are any significant changes in the parameters specified above in 6.a between the current licensing basis and the representative transition core. If so, explain. AREVA Response Exelon to provide a response separate from this document.
RAI-7 2.7In Table 6.1 of Attachment 5 to the supplement dated September 1, 2015, the limiting PCT was reported as 2127
° F. In Table 2.1 of Attachment 14 of the submittal dated February 6, 2015, the PCT was reported as 2138 ° F.
Clarify why the higher PCT value of 2138 ° F is not the limiting PCT.
AREVA Response The purpose of the LOCA break spectrum report (Attachment 5 to the supplement dated September 1, 2015) is to identify the characteristics of the limiting break including:  break location, break type, break size, limiting single failure, and the limiting axial power shape. The limiting break is the one with the highest peak clad temperature. The limiting break for Quad Cities is a 0.13 ft 2 split break in the recirculation pump discharge pipe with a top
-peaked axial power shape and a single failure of the HPCI system. The peak clad temperature is 2127 °F, which is the highest temperature reported in the break spectrum report. While the peak clad AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-6  temperatures in the break spectrum report are used for comparative purposes to identify the characteristics of the limiting break, the highest value is not the final licensing PCT.
Even though the characteristics of the limiting break do not change with exposure or nuclear fuel design, the PCT calculated for any given break is dependent on the exposure and local power peaking of the fuel assembly. Therefore, peak clad temperature results are dependent on the lattice design. As a result, heatup analyses are performed to determine the PCT versus exposure for the ATRIUM 10XM lattice designs in the core. The highest PCT result is the limiting or licensing PCT. For Quad Cities, the limiting or licensing PCT for ATRIUM 10XM fuel is 2138 °F as reported in Attachment 14 to the submittal dated February 6, 2015. Heatup analyses are performed for each new ATRIUM 10XM nuclear fuel design loaded in the core.
RAI-8 2.8Table 2.1 of Attachment 12 of the submittal dated February 6, 2015, provides a summary of the disposition of events for ATRIUM 10XM Fuel Introduction at QCNPS. Discuss whether any significant changes are observed in the transients and accidents analyzed, including potentially limiting events, with ATRIUM 10XM fuel in the core as opposed to OPTIMA2 fuel in the core. If so, discuss the specific events and reasons for the change.
AREVA Response The disposition of events identifies the potentially limiting events/analyses that need to be evaluated to ensure appropriate operating limits are established and design criteria are met. No new potentially limiting events were identified as a result of the introduction of the ATRIUM 10XM fuel design. The system response to the various transients and accidents will not be significantly impacted by the change in fuel design from OPTIMA2 to ATRIUM 10XM.
RAI-9 2.9 P age 7-1 of Attachment 12 to the submittal dated February 6, 2015, indicates that the limiting American Society of Mechanical Engineers (ASME) over pressurization event was the feedwater controller failure with turbine bypass valves out
-of-service and high neutron flux scram. For most of the boiling-water reactors, typically main steam isolation valve (MSIV) closure with high neutron flux scram is the limiting ASME over pressurization event.
: a. Address why the MSIV closure with high neutron flux scram is not the limiting over pressurization event at DNPS/QCNPS, and
: b. Discuss any significant changes in the results of the limiting ASME over pressurization analysis due to transition from OPTIMA2 to ATRIUM 10XM fuel.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-7  AREVA Response
: a. Consideration of the feedwater controller failure (FWCF) with turbine bypass valves out
-of-service (TBVOOS) in the ASME overpressurization evaluation is to support equipment out of service operation. With equipment in service, the potentially limiting events are consistent with other BWRs.
A review of various aspects of the MSIV closure and FWCF with TBVOOS events shown below indicates there are some aspects of the event that make the overpressure results more or less severe.
Characteristic MSIV Closure FWCF with TBVOOS Valve Closure rate Slow closure of the MSIV (3  sec) spreads out pressure wave Less Severe Turbine stop valve (TSV) closure (0.1 sec) results in a steeper pressure wave More Severe Steam line volume to pressurize Smaller portion of steam line to pressurize More Severe Larger portion of steam line to pressurize Less Severe Scram initiation relative to complete valve closure Occurs close to MSIV full closure    Less Severe Occurs a longer time after complete TSV closure - even when pressure wave propagation is considered More severe Downcomer and steam dome volume to pressurize Larger volume to pressurize Less Severe Smaller volume due to the increase in water level prior to TSV closure  More Severe Change in steam flow at the time of valve closure NA Slight increase in steam flow during the overcooling phase More Severe The comparison of the same characteristics of the events is qualitative in nature and does not support a conclusion that one event should be more severe than the other. They do however show that there are some things that tend to make one event more severe than the other. As a result, both events were included in the analysis. The results presented in Table 7.1 of Attachment 12 show that the peak vessel pressure and peak dome pressure results for the MSIV closure and FWCF with TBVOOS are within 2 psid of each other, so the differences are small.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-8  b. While there may be some cycle to cycle variation in the peak pressure results, no significant changes are expected as the plant transitions from OPTIMA2 to ATRIUM 10XM fuel. RAI-10 2.10Describe the methodology used to perform the anticipated transient without scram analysis, its applicability to mixed core of ATRIUM 10XM and OPTIMA2 fuels, and whether the computer code employed was NRC
-approved. AREVA Response The ATWS analyses that are performed on a cycle
-specific basis include ATWS overpressure and standby liquid control system (SLCS) shutdown margin (SDM). The results of these analyses for a representative cycle are provided in ANP
-3361P (or NP) which is included as Attachment 12 (or 22) of the fuel transition LAR (ML15055A154, dated February 6, 2015). Specifically, the ANP
-3361P reload safety analysis report addresses the ATWS overpressure and SLCS SDM analyses in Sections 7.2.1 and 7.3, respectively.
COTRANSA2 is used to perform the ATWS overpressurization analysis. COTRANSA2 is a system transient simulation code that includes models for the reactor vessel, core, recirculation lines and steam lines. COTRANSA2 is described in Reference
: 5. The code includes the capability to model pressurization events.
The core model input is based on the MICROBURN
-B2 (Reference
: 7) three-dimensional core simulator that explicitly models all fuel designs in the core.
As a result, the impact/influence of the mixed core from neutronic and thermal
-hydraulic perspectives are included in the analysis.
COTRANSA2 is approved by the NRC in the SER associated with Reference
: 5. The list of events for which COTRANSA2 can be used was expanded in Reference 6 to include the initial pressurization phase of ATWS events.
The SLCS SDM analysis is performed using the NRC approved CASMO4 / MICROBURN
-B2 methodology (Reference 7). Application of this methodology to both the OPTIMA2 and ATRIUM 10XM designs remain s within its approval basis and SER restrictions. Extensive benchmarking of past operating cycles for both Dresden and Quad Cities provide s additional assurance of the continued applicability of this methodology to the OPTIMA2 fuel design. Similarly, operating experience in the US with reloads of the ATRIUM 10XM design provide additional confirmation of the continued applicability of this methodology to this fuel design.
An additional non
-cycle specific evaluation was included in the submittal to address the impact of the introduction of ATRIUM 10XM fuel on long term ATWS response. The results of this evaluation are provided in Section 11.3 of the ANP
-3338P (or NP) methods applicability report which is included as Attachment 1 (or 8) of the supplement to the fuel transition LAR (ML15251A381 dated September 1, 2015). As described in this report, [  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-9  [            ]  RAI-11 2.11On page 4-2 of Attachment 12 to the submittal dated February 6, 2015, states that QCNPS has implemented the Boiling Water Reactor Owners Group (BWROG) Long Term Stability Solution Option III (oscillation power range monitor (OPRM)). It further stated that in cases where the OPRM system is declared inoperable backup stability protection (BSP) is provided in accordance with OG02
-0119-260, "Backup Stability Protection for Inoperable Option III Solution," General Electric Nuclear Energy, July 17, 2002. 
 
Discuss the long term stability solution option and BSP for DNPS.
AREVA Response Exelon to provide a response separate from this document.
RAI-12 2.12Chapter 4.2 of NUREG
-0800 "Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition
," states that the fuel system safety review should provide assurance that the fuel system is not damaged as a result of normal operation and AOOs.
Appendix A to Chapter 4.2 indicates that earthquakes and postulated pipe breaks in reactor coolant system would result in external forces on the fuel assembly. It further states that fuel system coolability should be maintained and that the damage should not be so severe as to prevent control rod insertion when required during these low probability accidents.
to the supplement dated September 1, 2015 (ADAMS Accession No. ML15251A381), summarizes the evaluation of fuel handling loads and structural evaluation of faulted conditions to confirm the structural integrity of the fuel assembly components.
Regarding the structural integrity of the fuel assembly components under normal and fault ed AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-10  conditions, provide the following information:
: a. Details of the stress calculations performed to confirm the design margin to establish a baseline for adding accident loads.
: b. Details of the analysis and testing that determined the maximum axial handling of the load by the fuel assembly without yielding.
: c. Details of the rod bow analysis for ATRIUM 10XM fuel design and its impact on thermal margin assessment.
: d. Discuss the results of the rod bow analysis for the OPTIMA2 fuel design for each core and the impact of rod bow on thermal margin assessment.
AREVA Response
: a. As discussed in Reference 9, Section 3.3.1, [  ] the fuel assembly structural components do not receive significant loads during normal and AOO conditions.
[    ]. No analyses are performed to confirm design margin under normal operating and AOO conditions
[  ]. The following text describes how AREVA's approved methodology was conservatively applied for the Quad Cities and Dresden ATRIUM 10XM stress evaluation.
To ensure the structural integrity of
[  ] Section III of the ASME Boiler and Pressure Vessel code (Reference
: 8) is used to establish acceptable design limits. To evaluate the stresses under normal operating conditions, [              ] The maximum normal operation
[  ] for Quad Cities and Dresden analyses is then compared against the limit to ensure that adequate margin is maintained.
To evaluate the stress under AOO and accident conditions, [  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-11  [                    ]  For the [  ] the normal operating stresses
[  ] The design margin is confirmed by comparing the resulting stress to the design limit as defined by Section III of the ASME Boiler and Pressure Vessel code (Reference 8). Information on the stress evaluation results and comparison to the load limits that show that the assembly structural component criteria are maintained under faulted conditions can be found in Table 3
-1, Section 3.4.4 of Reference
: 9. b. As discussed in Section 3.3.9 and Section 4.1 of Reference 9, the fuel assembly must withstand
[  ]  To demonstrate compliance with the criteria, an axial test is performed on the fuel assembly load chain by
[  ] The test is conducted to
[  ] to ensure that the criteria as established in Section 3.3.9 of ANF 98 is satisfied.
Each component is tested in a similar manner. The lower tie plate test is discussed in Section 4.3 of Reference
: 9. The lower tie plate is
[  ] The upper tie plate, upper tie plate locking mechanism, and connecting bolt are also tested in this manner. 
[    ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-12  [  ]  In addition to testing, the lower end plug and water channel handling is
[  ]  c. As described in Section 3.3.5 of Attachment 9 (Reference 9), the method from Reference 10 is used to estimate rod bow on AREVA BWR designs.
The current AREVA methodology for evaluating the impact of rod bow on thermal margins is composed of two steps:
: 1. [          ] The predicted gap closure versus exposure for ATRIUM 10XM is shown in  Figure 12.
1 (dashed red line).
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-13  Figure 12.1  Predicted Gap Closure versus exposure for ATRIUM 10XM and ATRIUM
-10A designs
: 2. [  ] A correlation to predict the CPR reduction due to the reduction in rod
-to-rod spacing was approved during the approval process for extended burnup of AREVA BWR fuel designs (Reference 11). The correlation of MCPR penalty versus gap closure is shown in Figure 12.
: 2. Figure 12.2  MCPR Penalty versus Fuel Rod Gap Closure
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-14  [    ] Rod bow is not considered an important phenomenon in determining the linear power limit; therefore, is not explicitly represented in the RODEX4 methodology.
Any impact of rod bow on the MAPLHGR limit evaluations would be to the pin power distribution.
[    ]  d. Exelon to provide a response separate from this document.
RAI-13 2.13Section 3.2.3 of Attachment 10 to the supplement dated September 1, 2015, addresses the overheating of fuel pellets. In relation to this section:
: a. Explain how radial depression of the thermal neutron flux is accounted for in defining the local volumetric heat generation rate. 
: b. Explain the term "neutronic fuel assembly types."
: c. Explain how linear heat generation rate (LHGR) uncertainties are calculated using the stated methodology.
: d. Explain how fuel rod power histories are created using the stated methodology. Specifically discuss an equilibrium core design and the upcoming cycle.
: e. Explain why transients are randomly selected using the stated methodology.
: f. Discuss the methodology used for power measurement and operational uncertainties. Address whether this a deviation from the RODEX4 methodology.
AREVA Response
: a. The radial depression of the thermal flux is one component of the radial power profile model of RODEX4 and is described in
[      ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-15  [  ]  The volumetric thermal power at any location in the fuel rod is the product of the value of the radial power profile factor at that radius, the input linear power at the axial location and the volume of
[  ] b. Neutronic fuel assembly typing is an identification scheme that groups fuel assemblies by the enrichment and gadolinia distribution within the fuel rods that comprise the assembly.
Thus all fuel assemblies within a given type have the same distribution of these two characteristics. For a given type, the mechanical fuel assembly designs are identical (e.g.,
number of fuel rods; number, location and length of part
-length fuel rods; plenum volumes for each fuel rod; spacer grid design, water channel design, etc.). A reload batch of BWR fuel usually consists of fuel assemblies with identical mechanical designs, but with two or three neutronic types
-and occasionally more.
: c. [              ] d. The preparation of the fuel rod power histories is performed using the nodal simulator. This is described in the topical report (Reference 14, §3.2). The relationship among the cycles that are analyzed for a given reload (denoted as "N") and the equilibrium cycle analysis is shown in the figure below. 
[          ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-16  [  ]  e. The Flow Run
-up and Control Rod Withdrawal Error (CRWE) transients documented in the report are Anticipated Operational Occurrences (AOO's), which are incidents of moderate frequency, and which may occur once during a calendar year of operation for a plant (Regulatory Guide 1.70, Revision 3, Chapter 15, §15.X.X.a).  [        ] f. The methodology used for power measurement and operational uncertainties was discussed as a part of the response to part "c" of this RAI. The analyses described in the subject report are fully consistent with the RODEX4 methodology.
RAI-14 2.14Page 3-3 of Attachment 19 to the submittal dated February 6, 2015 (ADAMS Accession No. ML15043A489), discusses the thermal
-hydraulic compatibility methodology. Specifically, it discusses the methodology used to determine fuel assembly groupings. In relation to this section:
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-17  a. Explain the process of
[ ]];  b. Describe how the core power distributions are input to the compatibility analysis code for the case where the [[ ]; and,  c. Address the assumption that
[ ]. AREVA Response
: a. The assemblies are
[                  ] The assembly grouping used in the hydraulic compatibility analysis of Attachment 9 (proprietary version of Attachment 19) to the submittal dated February 6, 2015, is presented in Table 14.1. b. There are two components of the core power distribution - the axial and radial power distributions. A core average axial power distribution is used in the analyses. Analyses were performed using the bottom
-, middle-, and top-peaked axial power distributions presented in Figure 3.1 of Attachment 9. The radial power distribution input was determined as follows:
: 1. [    ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-18  3. [    ] Table 14.1 presents an example of the assembly grouping and the radial power distribution used in the first transition core loading analysis presented in Attachment 9. Table 14.1  First Transition Core Hydraulic Channel Grouping and Radial Power Distribution
: c. The purpose of the thermal hydraulic compatibility analysis is to evaluate the impact on changes in core pressure drop, assembly flow, bypass flow and critical power performance as the core transitions from a full core of the legacy fuel to a full core of the new reload fuel.
Different axial power distributions will influence the core pressure drop and therefore the flow split between the bypass and assembly flows. While the absolute results change with different core average axial power shapes, the changes in the parameters of interest as the core transitions are very similar. 
[        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-19  RAI-15 2.15Provide the dimensional data associated with the resident OPTIMA2 fuel and include fuel channel dimensions, thickness, and length.
AREVA Response Exelon to provide a response separate from this document.
RAI-16 2.16With the introduction of ATRIUM 10XM fuel to the DNPS and QCNPS units, discuss for each unit how the licensee has accounted for sufficient clearance for the control rods and in
-core instrumentation.
AREVA Response Engineering tests and analyses have been completed to verify that the fuel assembly mechanical compatibility for each reactor's co
-resident fuel and core internals is in compliance with AREVA's generic mechanical design criteria documents for boiling water reactor (BWR) fuel designs (ANF 98(P)(A), Revision 1) and fuel channels (EMF 177(P)(A), Revision 1 and Supplement 1). Verification of the control blade and in
-core instrumentation clearances, specifically, is performed by compatibility analyses where specific bounding physical features and attributes of the fuel assembly, defined in detailed design drawings, are compared to dimensional data officially transmitted from the customer. Dimensional comparisons are made in the cold, beginning
-of-life (BOL) condition. Fuel channel long
-term creep deformation (bulge and bow) is also evaluated to verify that margin to a stuck control rod remains positive. Compatibility to ensure clearance is further discussed in Reference
: 9. RAI-1 7 2.17With the introduction of ATRIUM 10XM fuel to the DNPS and Q C units, discuss for each unit the impacts of channel bow (as a function of burnup) on the clearances for control rods, in-core instrumentation, and adjacent assemblies.
AREVA Response The ATRIUM 10XM with fuel channel BOL dimensional compatibility is discussed in the response for RAI-16. Fuel channel bow towards the control blade as a function of burnup is discussed in the response to RAI-16. Fuel channel bow towards the in
-core instrumentation as a function of burnup is not explicitly evaluated since it is not as limiting as the interference with the control blade.
In addition, channel bow towards the in
-core instrumentation does not impede control blade insertion and safe shutdown is not impacted.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-20  RAI-18 2.18Section 4 of Attachment 8 to the supplement dated September 1, 2015, discusses the AREVA critical heat flux (CHF)/CPR correlations for the co
-resident fuel (OPTIMA2) used for thermal margin analysis.
: a. Page 4-1 states that
[[ ]] 1. Provide a list of SPCB range of applicability with respect to pressure, inlet mass velocity, inlet sub
-cooling, design local peaking and tested local peaking.
: 2. Compare the range of applicability of SPCB correlation with the range of applicability of ACE [AREVA advanced critical power correlation] ATRIUM 10XM correlation. For any parameter where the range of applicability is different between the two correlations, specify the impact on the safety limit minimum CPR determination.
: b. Page 4-2 states that a penalty is applied to bring the [[  ]  Address how the Adder is developed and is implemented in the SPCB correlation.
AREVA Response a.1. The range of applicability of the SPCB correlation as applied to the OPTIMA2 fuel is shown in column two of Table 18.1. 
[    ] a.2. The range of applicability for the ACE ATRIUM 10XM correlation is reported in Reference 16 and also is presented in column three of Table 18.1. As shown, there are differences in the range of applicability between the SPCB and ACE/ATRIUM 10XM correlation for some of the parameters.
[    ] b.  [            ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-21  [  ] The codes that perform OPTIMA2 CPR calculations are MICROBURN
-B2, XCOBRA, XCOBRA-T, RAMONA5
-FA, and SAFLIM3D.
[  ]  MICROBURN-B2 is three
-dimensional core simulator code.
[    ]  XCOBRA is a steady state thermal hydraulics code. 
[  ]  XCOBRA-T is a transient thermal hydraulics code that predicts thermal margin performance during transient events.
[      ]  RAMONA5-FA is used in the Option III stability analysis to calculate the DIVOM (delta CPR over initial CPR versus oscillation magnitude). 
[    ]  SAFLIM3D is used to perform the safety limit minimum CPR calculations. 
[        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-22  Table 18.1  SPCB and ACE Range of Applicability Comparison for Quad Cities / Dresden RAI-19 2.19Appendix G of Attachment 8 to the supplement dated September 1, 2015, states that [[  ]]      ] [                                               
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-23  a. Provide a basis for the quadrant flow uncertainty listed in Appendix G.
: b. Describe the expressions used in calculations demonstrating how the total pressure drop uncertainty and sub
-assembly flow uncertainty are obtained.
AREVA Response
                  ] [
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-24  [  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-25  RAI-20 2.20Page 5-2 of Attachment 8 to the supplement dated September 1, 2015, states that [[  ]]  a. Address whether the computed fluence gradient for each unit is exceeded in the channel measurement database. If yes, provide a typical calculation demonstrating the impact on SLMCPR calculations;
: b. Provide a quantitative summary of how the uncertainty due to the water cross in the OPTIMA2 design impacts the SLMCPR calculations; and,  c. Address whether there any limitations associated with fuel channel bow uncertainty if the computed fluence gradient is not bounded by the channel measurement database.
AREVA Response
: a. A total of eight (8) assemblies in the representative cycle were identified to be outside of the fast fluence gradient lower bound of the channel measurement database, which is
[    ] . The maximum limiting fast fluence gradient for these assemblies occur at EOC (i.e. 16.36 GWd/MTU cycle exposure) with the values provided in the following table.
These are all OPTIMA 2 assemblies located in low power (i.e. non
-limiting) locations of the representative Quad Cities 2 Cycle 24 core. The location s of these assemblies are provided on the core map provided in Figure 20
-1. This figure also provides the location of face adjacent OPTIMA2 and ATRIUM 10XM assemblies located towards the center of the core. The non-limiting nature of assemblies with
[  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-26  [  ] is illustrated in Figure 20
-2 by a comparison of the bundle to core limiting MFLCPRs throughout the cycle. Corresponding plots are provided for the face adjacent OPTIMA2 and ATRIUM 10XM assemblies in Figures 20
-3 and 20-4, respectively. The face adjacent assemblies are provided since adjacent bundles may also be impacted by increased bow in a neighboring location.
[      ] b. [                    ] c. Page 5-2 of ANP-3338P (included as Attachment 1
* of the fuel transition submittal dated September 1, 2015) describes the process used by AREVA
[        ]
* Attachment 8 as identified in the question is the non
-proprietary version of the same report. The information referred to by this question is proprietary and is therefore redacted in Attachment 8.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-27  This calculational approach was previously reviewed and approved by the NRC as part of the approval for the use of the SAFLIM3D methodology (Reference
: 17) for another AREVA customer. In the associated SER (Accession Number ML13037A551) approving this previous application, the following is stated:
- the NRC staff concluded that the increased uncertainty approach will add a sufficient amount of conservatism to the SLMCPR calculation, since (1) the overall SLMCPR calculation is reasonably insensitive to the increased uncertainty value, and (2) the fuel bundles whose channels exceed the channel bow measurement data base tend to be in non
-limiting locations in the core. Based on thes e considerations, the NRC staff determined that the - licensee's application of the AREVA realistic channel bow model is acceptable.
  [      ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-28  Figure 20.
1  Location of Assemblies with Fast Fluence Gradients Outside of Channel Measurement Database
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-29  Figure 20.
2  Core Limiting MFLCPR Comparison, Assemblies with Channels Outside Measurement Database Figure 20.
3  Core Limiting MFLCPR Comparison, Face Adjacent OPTIMA2 AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-30    Figure 20.
4  Core Limiting MFLCPR Comparison, Face Adjacent ATRIUM 10XM RAI-21 2.21Section 7 and Appendix C of Attachment 8 to the supplement dated September 1, 2015, addresses Core Neutronics and Neutronic Methods, respectively. Using these references address the following;
: a. Page C-1 of the report states:
  [m]odels for nodal [[]] are used to improve the accurate representation of the in
-reactor configuration.
Explain how these identified models are used improve the accurate representation of the in-reactor configuration.
: b. Provide details of the interpolation methodology employed to produce Figures C
-10 through C-14 for all sets of intermediate void conditions.
: c. Page C-4 of the report states that the MICROBURN
-B2 (MB2) methodology models a AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-31  wide range of thermal hydraulic conditions including EPU [Extended Power Uprate] and extended power/flow operating map conditions.
 
Address how the MB2 methodology models the various thermal hydraulic conditions and extended range power/flow operating conditions and provide the range. Indicate whether the use of MB2 methodology, in this analysis, is error- free for the entire power/flow range. If not, provide the basis why the use of MB2 methodology is acceptable for this fuels transition.
AREVA Response
: a. [                    ]. These values are incorporated into the Advanced Nodal Expansion Method to determine accurate intra
-nodal flux and power distributions.
: b. Microscopic cross
-section and background cross
-section interpolation is performed using a quadratic Lagrangian method with the three instantaneous void state
-points. Cross section interpolation in exposure space is performed using a piecewise linear interpolation method. 
[  ]  c. The steady
-state thermal
-hydraulic methodology implemented in MICROBURN
-B2 is an expanded version of the AREVA thermal
-hydraulic design and transient analysis methodology XCOBRA. The expansion enables MICROBURN
-B2 to analyze a thermal
-hydraulic network consisting of several hundreds of heat sources, active coolant channels, water channels and bypass channels in the core without homogenizing them into smaller AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-32  number of channels. The expansion also assigns nodal specific heat deposition explicitly calculated with neutron and gamma transport calculation by CASMO
-4 and reconstructed by MICROBURN-B2 to each component of the network instead of the conventional constant heat deposition fractions. This detailed geometric modeling capability enables MICROBURN-B2 to predict local conditions in wide
-ranging conditions. The methodology contains two key constituent correlations in addition to the ASME steam table. One is the void-quality correlation and the other is the channel component flow friction correlation. To assure the validity of these correlations over the operating range of past and future reactor fuel cycles, void fractions and pressure drop measured for current or proposed fuel bundles are benchmarked. The range of thermal
-hydraulic parameters underlying the measurement covers anticipated normal operation transient conditions.
The methodology has been and continues to be evaluated on a large variety of conditions including plants already operating with EPU. The maximum void fraction remains below 0.90 under steady state conditions throughout the power / flow operating domain. Figure C-13 of ANP-3338(P) demonstrated that the nodal cross
-sections resulted in very low differences in k
-infinity for void fractions up to 0.90.
Recent analysis has shown that there was some concern about the void
-quality correlation under low flow conditions when 
[  ]  was implemented. This issue has been addressed and demonstrated that the limiting values reported in this LAR submittal either remain valid or are conservative. In the licensing analyses to be performed for the actual transition, the
[  ]  will be eliminated and the issue no longer has any impact on the results.
RAI-22 2.22On page 1-1 of Attachment 11 to the supplement dated September 1, 2015, Version 2, of MICROBURN-B2 (MB2) is listed as part of the cycle design analysis. a. Address whether Version 2 of MICROBURN
-B2 is the same version listed in "Reactors: Evaluation and Validation of CASMO
-4/M/CROBURN
-B2," Siemens Power Corporation (dated October 1999). b. Discuss the methodology used in the following MB2 modeling features;  1. Explicit control blade modeling.
: 2. Explicit neutronic treatment of spacer grids.
: 3. Explicit thermal
-hydraulic modeling of water rod flow.
AREVA Response
: a. The Version 2 of MICROBURN
-B2 is the same methodology as defined in EMF
-2158(P)(A) with a few enhanced constituent models which have been demonstrated to meet the same requirements specified in EMF
-2158(P)(A) and specifically allowed in the clarification letters included in the approval.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-33  b. 1. Modern control blades contain axially varying loading of absorber material. Absorber material includes B
-10 and Hf nuclides. This design exhibits axial height dependent reactivity variation. These blades may also contain varying absorber and non
-absorber composition along horizontal blade wing. Conventional modeling of control blades makes an implicit assumption that all control blades are close to the original equipment blade and therefore modeled as the original equipment blade, which is uniform axially
 
and horizontally. Calculations using CASMO
-4 and MCNP as well as measurements of local cold criticals show that different blade types exhibit significantly different reactivity effect as well as local peaking factor impact.
Explicit control blade modeling indicates that the specific material composition and mechanical design of various control blades present in the DNPS and QCPS operating cycles are modeled in CASMO
-4 and provided for MICROBURN
-B2 to apply to the appropriate conditions. The cross
-section representation in MICROBURN
-B2 has been expanded to include multiple controlled states defined by the specific control blade present in the core adjacent to the fuel. The explicit modeling also includes an explicit depletion of B
-10 neutron absorber while inserted adjacent to fuel assemblies. This accounts for reduction in control blade strength for sub
-critical reactor shutdown and transient reactor scram.
: 2. Explicit neutronic treatment of spacer grids is an enhanced model recognizing that
 
spacer grid (typically 2-3 cm) covers only a portion of an axial node (typically about 15 cm). When an explicit modeling of spacers for its exact dimension and axial location is sought, the nodal average cross
-section method is not representative of such an inhomogeneous configuration. A base cross section is defined as the standard nodal cross section without the influence of spacer grid. A spacer zone cross section is defined as the same lattice cross section but with the effect of spacer grid included. The spacer zone cross section represents only the block of lattice covered by the spacer grid.
[      ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-34  [              ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-35  Figure 22.1  Effect of Heated Perimeter on Spacer Cross Section Multiplier
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-36  Figure 22.2  Effect of Spacer Volume on Spacer Cross Section Multiplier
[      ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-37  Figure 22.3  Effect of Void and Heated Perimeter on Spacer Cross Section Multiplier
: 3. [          ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-38    [  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-39  [      ]  RAI-23 2.23Section 7.1 of Attachment 8 to the supplement dated September 1, 2015, states that for each transition cycle, shutdown margin is computed by performing restart solutions based on a shuffled core from a short window previous cycle condition. 
 
Discuss the impact on the shutdown margin for future cycles if the plants were to operate for either nominal or long cycles.
AREVA Response The shutdown margin (SDM) calculation referred to is performed as part of the core design and licensing process. The intent is to provide a high level of assurance that the plant will actually meet the SDM requirement in Technical Specification 3.1.1 even though the design is being performed before the final previous cycle shutdown exposure is known. Adherence to the SDM requirement in Technical Specification 3.1.1 is required to be demonstrated during the BOC startup, i.e. after initial criticality. As part of this demonstration, the actual previous cycle burn history is used.
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-40  As noted in the question above, the discussion provided in Section 7.1 of ANP
-3338P (or NP), included as Attachments 1 (or 8) to the fuel transition LAR supplement (ML15251A381 dated September 1, 2015), indicates that the SDM is based upon short window operation for the previous operating cycle. Although not specifically stated, this discussion assumes that no significant upset condition has been encountered during the previous cycle operation.
Fuel cycle economics requires that the bundle and cycle designs deplete the majority of the internal Gadolinia neutron absorber in the fresh fuel prior to the end of cycle. This is require d
since significant gadolinia present at the end of the cycle will impact the cycle energy by requiring a higher initial enrichment or larger batch fraction to meet energy targets. Therefore, in the absence of an upset condition the fresh fuel will have passed the point of peak reactivity and reactivity will continue to decrease as exposure continues to increase. The net impact is that the short cycle exposure window will provide the limiting result for SDM, i.e. the minimum SDM will continue to increase for previous cycle operation past the allowable short exposure window.
It is also recognized that the potential exists for a significant upset condition to occur that could impact this assumption. Specifically, if the fresh fuel does not reach the peak reactivity condition in the previous cycle short window it would be possible for a higher previous cycle exposure to result in a lower SDM. This occurs since the now once
-burnt fuel could still be increasing in reactivity. The most credible conditions that could cause this behavior would be either an early previous cycle shutdown or operation at a significantly reduced power from that planned for a significant portion of the cycle. In the presence of a significant upset condition like these, additional SDM calculations would be performed to ensure that the appropriate previous cycle exposure has been analyzed.
RAI-24 2.24Section 7.2 of Attachment 8 to the supplement dated September 1, 2015, addresses LHGR monitoring of advanced fuel designs. From this discussion:
: a. Explain the explicit low power range monitor (LPRM) model,  b. Explain the in
-core monitoring system and how the model is used to account for perturbations to the local peaking factors of the rods surrounding the LPRM, and
: c. Explain how the rod power biases due to the presence of LPRM detectors are accounted for in the monitoring of LHGRs.
AREVA Response
: a. The explicit LPRM model for LHGR monitoring of advanced fuel designs is a feature that is designed to account for the impact of displacement of moderator by the LPRM detector tube on the LHGRs of surrounding fuel rods. CASMO
-4 has the capability to model the LPRM tube and its internals or its alike such as startup range detector and neutron source tube. This capability provides reference points of comparison for lower order method or core AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-41  simulator. But it is not practical for use in actual core design or monitoring mainly because fuel bundles are shuffled to different positions during their lifetime in the core which may or may not be adjacent to LPRM tube. Hence a corner flux method is developed to account for LPRM tube effect without the use of CASMO
-4 modeling of LPRM tube during fuel lattice homogenization calculation. The corner flux method is described below.
: b. The incore
-monitoring system collects LPRM detector signal from fixed locations in the core and send the collected signal to the AREVA core
-monitoring system called POWERPLEX. POWERPLEX compares the measured LPRM detector signal to the calculated LPRM response by MICROBURN
-B2. The comparison produces adjustment factors called BLPRM values. A combination of BLPRM values and MICROBURN
-B2 calculated neutron TIP distribution produces a synthesized TIP distribution. The comparison of synthesized TIP distribution to the MICROBURN
-B2 calculated TIP distribution together with the MICROBURN-B2 calculated nodal power distribution produces a measured nodal power distribution. This process is described in mathematical form in Section 9.1, EMF2158 (P)(A)
Siemens Power Corporation Methodology for Boiling Water Reactors. The measured nodal power is multiplied to the calculated local peaking factors of fuel rods to produce measured LHGR values. Since the calculated local peaking factors already accounts for the LPRM tubes using the corner flux method described below, the measured LHGR values automatically account for the LPRM tube effect on surrounding fuel rods.
: c. The bias of rod power due to the presence of LPRM tube is described in the corner flux method below.
For a practical application of the corner flux method, the narrow/narrow (N/N) water gap corner occupied with an absorber material (LPRM tube, intermediate or startup range detector, or neutron source tube) presents a challenge. An absorber material in the N/N corner suppresses the neutron flux.
A lattice code can model the realistic geometry of an absorber material present in the N/N corner. But a lattice may be loaded adjacent to an absorber material only for a fraction of its lifetime. To be exact, different lattice types have to be created depending on the presence of an absorber material and its residence time. This is not practical, although not impossible, from a core design point of view. MICROBURN
-B2 deals with this problem in a simple way as presented below.
[        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-42  [    ]  [  ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-43  [        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-44  Figure 24.
1  Effect of LPRM Modeling on the Bundle GEH01 Axial level 1 Pin Gamma Scan Comparison for Quad Cities Unit 1 EOC 2
  [        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-45  [  ]  RAI-25 2.25Appendix B of Attachment 8 to the supplement dated September 1, 2015, states that even though the multi
-rod database used in the
[[]] was obtained through third party organizations, the database and prediction uncertainties are not available to AREVA.
Explain how the correlation has been independently validated by AREVA against public domain multi-rod data and proprietary data from ATRIUM
-10 and ATRIUM 10XM test assemblies.
AREVA Response The [  ], as implemented in MICROBURN
-B2, has been independently validated by AREVA against public domain multi
-rod data and proprietary data collected for AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-46  prototypical ATRIUM
-10 and ATRIUM 10XM test assemblies. Table B
-1 of Attachment 8 of the supplement dated September 1, 2015 presents the AREVA Multi
-Rod Void Fraction Validation database used to validate void fraction predictions. The FRIGG dataset is publicly available (References 16, 17, and 18 of Attachment 8 of the supplement). Specific void fraction measurement campaigns were performed at AREVA's KATHY test facility where void fraction measurements were taken for prototypic assemblies (ATRIUM
-10 and ATRIUM 10XM) under a wide range of conditions. Specific inputs were created to simulate each test and geometry. Figure B-1 and Figure B
-2 of Attachment 8 show the performance of the predictions against measured data.
RAI-26 2.26Figures B-3 and B-4 of Attachment 8 to the supplement dated September 1, 2015, exhibits comparisons of [[ ]]    In both these curves, there has been a shift in the data points outside the band for
-0.05 (predicted - measured) for a range of void fraction 0.40 to 0.80.
Provide the basis which validates the (Ohkawa
-Lahey using ATRIUM
-10 and ATRIUM 10 XM Void Data) database considering the scatter of the data points.
 
AREVA Response Figures B-3 and B-4 of Attachment 8 to the supplement present the data comparison of the Ohkawa-Lahey correlation to public domain multi
-rod data and proprietary data collected for prototypical ATRIUM
-10 and ATRIUM 10XM test assembles.
[      ] As described in Section D.1.2 of Attachment 8 to the supplement, the initial power distributions are based on the
[  ] used in MICROBURN
-B2 (Figures B
-1 and B-2) rather than Ohkawa
-Lahey. [        ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-47  [      ], the use of the correlation is considered appropriate and the overall methodology conservative as described in Sections B.3 and B.4 of Attachment 8 to the supplement. RAI-27 2.27Appendix C.2 of Attachment 8 to the supplement dated September 1, 2015, states the correspondence between the assembly powers of adjacent assemblies is quantified by a conservative multiplier as listed on Page C
-5. Additionally, this multiplier is based on the correlation coefficient that is statistically calculated and shown in Figure 9.1 and 9.2 of EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO
-4/MICROBURN
-B2," (ADAMS Accession No. ML003698495).
: a. Provide the basis of the calculations used to derive the conservative multiplier shown on
 
page C-5. b. In Section 8.2 of EMF
-2158(P)(A), the report states a combination of uranium oxide (UO 2) and plutonium oxide (PuO
: 2) bundles are used.
For each unit address the application from a measurement with UO 2 and PuO 2 to each core containing only UO 2 fuel. Describe the process used in this analysis.
AREVA Response
: a. [    ]
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-48  [      ]  b. Section 8.2 of EMF
-2158(P)(A) presents results for both UO 2 only and combined UO 2 - PuO 2 analyses. The combined results are larger than the UO 2 only results, thus it is conservative to use the combined results.
RAI-28 2.28Appendix D of Attachment 8 to the supplement dated September 1, 2015, describes the 11/2 group diffusion equation (page D
-4). 1f1 2 to reflect fission rates for groups 1 and 2.
AREVA Response There is a typographical error in the subscript of the equation. The equation should read:
The  f term is considered a single entity and separate subscripts are not required.
RAI-29 2.29Appendix D of Attachment 8 to the supplement dated September 1, 2015, states that there are [[ ] two-group cross sections.
Explain what is meant by the [[]]. AREVA Response Cross sections used for input in the COTRANSA
-2 code come from a MICROBURN
-B2 solution where the 3
-D cross sections are collapsed to 1-D set. For each plane of the core the cross sections are evaluated where none of the control blades are present and flux weighted to provide the 1
-D cross section as well as radial buckling. This is the fully uncontrolled state cross section for that plane. Similarly flux weighting is performed where all of the control blades are present (fuel assemblies that are located on the periphery where there may not be a control 1212 2ff a AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-49  blade adjacent are evaluated as uncontrolled). This is the fully controlled state cross section for that plane. This operation is performed for every plane in the core so that every plane has fully uncontrolled and fully controlled cross sections including corresponding radial buckling.
RAI-30 2.30Appendix F of Attachment 8 to the supplement dated September 1, 2015, summarizes the impact and treatment of fuel thermal conductivity degradation (TCD) with fuel burnup for licensing safety analyses such as AOOs , LOCA analyses. a. Address whether TCD was applied to the models provided in letters dated July 14, 2009 and April 27, 2012 (ADAMS Accession Nos. ML092010157 and ML121220377, respectively).
: b. For each unit, discuss how AREVA intends to implement TCD models.
AREVA Response
: a. The references cited discuss the impact of TCD on the design and licensing analyses for BWR fuel. The following codes include some time dependent aspects in the fuel rod modeling and are therefore potentially impacted by TCD.
Code Code Usage TCD Model RODEX4 Fuel rod thermal
-mechanical design and analyses Explicitly models TCD RAMONA5-FA calculates the DIVOM (delta CPR over initial CPR versus oscillation magnitude) used in Option III stability analyses  Explicitly models TCD STAIF Backup stability protection evaluation Explicitly models TCD RODEX2 Fuel rod performance analyses provide gap conductance inputs AOO and overpressurization analyses, stored energy inputs to RELAX LOCA analyses and fuel mechanical inputs to heatup analyses No TCD model COTRANSA2 One-dimensional system transient analysis - anticipated operational occurrences and overpressurization analyses No TCD model
 
AREVA Inc.
Response to RAI's for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2 Transition to AREVA Fuel ANP-3463 N P Revision 0 Page 2-50  Code Code Usage TCD Model XCOBRA-T One-dimensional transient thermal hydraulic analysis for thermal margin evaluations No TCD model RELAX LOCA system and hot channel analysis No TCD model HUXY Exposure-dependent fuel assembly heatup calculations to calculate PCT and local clad oxidation No TCD model The approach to include the effects of TCD in the analyses using the codes that do not explicitly model TCD is summarized in Appendix F of Attachment 8.
: b. As indicated in the Part a response of this RAI, not all the codes include TCD models. The impact of TCD will be included in the potentially affected analysis results as discussed in Appendix F of Attachment 8. The discussion concludes that the potentially affected analysis results are the fuel assembly analyses to determine exposure dependent PCT and local clad oxidation, and the ASME and ATWS overpressurization analyses.
[              ] For the overpressurization events, the impact of TCD is included by reducing the thermal conductivity input to COTRANSA2 to account for the effects of exposure using the exposure-dependent thermal conductivity model in RAMONA5
-FA.}}

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