ML13240A002

Peach Bottom Atomic Power Station, Units 2 & 3 - Extended Power Uprate License Amendment Request - Supplement 9 Response to Request for Additional Information

ML13240A002
Person / Time
Site:	Peach Bottom
Issue date:	08/22/2013
From:	Borton K F Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13240A002 (44)

Text

{{#Wiki_filter:Am-Exelon Generation PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.39010 CFR 50.9010 CFR 2.390August 22, 2013U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3Renewed Facility Operating License Nos. DPR-44 and DPR-56NRC Docket Nos. 50-277 and 50-278

Reference:

Extended Power Uprate License Amendment Request -Supplement 9Response to Request for Additional Information

1. Exelon letter to the NRC, "License Amendment Request -Extended Power Uprate,"

dated September 28, 2012(ADAMS Accession No. ML122860201)

2. NRC letter to Exelon, "Request for Additional Information Regarding License Amendment Request for Extended PowerUprate (TAC Nos. ME9631 and ME9632),"

dated July 23, 2013(ADAMS Accession No. ML1 3203A1 00)In accordance with 10 CFR 50.90, Exelon Generation

Company, LLC (EGC) requested amendments to Facility Operating License Nos. DPR-44 and DPR-56 for Peach BottomAtomic Power Station (PBAPS) Units 2 and 3, respectively (Reference 1). Specifically, the proposed changes would revise the Renewed Operating Licenses to implement anincrease in rated thermal power from 3514 megawatts thermal (MWt) to 3951 MWt.During their technical review of the application, the NRC Staff identified the need foradditional information.

Reference 2 provided the Request for Additional Information (RAI).This letter addresses requests from the staff of Reactor Systems (SRXB) and AccidentDose (AADB) Branches of the U. S. Nuclear Regulatory Commission to provideinformation in support of the request for amendment for the extended power uprate.Responses to these questions are provided in the attachments to this letter:Attachment 1 -Reactor Systems Branch question responses, proprietary versionAttachment 2 -Reactor Systems Branch question responses, non-proprietary, Attachment 3 -Accident Dose Branch question responses Attachment I contains Proprietary Information. When separated from Attachment 1, this document is decontrolled. AuJ EPU LAR Supplement 9Response to Requests for Additional Information August 22, 2013Page 2 of 3GE Hitachi Nuclear Energy America (GEH) considers portions of the information provided in the responses in Attachment 1 to be proprietary and, therefore, exempt frompublic disclosure pursuant to 10 CFR 2.390. The proprietary information in Attachment 1is identified; a non-proprietary version of this information is provided in Attachment

2. Inaccordance with 10 CFR 2.390, EGC requests Attachment 1 be withheld from publicdisclosure.

An affidavit supporting this request for withholding is included asAttachment 4.EGC has reviewed the information supporting a finding of no significant hazardsconsideration and the environmental consideration provided to the U. S. NuclearRegulatory Commission in Reference

1. The supplemental information provided in thissubmittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.

Further, the additional information provided in this submittal does not affect the bases for concluding that neither anenvironmental impact statement nor an environmental assessment needs to be preparedin connection with the proposed amendment.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the Commonwealth of Pennsylvania and the State ofMaryland of this application by transmitting a copy of this letter along with the non-proprietary attachments to the designated State Officials. There are no regulatory commitments contained in this letter.Should you have any questions concerning this letter, please contact Mr. David Neff at(610) 765-5631. I declare under penalty of perjury that the foregoing is true and correct. Executed on the22nd day of August 2013.Respectfully, Kevin F. BortonManager, Licensing -Power UprateExelon Generation

Company, LLCAttachments:

1. Response to Request for Additional Information

-SRXB -Proprietary

2. Response to Request for Additional Information

-SRXB3. Response to Request for Additional Information -AADB4. Affidavit in Support of Request to Withhold Information EPU LAR Supplement 9Response to Requests for Additional Information August 22, 2013Page 3 of 3cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPSUSNRC Project Manager, PBAPSR. R. Janati, Commonwealth of Pennsylvania S. T. Gray, State of Marylandw/attachments w/attachments w/attachments w/o proprietary attachment w/o proprietary attachment Attachment 2Peach Bottom Atomic Power Station Units 2 and 3NRC Docket Nos. 50-277 and 50-278Response to Request for Additional Information -SRXB EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 1Response to Request for Additional Information Reactor Systems BranchBy letter dated September 28, 2012, Exelon Generation

Company, LLC (Exelon)submitted a license amendment request for Peach Bottom Atomic Power Station(PBAPS),

Units 2 and 3. The proposed amendment would authorize an increase in themaximum power level from 3514 megawatts thermal (MWt) to 3951 MWt. Therequested change, referred to as an extended power uprate (EPU), represents anincrease of approximately 12.4 percent above the current licensed thermal power level.The response to SRXB RAI-1 was provided in Supplement 2, dated May, 7, 2013(ADAMS Accession No. ML13129A143). The NRC staff has reviewed the information supporting the proposed amendment and by letter dated July 23, 2013 (ADAMSAccession No. ML1 3203A1 00) has requested additional information. The response tothat request is provided below.Note -an Acronym List is provided at the end of this Attachment. SRXB RAI-2Section 2.8.2.1 of the Power Uprate Safety Analysis Report (PUSAR2) indicates that theaverage bundle power would increase from the current licensed thermal power (CLTP)value of 4.60 Megawatts (MW)/bundle to a value of 5.17 MW/bundle under EPUconditions. This 12.4% change in average bundle power level corresponds to the samepercent increase of total core power from CLTP to EPU conditions. For constantpressure power uprate (CPPU) for boiling-water reactors (BWRs), it is assumed that theadditional core power is obtained by flattening the core power profile (i.e., raising theaverage bundle power, but keeping the peak bundle power the same). However, pastBWR EPU operations have demonstrated that peak bundle power can increase by alimited amount. Please provide the current peak bundle power level and the expectedvalue of peak bundle power for EPU operation at PBAPS.RESPONSEThe peak bundle power for CLTP (3514 MWt) operation is 7.05 MW (based on Unit 2Cycle 19), 7.34 MW (based on Unit 2 Cycle 20), 7.11 MW (based on Unit 3 Cycle 19)and 7.24 MW (based on Unit 3 Cycle 20). The peak bundle power for the representative equilibrium GNF2 core design in the PUSAR (Reference 2-1) is 7.65 MW. Smallvariation is expected depending upon the cycle specific nuclear design and associated operating limits.2 A proprietary (i.e., non-publicly available) version of the PUSAR is contained in Attachment 6 tothe application dated September 28, 2012. A non-proprietary (i.e., publicly available) version ofthe PUSAR is contained in Attachment 4 to the application dated September 28, 2012. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 2References 2-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML1 22860201), Attachments 4 and 6.SRXB RAI-3As stated in the PUSAR, PBAPS EPU analyses assumed a representative "equilibrium" core comprised of GNF2 fuel only. Describe the PBAPS core design for the first EPUcycle. If the actual EPU core will be comprised of other GE fuel designs, for exampleGE14, in addition to GNF2, then justify why using a GNF2 equilibrium core for EPUcalculations provides bounding results for EPU transient and accident

analyses, inparticular the safety limits (minimum critical power ratio (MCPR), linear heat generation rate, maximum average planar linear heat rate, peak cladding temperature (PCT), etc.).RESPONSEThe PBAPS core at the time of EPU implementation for each PBAPS unit is expected toconsist only of GNF2 fuel. The response to SNPB-RAI-1, part d, (Reference 3-1)concerning the PBAPS core design for the first cycle, and the response to SNPB-RAI-7 (Reference 3-1), concerning details of the PBAPS equilibrium core which supported theanalysis for EPU, provides the justification for the use of this equilibrium core for EPUtransient and accident analyses.

References 3-1 Exelon letter to USNRC, "Extended Power Uprate License Amendment Request-Supplement 6 -Response to Request for Additional Information -SNPB",dated July 30, 2013. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 3SRXB RAI-4Pellet clad interaction (PCI) and stress-corrosion cracking (SCC) phenomena can causeclad perforation resulting in leaking fuel bundles and resultant increased reactor coolantactivity. Therefore, the NRC staff requests the licensee to provide the following additional information regarding PCI/SCC for PBAPS at EPU conditions: a) Describe any differences in operating procedures associated with PCI/SCC at EPUconditions versus pre-EPU operations. b) From the standpoint of PCI/SCC, discuss which of the Anticipated Operational Occurrences (AOOs), if not mitigated, would most affect operational limitations associated with PCI/SCC.c) For the AOOs in part (b), discuss the differences between the type of requiredoperator

actions, if any, and the time to take mitigating actions between pre-EPU andEPU operations.

d) If the EPU core will include fuel designs with non-barrier cladding which have lessbuilt-in PCI resistance, then demonstrate by plant-specific analyses that the peakclad stresses at EPU conditions will be comparable to those calculated for thecurrent operating conditions. e) Describe operator training on PCI/SCC operating guidelines. RESPONSEa) There are no differences in operating procedures associated with PCI/SCC atEPU conditions versus pre-EPU conditions. The fuel vendor, Global NuclearFuel (GNF), provides operating recommendations associated with PCI/SCC.Those operating recommendations are the same for both pre-EPU and post-EPUconditions. There is extensive successful operating experience with the GNFoperating recommendations at both EPU and non-EPU conditions. b) From the standpoint of PCI/SCC the AOO that would most affect operating limitations associated with PCI/SCC, if not mitigated, is the Loss of Feedwater Heater AOO. This is a relatively slow transient, however the power increases associated with this event might exceed the ramp rate increases recommended by the 'soft duty guidelines'. This could reduce the margin to fuel failuresassociated with PCI/SCC.

However, with GNF barrier fuel, no fuel failuresassociated with PCI/SCC are expected associated with this AOO; and no fuelfailures in GNF barrier fuel have occurred to date in any plants from a loss offeedwater heating AOO. Furthermore, operating procedures specify a reduction in core recirculation flow early in the event, which reduces the increases in nodalpowers, so that failures, even with non-barrier fuel, are not likely or expected.

EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 4Operating procedures also specify that any control rods (after recirculation flowreduction) be continuously and fully inserted to 00, to avoid the axial powerpeaking that can occur at the tip of partially inserted control rods.c) There are no differences in the type or timing of the operator actions in responseto the Loss of Feedwater (LOFW) AOO between pre-EPU and EPU operation. d) The PBAPS EPU core will only consist of fuel with barrier cladding. e) Operator training on PCI is integrated into the Core Thermal Limits lesson andSCC is covered in the Operations chemistry lesson in the initial licensed andnon-licensed operator training programs. Continuing training includes topicsselected to reinforce fundamental knowledge; PCI is currently included in thebiennial thermal limits review.SRXB RAI-5Characterize the expected amount of bypass voiding under CPPU conditions. Providethe expected bypass void level at points C, D, and E of Figure 1-1 of the PUSAR, usinga methodology equivalent to that used by ISCOR for both hot and average channel.RESPONSEISCOR was used to characterize the expected amount of bypass voiding under CPPUconditions. ISCOR conservatively calculates the hot channel bypass voiding using itsdirect moderator-heating model and provides no credit for cross flow while applyingconservative hot channel bypass heating. Points C, D, and E of PUSAR (Reference 5-2)Figure 1-1 are at the following Power/Flow Statepoints: Table SRXB RAI-5 -PBAPS EPU Bypass Voiding (ISCOR)Core Hot Core HotPower Average Channel Average ChannelPoint (EPU) LPRM LPRM TAF TAF(% Rated) (% Rated) D-Level D-Level Bypass BypassBypass Bypass Voids aVoidsVoids (%) Voids (%)E 100 100D 100 99C 54.9 38For Points E and D, the Core Average and Hot Channel LPRM D-Level Bypass Voidingassociated with the representative equilibrium GNF2 core is shown to be less than 5% EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 5when operating at steady state conditions within the MELLLA boundary. For Point C,the Core Average and Hot Channel Bypass Voiding values at the LPRM D-level areshown above. The Core Average and Hot Channel Bypass Voiding values are alsoshown at the Top of Active Fuel (TAF) location and confirmed to be less than [[ 1]and [[ ]], respectively. These values are not specific

criteria, but ranges for bypassvoiding for the MELLLA operating domain as shown in Sections 5.4, 6.1.1.1, and 6.2 ofthe Reference 5-1.References 5-1 GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains,"

NEDC-33173-P-A, Revision 4, November 2012.5-2 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.SRXB RAI-6Reliability of the local power range monitor (LPRM) instrumentation and accurateprediction of in-bundle pin powers typically requires operation with bypass voids lowerthan 5% at nominal conditions (e.g., point E of Figure 1-1 of the PUSAR). If theexpected bypass void conditions at CPPU are greater than 5%, evaluate the impact on:(1) reliability of LPRM instrumentation, (2) accuracy of LPRM instrumentation, and (3) in-bundle pin powers.RESPONSEPer results noted in Section 2.8.2.4.1 of Reference 6-1 (and in response to SRXBRAI-5), bypass void conditions at EPU are not expected to be greater than 5% at Point Dand Point E of the power/flow map.

References:

6-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 6SRXB RAI-7The presence of bypass voids affects the LPRM calibration. Evaluate the expectedcalibration error on Oscillation Power Range Monitor (OPRM) and Average PowerRange Monitor cells induced by the expected level of bypass voids. Document theimpact of this error on the detect-and-suppress Option III scram setpoint. RESPONSEThe effect of LPRM calibration errors on the OPRM system scram amplitude due tobypass voiding would be less than 5% (see Section 6.2 of Reference 7-1). Thistranslates to an approximate 0.01 difference in OPRM amplitude setpoint. In accordance with the Stability Setpoints Adjustment Limitation in Section 6.2 of theSafety Evaluation in Reference 7-1, a 5% penalty was applied to the calculated OPRMamplitude

setpoint, which translates to an approximate 0.01 decrease.

The OPRMamplitude setpoints presented in Section 2.8.3.1.2 and Table 2.8-2 of Reference 7-2include the 5% setpoint penalty due to LPRM calibration errors.The Average Power Range Monitor (APRM) system is not used for detection andsuppression of thermal-hydraulic oscillations; therefore, there is no effect of APRMcalibration errors on the Detect and Suppress Option III scram setpoint.

References:

7-1 GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains," NEDC-33173P-A, Revision 4, November 2012.7-2 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.SRXB RAI-8PUSAR Table 2.8-2 only shows the Option III Setpoints Demonstration. Please providean example setpoint calculation for the EPU cycle including an uncertainty termreflecting the possible LPRM miscalibration under bypass void conditions. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 7RESPONSETable 2.8-2 of Reference 8-1 shows the Option III setpoint for the EPU equilibrium core.The following is an example of the calculation method used to determine the OPRMsetpoint for the two-pump trip event, which is the more limiting of the two boundingstability events for the EPU equilibrium core.For an OPRM setpoint of 1.12, the hot channel oscillation magnitude was calculated asAh = [[ ]] for PBAPS based on the statistical methodology described in Reference 8-2. The DIVOM slope was calculated as [[ ]] for the PBAPS EPU equilibrium core.In order to protect SLMCPR, the initial MCPR is determined using the following relationship (Section 4.5.1 of Reference 8-2),IMCPR > SLMCPR / (1 -ACPR/ICPR) ACPR/ICPR in the above equation is the product of Ah and DIVOM slope. For the EPUequilibrium core, SLMCPR is 1.09 including the steady-state uncertainties resulting fromLPRM calibration. Therefore, MCPR at the start of the oscillations is calculated asfollows,IMCPR >[It was determined from a PANAC1 1 analysis that this IMCPR after the pump trip will beattained if the MCPR at rated conditions prior to the pump trip, OLMCPR(2PT), isThis OLMCPR(2PT) is to be compared to the transient-based OLMCPR. Per Limitation and Condition 9.19 of Reference 8-3, 0.01 is added to OLMCPR(2PT), OLMCPR(2PT) = [[The OLMCPR(2PT) value of [[ ]] was calculated for an OPRM setpoint of 1.12. Inorder to account for impact of the setpoint uncertainties resulting from bypass voidingdiscussed in the SRXB RAI-7 response, 0.01 is subtracted from the setpoint and theresult is conservatively reported as applying to the reduced setpoint (as reported inTable 2.8-2 of Reference 8-1). Hence, the minimum OLMCPR that can be supported based on a two recirculation pump trip event is [[ ]] for an OPRM setpoint of 1.11.Thus, this OLMCPR value accounts for the LPRM calibration uncertainty due to bypassvoiding. Exelon will apply the above methodology to the EPU implementation cycle coredesign to determine the cycle-specific OPRM setpoint. This setpoint will be reported inthe PBAPS Supplemental Reload Licensing Report. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 8

References:

8-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.8-2 GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, Class I (Non-proprietary), August 1996.8-3 GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains," NEDC-33173P-A, Revision 4, November 2012.SRXB RAI-9The Delta Critical Power Ratio (CPR) over Initial CPR Versus Oscillation Magnitude (DIVOM) slope is not included in PUSAR Table 2.8-2 under CPPU conditions. Pleasedocument which DIVOM slope will be used for future CPPU cycles and whichmethodology will be used to: (1) calculate it, or (2) evaluate the adequacy of an olderslope.RESPONSEThe DIVOM slope calculated for the EPU equilibrium core is [[ ]]. The DIVOMslope for each Peach Bottom Unit 2 and 3 operation cycles is calculated as part of thecycle-specific reload licensing analysis and the DIVOM slope will be evaluated on acycle-specific basis per References 9-1 and 9-2. It is limited to no less than the genericDIVOM slope of 0.45 as prescribed in References 9-2 and 9-3.

References:

9-1 GE Hitachi Nuclear Energy, "Migration to TRACG04/PANAC1 1 fromTRACG02/PANAC10 for Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465 Supplement 1, September 2011.9-2 Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel(GESTAR II), "NEDE-24011-P-A-19 and the US Supplement NEDE-24011-P-A-19-US, May 2012.9-3 GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, Class I (Non-proprietary), August 1996. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 9SRXB RAI-10Assuming a conservative OPRM setpoint of 1.15, provide the hot-spot fuel temperature as a function of time before the scram. Evaluate this fuel temperature oscillation againstPCI limits. Assume that the steady-state fuel conditions before the oscillations are thoseof point Al of PUSAR Figure 2.8-21 (the highest power point in the backup stability protection (BSP) scram region).RESPONSEThe current licensing criteria applicable to SRXB RAI-10 are [[]]. Additionally, the current licensing criterion that cladding fatiguelife usage be less than or equal to 1.0 applies to SRXB RAI-M0. These criteria areaddressed in this response. This response also addresses the issue of the potential forincreased pellet-cladding interaction (PCI) raised in SRXB RAI-10. Because design orlicensing criteria for PCI currently do not exist, the issue is addressed qualitatively interms of impact on reliability. A thermal-mechanical based power-exposure limits envelope is specified [[]] The LHGR limits arespecified to assure compliance with several primary fuel rod thermal-mechanical licensing criteria; these criteria address fuel centerline temperature, [[]], and fuel rod internal pressure. [[A major GNF fuel rod design objective is to specify the LHGR limits curves to achievebalanced margins and a balanced design with high reliability over the rod lifetime. During the core design process, a specified margin is typically maintained between theLHGR limits and the anticipated operation for each bundle. Operation under poweruprate conditions will result in more rods in some bundles operating near the specified margin for a larger fraction of the bundle lifetime, thus increasing the potential for fuelfailure. The potential for increased failure under power uprate conditions is assessed interms of available GNF operational experience and experimental information below. Theimpact of power uprate on thermal-mechanical licensing analyses for the GNF2 fueldesign is also discussed below. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 10]] The resultsfrom this Severe Power Ramp testing, as compared to the LHGR limits curves for thefuel designs noted above, are also provided in Figure 10-1. It is observed from Figure10-1 that significant margin exists to the apparent failure threshold represented by theavailable ramp test results. In addition to barrier fuel's resistance to ramping, ramp ratesat power uprate conditions versus non-power uprate conditions are not appreciably different. Thus it is judged that the possible increased cladding mechanical dutyassociated with operation under power uprate conditions will have negligible impact onthe reliability of GNF fuel. It is further noted that the margin to failure is reasonably well-balanced over the entire exposure range, consistent with the design objective notedabove.In addition to possible increased fuel duty, other potential effects of power uprate aresmall changes in core conditions such as increased coolant pressure (and temperature) and changes in flow conditions. [[1]For instability oscillations indicated in SRXB RAI-10, the incremental fatigue usage dueto the oscillations is negligible in an absolute sense and relative to the margin to the limit(1.0) calculated for the cyclic loading assumed in the fuel rod thermal-mechanical licensing analyses. This criterion is based upon preventing wide spread cladding fatiguefailures during normal operation. The fuel rod time constant is higher than the period ofthe power oscillations. As a result, the power oscillations result in insignificant fueltemperature oscillations relative to the PCI margin shown in Figure 10-1. These resultsindicate that the instability oscillations will have negligible impact on fuel reliability. In summary, on the basis of the generic licensing analyses and the specific analyses toaddress operation under power uprate conditions summarized above, it is concluded that the [[ ]] fuel design is fully compliant with existing licensing requirements foroperation under power uprate conditions. Based upon available operational experience and experimental data, it is also concluded that operation under power uprate conditions will not significantly affect GNF2 fuel reliability. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 11Figure 10-1 LHGR Limits and Severe Ramp Test Failure Data1]References 10-1. H. Sakurai, et. al., 'Irradiation Characteristics of High Burnup BWR Fuels', paperpresented at the ANS Light Water Reactor Fuel Performance Conference held atPark City, Utah, April 10-13, 2000.SRXB RAI-11Describe any effects or impacts of EPU on the long-term stability implementation. RESPONSEThe EPU expands the operating domain in a region of the power to flow map where theplant is not susceptible to thermal-hydraulic instability events. The effects of PBAPSEPU implementation for the Option III stability solution are described in Section 2.8.3 ofReference 11-1. Furthermore, the OPRM setpoints and Backup Stability Protection regions are generated, and the OPRM Trip-enabled region boundaries are confirmed each reload. Therefore EPU does not affect the applicability of the Option III solution. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 12

References:

11-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.SRXB RAI-12For the BSP calculations, describe how the stability curves for the scram region and thecontrolled entry region shown in PUSAR Figure 2.8-21 are calculated for EPUconditions. Specifically, provide the associated feedwater temperature assumptions thatallow the use of the same decay ratio criteria shown in Table 2.8-3 for the Scram andControlled Entry boundary. RESPONSEThe BSP Scram and Controlled Entry region for Option III methodology are calculated inthe fuel cycle reload stability analysis (References 12-1 to 12-3). The samemethodology is applied for EPU. To calculate the BSP Scram and Controlled EntryRegion boundaries, ODYSY decay ratio calculations are performed on the highestlicensed flow control line and on the natural circulation line. Rated feedwater temperature and rated xenon concentrations are assumed for calculating the BSPScram Region boundary points, and the points where a 0.8 core wide decay ratio iscalculated are connected using well defined Shape Function (i.e., Generic or ModifiedShape Function) to define the Scram region boundary. The BSP Controlled EntryRegion is calculated in a similar manner, also using a core wide decay ratio of 0.8 todefine the region boundary; the difference being that the decay ratio calculation of thepoint on the highest flow control line assumes equilibrium feedwater temperature at off-rated operating conditions and xenon concentration (rather than rated), and the point onthe natural circulation line assumes equilibrium feedwater temperature and xenon freeconditions. This is why the two different curves can have almost identical calculated core wide decay ratios.

References:

12-1 "Backup Stability Protection (BSP) for Inoperable Option III Solution", OG 02-0119-260, July 2002.12-2 "ODYSY Application for Stability Licensing Calculations Including Option I-D andII Long Term Solutions," NEDE-33213P-A, April 2009.12-3 Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel(GESTAR II)," NEDE-24011-P-A-19 and the US Supplement NEDE-24011-P-A-19-US, May 2012. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 13SRXB RAI-13Provide plant-specific information relevant to an anticipated transient without scram(ATWS) event under EPU conditions. Specifically, provide the location of the boroninjection, a description of the standby liquid control system actuation logic and itsoperability requirements, boron enrichment level, turbine bypass capacity, and locationof the steam extraction points for the feedwater heaters.RESPONSEThe equipment performance parameters used in the PBAPS EPU ATWS analysis areprovided below:a. The RPV lower plenum is the location for boron injection from the SLC system(SLCS). This is not a change from the current plant configuration.

b. There is no automatic actuation logic for SLCS. Operators manually initiate SLCSvia key lock switches in the main control room consistent with PBAPS emergency operating procedures.

ATWS analysis assumes SLCS is manually initiated atthe later of either: 1) the time of high pressure ATWS RPT plus 120 secondsoperator action time, or 2) the time at which the suppression pool temperature reaches the Boron Injection Initiation Temperature (BIIT). SLCS operability requirements are stated in PBAPS Technical Specification 3.1.7. Note thatrevised SLCS Technical Specifications for EPU are contained in the EPU LARAttachments 2 (Unit 2) and 3 (Unit 3).c. B10 enriched to at least 92%.d. Turbine bypass capacity at EPU rated thermal power is 2.82x106 Ibm/hr, which isunchanged from the turbine bypass capacity at current licensed thermal power.The turbine bypass is not credited in the PBAPS rated power ATWS analysis.

e. Steam extraction points for FW heaters are downstream of the MSIVs, such thatFW heating is lost following isolation.

The steam extraction points are listedbelow: EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 14FWH Extraction Steam Point Location5th Stage FWH HP Exhaust (Cross-around Steam)*4th Stage FWH AS2 stage Low Pressure Turbine*3rd Stage FWH AS3 stage Low Pressure Turbine*2nd Stage FWH AS6 stage Low Pressure Turbine*1st Stage FWH AS8 stage Low Pressure Turbine*Drain Cooler N/A -No Extraction Point* see PBAPS Piping & Instrumentation Drawing M-304SRXB RAI-14Provide a short summary of the Solution III hardware currently installed in PBAPS.Provide justification that the hot channel oscillation magnitude portion of the Option IIIcalculation is not affected by EPU because the OPRM hardware does not change.RESPONSEThe stability Option III hardware for PBAPS is fully integrated into the NUMACTM PowerRange Neutron Monitoring (PRNM) System. The licensing basis for the PRNM retrofit atPBAPS is contained in References 14-1 and 14-2. The Option III Oscillating PowerRange Monitor (OPRM) Channel is integral with each channel of the PRNM. [[]]

References:

14-1 GE Nuclear Energy, "Nuclear Measurement Analysis and Control Power RangeNeutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," NEDC-32410P-A, Volume 1 and 2, October 1995.14-2 GE Nuclear Energy, "Nuclear Measurement Analysis and Control Power RangeNeutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," NEDC-32410P-A Supplement 1, November 1997.14-3 GE Nuclear Energy, "Constant Pressure Power Uprate," NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003; and NEDO-33004-A, Revision 4,Class I (Non- Proprietary), July 2003.'A EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 15SRXB RAI-15Provide a summary of the ATWS emergency operating procedure (EOP) actions. Whatversion of emergency operating guidelines is currently implemented in PBAPS? Providea short description of the process used to ensure that the emergency procedure guideline variables (e.g., hot shutdown boron weight, heat capacity temperature limit,etc.) are adequate under CPPU conditions. RESPONSEThe ATWS related EOP operator actions are summarized as follows:* Manually SCRAM the reactor by placing the mode switch in shutdown.

• Manually initiate alternate rod insertion (ARI) to insert the control rods* Manually trip the reactor recirculation pumps.* Manually initiate boron injection with the SLCS if sustained power oscillations exceed 25% peak to peak.* Manually initiate boron injection with the SLCS before torus temperature reaches110°F.* Perform actions to manually insert the control rods.* Perform manual actions to minimize coolant injection to the Reactor PressureVessel (RPV) in order to lower RPV water level to between below -60 inches andabove top of active fuel until all control rods are inserted or sufficient boroninjection has occurred.
• Inhibit Automatic Depressurization System.* Bypass Main Steam Isolation Valve isolation.

Revision 2 of the BWROG Emergency Procedure and Severe Accident Guidelines (EPG/SAGs) is currently implemented at PBAPS.Calculation revisions to address these EPG variable changes are being made inaccordance with the EGC configuration control process. This process also ensuresimpacted EOPs are updated to reflect the changes from the calculations as applicable. SRXB RAI-16Provide a short description of how the Stability Mitigation Actions (e.g., immediate waterlevel reduction and early boron injection) are implemented in PBAPS. Does operation atCPPU conditions require modification of any operator instructions? EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 16RESPONSEPBAPS implemented the Option III stability solution to address Stability Mitigation Actions. This includes use of the power range neutron monitoring (PRNM) system toprovide a signal to shut down the reactor when a thermal-hydraulic instability (THI)condition is detected. Oscillations in the neutron flux are used as an indicator of THI.The Oscillation Power Range Monitor (OPRM) Upscale Function provides compliance with GDC 12 by providing a hardware system that detects and acts to suppress THIconditions. If a transient occurs (e.g., trip of a reactor recirculation pump at 100% RTP),the PRNM system will automatically trip the reactor when the OPRM trip setpoint isexceeded. If THI conditions are observed, procedures direct the operators to take the following actions:a Manually insert control rods until a THI condition no lonaer exists and monitor0Sindications for THI.SCRAM the reactor if APRM flux oscillations exceed an amplitude of 15% RTP.If a THI condition exists and a reactor SCRAM is unsuccessful (i.e., an ATWSevent), then the operators will respond as follows:* Manually initiate Alternate Rod Insertion (ARI) to insert the control rods.* Manually trip the operating reactor recirculation pump(s).* Manually initiate boron injection with the SLCS if sustained power oscillations exceed 25% peak to peak." Manually initiate boron injection with the SLCS before torus temperature reaches 110°F." Perform actions to manually insert the control rods.* Perform manual actions to minimize coolant injection to the Reactor PressureVessel (RPV) in order to lower RPV water level to between below -60 inchesand above top of active fuel until all control rods are inserted or sufficient boron injection has occurred.

• Inhibit Automatic Depressurization System.* Bypass Main Steam Isolation Valve isolation.

There are no changes to operator instructions for the stability mitigation actionsdiscussed above. However, due to the increase in boron-10 enrichment, EOP operatorinstructions will be revised to reflect a reduction in the percentage of SLC tank volumerequired to be injected by the SLCS to achieve hot shutdown boron weight for EPUconditions. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 17SRXB RAI-17PBAPS currently operates under the Option III solution. Please provide clarification forthe following areas:a) Describe the process that was followed by PBAPS to implement Option III Long-Term Stability Solution and to verify that Option III is still applicable under CPPUoperation. b) Describe the expected effects of CPPU operation on Option Il1.c) Describe any alternative method to provide detection and suppression of any modeof instability other than through the current OPRM scram.d) Provide a summary of the PBAPS Technical Specifications affected by the Option IIIimplementation and future CPPU operation. e) List the approved methodologies used to calculate the OPRM setpoint by the currentoperation and future PBAPS CPPU operation. RESPONSEa) The process followed by PBAPS to implement the Option III Long Term Stability Solution is described in the PRNM LARs (References 17-1 and 17-2) and therelated NRC Safety Evaluation Report (Reference 17-8.) Validity of the Option IIIsolution at EPU conditions has been shown generically in Reference 17-3. TheOption III solution has plant and cycle-specific

features, such as the OPRM Trip-Enabled region, OPRM trip setpoints, and Backup Stability Protection regions.Section 2.8.3 of Reference 17-4 established the basis for the plant-specific feature,namely the OPRM Trip-Enabled region at EPU conditions.

A demonstration analysis for the EPU conditions is also presented in Section 2.8.3 of Reference 17-4. The cycle-specific features are included with the reload analysis. b) The EPU expands the operating domain in a region of the power to flow mapwhere the plant is not susceptible to thermal-hydraulic instability events. Theeffects of PBAPS EPU implementation for the Option III stability solution aredescribed in Section 2.8.3 of Reference 17-4. Furthermore, the OPRM Setpoints and Backup Stability Protection regions are generated, and the OPRM Trip-Enabled region boundaries are confirmed each reload. Therefore EPU does notaffect the applicability of the Option III solution. c) The Backup Stability Protection at PBAPS is discussed in the response to RAI-3for the PRNM LAR (Reference 17-5.) There is no change to the Backup Stability Protection implementation with EPU. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 18d) The changes to the Technical Specifications (TS) due to implementation andactivation of the Option III long-term stability solution for CLTP conditions atPBAPS are described in the related NRC SER (Reference 17-8). EPU affects TSLCO 3.3.1.1 Condition J, SR 3.3.1.1.19 and Table 3.3.1.1-1 Function 2.f. Thechanges to update the PRNMS TS for EPU are described in Section 3.1.8 of theEPU LAR (Reference 17-4, Attachment

1) and in EPU LAR Supplement No. 5(Reference 17-9).e) There are no differences in the methodology used to determine the OPRMsetpoints for either CLTP or EPU conditions; that methodology is as specified inReferences 17-6 and 17-7. However, it should be noted that the setpoint penalties discussed in the SRXB RAI-8 response are applied at EPU conditions.

The OPRMsetpoints are determined each cycle as a part of the reload analysis.

References:

17-1 PECO Energy Company letter to the NRC, "Peach Bottom Atomic Power Station,Units 2 and 3 License Change Request ECR 98-01802," dated March 1, 1999.17-2 Exelon letter to the NRC, "License Amendment

Request, Activation of the TripOutputs of the Oscillation Power Range Monitor Portion of the Power RangeNeutron Monitoring System,"

dated February 27, 2004 (NRC Accession NumberML0407008073.) 17-3 GE Nuclear Energy, "Constant Pressure Power Uprate," NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003; and NEDO-33004-A, Revision 4,Class I (Non- Proprietary), July 2003.17-4 Exelon letter to the NRC, "License Amendment Request -Extended PowerUprate," dated September 28, 2012, Attachments 4 and 6 (NRC Accession No.ML122860201.) 17-5 Exelon letter to the NRC, "Responses to Request for Additional Information, License Amendment

Request, Activation of the Trip Outputs of the Oscillation Power Range Monitor Portion of the Power Range Neutron Monitoring System,"dated September 13, 2004 (NRC Accession No. ML042580401.)

17-6 GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, Class I (Non-proprietary), August 1996.17-7 GE Hitachi Nuclear Energy, "Migration to TRACG04/PANAC11 fromTRACG02/PANAC10 for Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465 Supplement 1, September 2011.17-8 NRC letter to Exelon, "Activation of Oscillation Power Range Monitor Trip (TACNos. MC2219 and MC2220)," dated March 21, 2005 (NRC Accession No.ML050270020.) 17-9 Exelon letter to the NRC, "Supplemental Information Supporting Request forLicense Amendment Request -Extended Power Uprate -Supplement No. 5,"dated June 27, 2013 (NRC Accession No. ML042580401.) EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 19SRXB RAI-18Provide a table of hot channel and core-wide decay ratios at the most limiting state pointfor the last cycles and the proposed CPPU condition. The purpose is to evaluate theimpact of CPPU on relative stability of the plant, and the applicability of Option III toPBAPS under these new conditions. RESPONSEThe core decay ratio and hot channel decay ratio were calculated at the intersection ofthe Natural Circulation Line and High Flow Control Line for the same EPU equilibrium core that was used in the demonstration analyses in Section 2.8.3 of Reference 18-1.The decay ratios were also calculated for the current Peach Bottom 2 Cycle 20 reloadcore design, at the same absolute power / flow values. The results are summarized inthe table below. As can be seen from the table, the difference between the decay ratioscalculated for EPU and CLTP conditions is small. These results are representative ofboth Peach Bottom Units 2 and 3.Also note that the Backup Stability Protection regions are generated for each reload.Therefore, while a change in decay ratio may affect the size of the scram and controlled entry regions, it does not affect the applicability of Option Ill.Rated Power Core Flow Core HotPower Decay Channel(MWt) (MWt) % (MIb/hr) % Ratio DecayRatio3951 1953.1 49.4 32.08 31.3 1.03 0.353514 1953.9 55.6 32.08 31.3 1.08 0.36

References:

18-1 Exelon letter to the NRC, "License Amendment Request -Extended PowerUprate," dated September 28, 2012 (NRC Accession No. ML122860201), Attachments 4 and 6.SRXB RAI-19Describe the effects or impacts, if any, of EPU on suppression pool cooling duringisolation ATWS events and/or EOPs. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 20RESPONSEAs noted in the NRC Safety Evaluation for the GEH Constant Pressure Power UprateLicensing Topical Report (Reference 19-1), "[[f]" These actions are consistent with the BWROG EPG/SAGs. The peak suppression pool temperature response to an ATWS event at PBAPS is lowerat EPU conditions as compared to CLTP conditions due to elimination of Containment Accident Pressure (CAP) credit (Reference 19-2, Table 2.8-8).The EOPs will require revision to incorporate the changes associated with modifications for CAP credit elimination (i.e., the enriched boron-10 modification and the Condensate Storage Tank modification), as described in PUSAR Section 2.11, Human Factors(Reference 19-3, Attachment 4).References 19-1 GE Nuclear Energy, "Constant Pressure Power Uprate," NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003; and NEDO-33004-A, Revision 4,Class I (Non- Proprietary), July 2003.19-2 Exelon letter to the NRC, "License Amendment Request -Extended PowerUprate," dated September 28, 2012, Attachments 4 and 6 (NRC Accession No.ML122860201.) 19-3 Exelon letter to the NRC, "Supplemental Information Supporting Request forLicense Amendment Request -Extended Power Uprate -Supplement No. 5,"dated June 27, 2013 (NRC Accession No. ML042580401.) SRXB RAI-20Please provide a short description of the simulator neutronic core model. Also, providethe schedule to show when the PBAPS simulator will be upgraded for EPU conditions. RESPONSEThe PBAPS simulator neutronic core model is a Studsvik Simulate 3 Real-Time (S3R)model, based upon the Studsvik-Scandpower CASMO-SIMULATE engineering code.S3R is the real-time version of SIMULATE-3K, a best-estimate transient analysis code.The upgrades to the PBAPS simulator for EPU conditions are currently scheduled to becompleted in May 2014 in order to support operator training prior to the EPUimplementation outage (Fall 2014.) EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 21SRXB RAI-21PUSAR Section 2.8.5.7.3 states that the highest calculated PCT for ATWS events is1342 OF, during the Pressure Regulator Failure Open event. The submittal states that:Local cladding oxidation is not explicitly analyzed

because, with PCT lessthan 1600 OF, cladding oxidation has been demonstrated to beinsignificant compared to the acceptance criteria of 17% of claddingthickness.

Therefore, the local cladding oxidation for the PBAPS ATWSevents is qualitatively evaluated to show compliance with the acceptance criteria of 10 CFR 50.46.Please provide a reference to show where cladding oxidation has been demonstrated tobe insignificant when the PCT is less than 1600 OF during ATWS events.RESPONSEThe discussion that follows provides references and discussion for why claddingoxidation has been demonstrated to be insignificant when PCT is less than 16001F.Section 3.4.3 of the Safety Evaluation for Reference 21-1 originally restricted upperbound PCT to 1600°F because: (a) the range of test data submitted as part of the codequalification extended only to 1600°F, and (b) the Monte Carlo Simulation presented inthe SAFER Licensing Topical Report (LTR) was performed over a temperature rangewhere effects such as metal-water reaction are negligible. Reference 21-2 was issuedto remove the 1600°F limitation for the licensing basis PCT. The following is an excerptdescribing the metal-water reaction as a function of temperature: "The metal-water reaction does not become a factor until the claddingtemperatures reach 1700°F and does not become significant until thecladding temperatures exceed 1800°F. When the upper bound PCTapproaches 1800°F (where metal-water reaction is just beginning tobecome significant), the licensing basis PCT will be approaching 2200°Fwhere it would be restricted by the 50.46 limit."The following figure from Reference 21-3 illustrates the Baker-Just zircaloy-water reaction equation used in the SAFER method which demonstrates that claddingoxidation is not significant below 18000F. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 220DATA FROM SU14OLES ZV 3 & ZI 4M00 ASSUMED Z1020 MAXJNMU TEMPERATURE 00- 000 0 00000 000001 0 000 I I I I I18D 9 20M 210D 220D 230D 2400 2000MdAXIMUM TEMPERATURE oF1Fig." 8- f ZrO, "hkk:nm ar a Furwon of MA.timum Twnerawwuu Because the criteria to assure coolable core geometry (2200°F PCT and 17% localcladding oxidation thickness limit) for a loss of coolant accident are also applicable to anATWS, the above references and discussion that demonstrate cladding oxidation isinsignificant when PCT is less than 16001F are also applicable to the ATWS analysis. References 21-1 GE Nuclear Energy, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident Volume III," NEDE-23785-1-PA, Revision 1,(Proprietary), October 1984.21-2 GE Nuclear Energy, "GESTR-LOCA and SAFER Models for the Evaluation of theLoss-of-Coolant Accident Volume III, Supplement 1, Additional Information forUpper Bound PCT Calculation," NEDE-23785P-A, Vol. III, Supplement 1,Revision 1, March 2002.21-3 BWR FLECHT Final Report, "Emergency Cooling in Boiling Water ReactorsUnder Simulated Loss-of-Coolant Conditions," GEAP-13197, June 1971. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 23SRXB RAI-22With respect to overpressure protection (i.e., Section 2.8.4.2 of the PUSAR), if ananalysis was performed for the Turbine Trip with Bypass Failure and Scram on High Flux(TTNBPF) event, as it is required in Table E-1 of ELTR-1, please provide a plotcomparing the pressure transients for the Main Steam Isolation Valve Closure withScram on High Flux and the TTNBPF events. If a TTNBPF analysis was not performed for EPU, then justify why not.RESPONSEAn analysis was performed for the TTNBPF event, as required by Table E-1 of ELTR-1.The comparison plot of the MSIVF and TTNBPF events is provided in Figure RAI-22-1. The MSIVF event is clearly more limiting for both dome and reactor vessel bottompressure. 130120110100900807060504030201001375135013251300127512501225 a2V1200 1175CL1150112511001075105002 3 4Time (s)-0 Vessel Dom~e Presswe (TTTNBPF) .4 Safey Valve Fl-~ (TTNBPF)ReliDef Valve Fl- 0-rNBPF)Z Veseel Bottom, -. ' (T-TNBPF) --Vessel Dome Press,-. (MSIVF)~-Safety Valv -lo (MSIVF)--Relief Valve Flow (MSIVF)--Vessel Bottonm Preaaoe (MSIVF)5 6--1-Figure RAI-22-1: Results Comparison for MSIVF and TTNBPF Events EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 24SRXB RAI-23Do the decay heat removal requirements change between current and EPU power levelsdue to any changes in decay heat load or suppression pool temperature? If so, what arethe new requirements and how does the reactor core isolation cooling (RCIC) systemmeet the new criteria without updating the system performance? RESPONSEAs stated in Section 2.8.4.3.1 of Reference 23-1, the only design requirement of theRCIC system is to maintain sufficient water inventory in the reactor to permit adequatecore cooling following a reactor vessel isolation event accompanied by loss of flow fromthe FW system. The system design injection rate must be sufficient for compliance withthe system limiting criteria to maintain the reactor water level above Top of Active Fuel(TAF) at EPU conditions. EPU does increase the amount of decay heat for the RCICsystem to remove, however the requirements remain unchanged. The RCIC systemdesign capabilities (flow, head, etc.) are sufficient to accomplish this design requirement as demonstrated by analysis. The results of the analysis presented in Section 2.8.5.2.3 of Reference 23-1 demonstrate that the RCIC system meets this design requirement atEPU conditions with no changes to the RCIC hardware or flow capability that currently exists at PBAPS current licensed thermal power level (CLTP).The analysis of events in which RCIC operation may be credited, Appendix R Method A(described in Reference 23-1 Section 2.5.1.4), Station Blackout (described in Reference 23-1 Section 2.3.5), and ATWS (described in Reference 23-1 Section 2.8.5.7), did notassume any increased flow capability of the RCIC system from CLTP. CLTPperformance characteristics of the RCIC system are adequate to mitigate these eventsfor EPU.References 23-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.SRXB RAI-24Table 4.7.1 in the PBAPS Updated Final Safety Analysis Report (UFSAR) shows thatthe RCIC system pump has a design temperature range of 40 OF to 140 OF. Are thereany instances under EPU conditions where the pump would be operating outside of thistemperature range? If so, what are the conditions and how are they addressed for thisEPU? EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 25RESPONSEThe only EPU analyses in which RCIC operation is credited are: Appendix R Method A(described in Reference 24-1 Section 2.5.1.4), Station Blackout (described in Reference 24-1 Section 2.3.5), ATWS (described in Reference 24-1 Section 2.8.5.7) and Loss ofFeedwater Flow Event (described in Reference 24-1 Section 2.8.5.2.3.1). For the loss offeedwater flow event, there is no elevated suppression pool temperature. For theAppendix R, Station Blackout and ATWS analyses, the RCIC pump suction sourcecredited is exclusively from the condensate storage tank, which has a temperature rangeof 40 OF to 140 OF. Therefore, there are no safety analyses for EPU where RCIC wouldoperate outside the design temperature range of 40 OF to 140 OF.References 24-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.SRXB RAI-25What is the effect on net positive suction head for the reactor recirculation system forEPU? The PUSAR stated this result is based on past uprate analyses. Explain the pastanalyses and their relevance. RESPONSEPlant-specific evaluation of the Reactor Recirculation (RR) Pump NPSH was performed for the PBAPS EPU. The RR pump NPSH available at EPU conditions increases to563.03 feet from the CLTP value of 514.71 feet. This increase in NPSH available is dueto the higher feedwater flow contribution as a function of total core flow, resulting incolder recirculation pump flow. The net increase in subcooling increases the RR pumpNPSH available at EPU conditions. Because the maximum core flow does not changeat EPU conditions and the core flow resistance at EPU conditions is only slightlyincreased, the NPSH required for the RR pumps is essentially unchanged from CLTP.Therefore, the RR pump NPSH margin for PBAPS (available NPSH minus requiredNPSH) increases at EPU conditions. The statement, "Based on past uprate analyses, the NPSH required at full power doesnot significantly increase or reduce the NPSH margin because the required increase inrecirculation flow is small," contained in Section 2.8.4.6.1 of Reference 25-1 essentially refers back to the [[ EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 26]] concerning RR pump NPSH. The plant-specific evaluation performed forPBAPS reconfirms that the Reference 25-2 [[]]References 25-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6.25-2 GE Nuclear Energy, "Constant Pressure Power Uprate," NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003; and NEDO-33004-A, Revision 4,Class I (Non- Proprietary), July 2003.SRXB RAI-26Section 2.8.5.6.2.5 of the PUSAR states that the licensing basis PCT is 1925 'F basedon the most limiting Appendix K case, including a variable plant uncertainty term.Please provide further explanation regarding the "plant variable uncertainty term."RESPONSEThe ECCS-LOCA analysis is performed with the SAFER model, which is considered asa representative or non-mechanistic model. Nominal plant operating parameters andconditions are input as the basis of the analysis to calculate the peak claddingtemperature. Conservatisms are then explicitly added to the peak cladding temperature based on nominal conditions so the result assures compliance with the regulatory requirements. This is summarized in an equation in Reference 26-1 Section 3.1 with thepeak cladding temperature based on nominal conditions (PCTNomjnaI) calculated by theSAFER model and conservatisms explicitly added (ADDER) to determine the licensing basis peak cladding temperature (PCTLicensing Basis). The equation from Reference 26-1Section 3.1 is also shown below.PCTLicensing Basis -- PCTNominai + ADDERThe ADDER in the above equation to account for conservatisms is calculated as follows:ADDER2 = [PCTAppendix K -PCTNominal] 2 + Y(6PCTi)2Where:PCTAppendix K = Peak cladding temperature from calculation using Appendix Kspecified models and inputs.PCTNominal = Peak cladding temperature from the nominal conditions case. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 277(6PCT )2 = Plant variable uncertainty term.The nominal case (PCTNominal) is based on the Appendix K case with the most limitingPCT for all analyzed operating conditions, break locations, break sizes and single failure.The ADDER comprises two terms:* The first term, [PCTAppendix K -PCTNominaI] 2, in the ADDER incorporates modelspecifications required by Appendix K that are not already included in thenominal calculation. The Appendix K model specification for calculation ofLicensing Basis PCT using the SAFER model is listed in Table 26-1.* The second term, 7(6PCTi)2, in the ADDER is the plant variable uncertainty term.The intent of the plant variable uncertainty term is to include uncertainties in plantvariables not specifically required in the Appendix K model specifications listed inTable 26-1. Reference 26-1 documents a survey spanning many sources ofvariable uncertainty from which a set of prominent items were justified forinclusion in the standard methodology. The uncertainties in plant variables arefuel product line dependent and listed in Table 26-2.The plant variable uncertainty term is a sum of squares regarding the change incalculated PCT when a single plant variable listed in Table 26-2 is perturbed to an upperbound value while the other plant variables listed in Table 26-2 are at best estimatevalues.Table 26-1 -Appendix K Model Specification for Licensing PCT Using SAFER1971 ANS + 20% Decay HeatMoody Slip Flow Model with discharge coefficients of [[ ]Baker-Just Metal Water Reaction RateTransition boiling allowed during blow down only until cladding superheat exceeds [[102% bundle power and at least 102% core powerTechnical Specification MCPR limitPLHGR consistent with Technical Specification MAPLHGR for selected bundle typeWorst Single FailureFuel Exposure which maximizes PCT or stored energyTable 26-2 -Plant Variables Perturbed I] EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 28Reference 26-1 GE Nuclear Energy, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss of Coolant Accident Volume III," NEDE-23785-1-PA, Class Ill,Revision 1, October 1984.SRXB RAI-27Page 2-396 of the PUSAR states that, independent of the EPU, the licensee will bereplacing the recirculation system pump motor power supplies from motor/generator setpower supplies to adjustable speed drives (ASDs). Section 2.8.5.2.1 of the PUSARdiscusses Loss of External Load and Turbine Trip events with specific evaluations forthe Generator Load Rejection with Steam Bypass Failure (LRNBP) event and the Tripwith Steam Bypass Failure event. Results of the transient

analysis, shown in PUSARTable 2.8-12, indicates that LRNBP is the limiting event with a delta CPR of 0.27.Section 2.8.5.2.1 indicates that, [[]] Please specify the resulting delta CPRIRESPONSEThe delta CPR with the ASDs installed and the EOC-RPT out of service is 0.30. Thisresult cannot be directly compared to the LRNBP delta CPR of 0.27 in Table 2.8-12 ofReference 27-1 to determine the effect of the ASD because the Table 2.8-12 resultconsiders the EOC-RPT in service.

When the ASD with EOC-RPT out of servicedelta CPR (0.30) result is compared to the M/G set with EOC-RPT out of servicedelta CPR (0.30), the effect of the ASD is negligible. This is due to the ATWS-RPToccurring approximately one second into the transient, thus limiting any benefit due tothe ASD, similar to the effect of the EOC-RPT when in service.PBAPS is not planning to install the ASD modification until 2015 for Unit 3 and 2016 forUnit 2. EPU does not rely on this modification, nor is approval of this modification requested. References 27-1 Letter from K. F. Borton (Exelon Generation

Company, LLC) to U. S. NuclearRegulatory Commission, "License Amendment Request -Extended PowerUprate,"

dated September 28, 2012. (ML122860201), Attachments 4 and 6. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 29SRXB RAI-28Recent operating experience has shown that, at a similar BWR/4, the events that followa loss of stator cooling (LOSC) could cause a situation that is limiting with respect to theMCPR. Please explain whether the LOSC has a potential to be CPR-limiting at PBAPS.If the LOSC is non-CPR limiting, explain what design features exist to provide protection from a LOSC. If the LOSC is a CPR-limiting event, please explain what affect the EPUcould have on the severity of the event, and how the EPU safety analyses address theevent.RESPONSEThe LOSC event was evaluated for PBAPS and was determined to not be potentially limiting with respect to the minimum critical power ratio (MCPR). At the high powerconditions (including EPU conditions), the plant is operating closer to the high pressurereactor scram (mitigating scram) and the LOSC is not limiting at these conditions. Atoff-rated conditions, the plant is operating further away from the high pressure scram.Evaluations at these off-rated conditions demonstrate that the PBAPS off-rated criticalpower ratio limits bound the LOSC event.The design feature that caused the MCPR to be limiting for the LOSC for the similarBWR/4 was a recirculation pump trip at the initiation of the LOSC. The LOSC sequenceof events for PBAPS does not have an automatic recirculation system pump trip orrecirculation runback. EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 30ACRONYM LISTACRONYM DEFINITION ACPR Delta critical power ratio2PT Two pump tripADS Automatic Depressurization SystemAOO Anticipated operating occurrence APLHGR Average planar linear hear generation rateAPRM Average power range monitorARI Alternate rod insertion ASD Adjustable speed driveATWS Anticipated transient without scramBIIT Boron injection initiation temperature BSP Backup stability protection BWROG Boiling Water Reactor Owners GroupCAP Containment accident pressureCASMO Computer code nameCLTR Constant Pressure Power Uprate topical report -Reference 2CPPU Constant pressure power uprateCPR Critical Power RatioDIVOM Delta CPR over Initial CPR Versus Oscillation Magnitude ECCS Emergency core cooling systemEGC Exelon Generation CompanyELTRI Extended power uprate topical report -Reference 3ELTR2 Extended power uprate topical report -Reference 6EOC-RPT End of cycle recirculation pump tripEOP Emergency operating procedure EPU Extended power uprateFW Feedwater FWH Feedwater heater EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 31ACRONYM DEFINITION GEH General Electric -HitachiGESTAR-II Core design topical report -Reference 5GExx A fuel type (e.g., GE9, GEl4)GNF Global Nuclear FuelGNF2 A fuel typeGWD/MTU Unit of exposure; gigawatt day per metric ton uraniumHP High pressureICPR Incremental critical power ratioIMCPR Initial minimum critical power ratioISCOR A computer codeLAR License amendment requestLHGR Linear heat generation rateLOCA Loss of coolant accidentLOFW Loss of feedwater LOSC Loss of stator coolingLPRM Local power range monitorLRNBP Generator load rejection with steam bypass failure eventLTR Licensing Topical ReportMAPLHGR Maximum average planar linear heat generation rateMCPR Minimum critical power ratioMELLLA Maximum extended load line limits analysis; currentoperating domainM/G Motor -generator MIb/hr Thousand pounds per hourMSIV Main steam isolation valveMSIVF Main steam isolation valve closure with SCRAM on high fluxMW MegawattMWt Megawatt thermalN/A Not applicable EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 32ACRONYM DEFINITION NPSH Net positive suction headNRC Nuclear Regulatory Commission NUMAC Trademark brand of the Power Range Neutron Monitoring SystemODYSY A computer codeOLMCPR Operating limit minimum critical power ratioOPRM Operating power range monitorPANAC A computer codePBAPS Peach Bottom Atomic Power StationPCI Pellet-clad interaction PCT Pellet clad temperature PLHGR Planar linear heat generation ratePRNM Power range neutron monitoring PUSAR Power uprate safety analysis report -Reference 1RAI Request for additional information (from NRC)RCIC Reactor core isolation cooling systemRPV Reactor pressure vesselRR Reactor recirculation S3R A computer codeSAFDL Specified acceptable fuel design limitSAFER A computer codeSAG Severe accident guidelines SCC Stress corrosion crackingSLCS Standby liquid control systemSLMCPR Safety limit minimum critical power ratioSNPB Performance and Code Review Branch of the NRCSRXB Reactor Systems Review Branch of the NRCTAF Top of active fuelTHI Thermal-hydraulic instability EPU LAR Supplement 9Attachment 2 -Response to RAI -SRXBAugust 22, 2013Page 33ACRONYM DEFINITION TS Technical Specification Turbine trip with bypass failure and SCRAM on high fluxTTNBPF evneventUFSAR Updated Final Safety Analysis Report Attachment 3Peach Bottom Atomic Power Station Units 2 and 3NRC Docket Nos. 50-277 and 50-278Response to Request for Additional Information -AADB EPU LAR Supplement 9Attachment 3 -Response to RAI -AADBAugust 22, 2013Page 1Response to Request for Additional Information Accident Dose BranchBy letter dated September 28, 2012, Exelon Generation

Company, LLC (Exelon)submitted a license amendment request for Peach Bottom Atomic Power Station(PBAPS),

Units 2 and 3. The proposed amendment would authorize an increase in themaximum power level from 3514 megawatts thermal (MWt) to 3951 MWt. Therequested change, referred to as an extended power uprate (EPU), represents anincrease of approximately 12.4 percent above the current licensed thermal power level.Exelon provided a response to an initial request for additional information from the AADBin its EPU LAR Supplement 3, dated May 24, 2013 (ADAMS Accession No.ML13149A145.) The NRC staff has reviewed the information supporting the proposedamendment and by letter dated July 23, 2013 (ADAMS Accession No. ML1 3203A1 00)has requested information to clarify the submittal. The response to that request isprovided below.AADB-RAI-2 In Section 2.1.5 of Attachment 1 to Exelon's letter dated May 24, 2013, the licenseestated that:The EPU Main Steam Line Break (MSLB) Exclusion Area Boundary(EAB) and Low Population Zone (LPZ) atmospheric dispersion factors areupdated using a site-specific X/Q calculation. This differs from the CLB[current licensing basis] MSLB evaluation which used X/Q valuescalculated using guidance from RG [Regulatory Guide] 1.5.Please provide a description of the calculation used for the updated MSLB X/Q values.Include a discussion of how it differs from the CLB MSLB evaluation, a justification for itsuse, and all inputs and assumptions used to make the calculation. RESPONSEAll atmospheric dispersion factors (x/Q) utilized by the PBAPS EPU dose calculations were previously supplied to the NRC Staff during the PBAPS AST submittal (Reference 2-1). The supplied information included code inputs, code outputs, and calc notesdescribing their generation and use.The CLB AST LOCA, CRDA and FHA EAB and LPZ y/Q values were generated by asite-specific PAVAN calculation. For each location, all three values are the same, andfor simplicity, they are called "LOCA" in this response. The CLB AST MSLB EAB and LPZ atmospheric dispersion factors were based uponRG 1.5. EPU LAR Supplement 9Attachment 3 -Response to RAI -AADBAugust 22, 2013Page 2Atmospheric dispersion factors are a function of release point, receptor point, and othersite-specific geography, layout, and meteorological data. Because there is nodependence upon licensed core thermal power, the AST atmospheric dispersion factorsgenerated at CLB are applicable to EPU.For the EPU MSLB dose calculation, additional conservatism was added to the doseresults by applying the higher PAVAN-calculated AST LOCA, 0-2 hr ground release,values rather than the RG 1.5 AST MSLB values. Consequently: (X/Q)EABMSLB,EPU = (X/Q)EAB,LOCA,AST = 9.11 x 10-4 s/m3 and(X/Q)LPZ,MSLB,EPU = (x/Q)LPZ,LOCA,AST = 1.38 x 10-4 s/m3Larger X/Qs are more conservative because the x/Q is a multiplier within the dosecalculation. Therefore, larger y/Qs generate higher dose results, and it is acceptable touse the larger CLB LOCA X/Q values for the EPU MSLB accident. References 2-1. Exelon letter to U. S. Nuclear Regulatory Commission, "License Amendment Request -Application of Alternative Source Term," dated July 13, 2007. Attachment 4Peach Bottom Atomic Power Station Units 2 and 3NRC Docket Nos. 50-277 and 50-278AFFIDAVIT NoteAttachment 1 contains proprietary information as defined by10 CFR 2.390. GEH, as the owner of the proprietary information, hasexecuted the enclosed affidavit, which identifies that the proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. Theproprietary information has been faithfully reproduced in the attachment such that the affidavit remains applicable. GE-Hitachi Nuclear Energy Americas LLCAFFIDAVIT I, James F. Harrison, state as follows:(1) I am the Vice President Fuel Licensing of GE-Hitachi Nuclear Energy Americas LLC(GEH), and have been delegated the function of reviewing the information described inparagraph (2) which is sought to be withheld, and have been authorized to apply for itswithholding. (2) The information sought to be withheld is contained in Enclosure 1 of GEH letter, GEH-PBAPS-EPU-427, "GEH Response to NRC SRXB RAIs 2-14, 17-19, 21-28," dated August16, 2013. The GEH proprietary information in Enclosure 1, which is entitled "GEHResponse to NRC SRXB RAIs 2-14, 17-19, 21-28," is identified by a dark red dottedunderline inside double square brackets. .This sentence is an example..13ý. In each case,the superscript notation

13) refers to Paragraph (3) of this affidavit that provides the basis forthe proprietary determination.

(3) In making this application for withholding of proprietary information of which it is theowner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedomof Information Act (FOIA), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C.Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets(Exemption 4). The material for which exemption from disclosure is here sought alsoqualifies under the narrower definition of trade secret, within the meanings assigned tothose terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass EnergyProject v. Nuclear Regulatory Commission, 975 F.2.d 871 (D.C. Cir. 1992), and PublicCitizen Health Research Group v. FDA, 704 F.2.d 1280 (D.C. Cir. 1983).(4) The information sought to be withheld is considered to be proprietary for the reasons setforth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit intothe definition of proprietary information are:a. Information that discloses a process, method, or apparatus, including supporting dataand analyses, where prevention of its use by GEH's competitors without license fromGEH constitutes a competitive economic advantage over GEH or other companies.

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture,

shipment, installation, assurance of quality, or licensing of a similar product.c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, that may include potential products of GEH.d. Information that discloses trade secret or potentially patentable subject matter forwhich it may be desirable to obtain patent protection.

Affidavit for GEH-PBAPS-EPU-420 Enclosure IPagel1 of 3 GE-Hitachi Nuclear Energy Americas LLC(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted tothe NRC in confidence. The information is of a sort customarily held in confidence byGEH, and is in fact so held. The information sought to be withheld has, to the best of myknowledge and belief, consistently been held in confidence by GEH, not been disclosed

publicly, and not been made available in public sources.

All disclosures to third parties,including any required transmittals to the NRC, have been made, or must be made, pursuantto regulatory provisions or proprietary or confidentiality agreements that provide formaintaining the information in confidence. The initial designation of this information asproprietary information, and the subsequent steps taken to prevent its unauthorized disclosure are as set forth in the following paragraphs (6) and (7).(6) Initial approval of proprietary treatment of a document is made by the manager of theoriginating component, who is the person most likely to be acquainted with the value andsensitivity of the information in relation to industry knowledge, or who is the person mostlikely to be subject to the terms under which it was licensed to GEH. Access to suchdocuments within GEH is limited to a "need to know" basis.(7) The procedure for approval of external release of such a document typically requires reviewby the staff manager, project manager, principal scientist, or other equivalent authority fortechnical

content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside GEH are limited to regulatory bodies, customers, andpotential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the infornation, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements. (8) The information identified in paragraph (2) above is classified as proprietary because itcontains results of analyses performed using the GEH EPU methodology including proprietary technical methods and processes. Development of these methodologies and thesupporting analysis techniques and information, and their application to the design,modification, and processes were achieved at a significant cost to GEH.The development of the evaluation methodology along with the interpretation andapplication of the analytical results is derived from the extensive experience database thatconstitutes a major GEH asset.(9) Public disclosure of the information sought to be withheld is likely to cause substantial hann to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety andtechnology base, and its commercial value extends beyond the original development cost.The value of the technology base goes beyond the extensive physical database andanalytical methodology and includes development of the expertise to determine and applythe appropriate evaluation process. In addition, the technology base includes the valuederived from providing analyses done with NRC-approved methods.Affidavit for GEH-PBAPS-EPU-420 Enclosure 1Page 2 of 3 GE-Hitachi Nuclear Energy Americas LLCThe research, development, engineering, analytical and NRC review costs comprise asubstantial investment of time and money by GEH. The precise value of the expertise todevise an evaluation process and apply the correct analytical methodology is difficult toquantify, but it clearly is substantial. GEH's competitive advantage will be lost if itscompetitors are able to use the results of the GEH experience to normalize or verify theirown process or if they are able to claim an equivalent understanding by demonstrating thatthey can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to thepublic. Making such information available to competitors without their having beenrequired to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these veryvaluable analytical tools.I declare under penalty of perjury that the foregoing affidavit and the matters stated therein aretrue and correct to the best of my knowledge, information, and belief.Executed on this 16th day of August, 2013.James F. HarrisonVice President Fuel Licensing GE-Hitachi Nuclear Energy Americas LLC3901 Castle Hayne RdWilmington, NC 28401james.harrison@ge.com Affidavit for GEH-PBAPS-EPU-420 Enclosure IPage 3 of 3}}