ML14181A538
| ML14181A538 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/05/2014 |
| From: | Indiana Michigan Power Co, Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2014-42, LTR-PL-14-22, TAC MF2916 | |
| Download: ML14181A538 (16) | |
Text
ENCLOSURE 5 TO AEP-NRC-2014-42Attachment #2 (NP-Attachment) of Westinghouse Letter, LTR-PL-14-22,Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit I -Request for Additional Information on the Application for Amendment to RestoreNormal Reactor Coolant System Pressure and Temperature Consistent withPreviously Licensed Conditions (TAC No. MF2916)," dated May 28, 2014RAIs: SCVB RAI-la&b, SCVB RAI-2a-c, SCVB RAI-3a&b, SCVB RAI-4a-f, SCVBRAI-5a, SCVB RAI-6, SCVB RAI-7, SCVB RAI-8, SCVB RAI-12, SNBP RAI-1, SNBPRAI-2, SNBP RAI-3a&bNP-Attachment(Non-Proprietary)
Westinghouse Non-Proprietary Class 3LTR-PL-14-22, NP-AttachmentWestinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit I -Request for Additional Information on the Application for Amendment toRestore Normal Reactor Coolant System Pressure and TemperatureConsistent with Previously Licensed Conditions (TAC No. MF2916)" Set #2SCVB RAI-la & b; SCVB RAI-2a-c; SCVB RAI-3a & b; SCVB RAI-4a-f; SCVB RAI-Sa;SCVB RAI-6; SCVB RAI-7; SCVB RAI-8; SCVB RAI-12; SNBP RAM-i; SNBP RAI-2;SNBP RAI-3a & b.NP-Attachment(Non-Proprietary)Westinghouse Electric Company LLC1000 Westinghouse DriveCranberry Township, PA 1606602014 Westinghouse Electric Company LLCAll Rights ReservedNP-1 LTR-PL-14-22, NP-AttachmentSCVB RAI-NReference 1, Enclosure 6, Section 5.4.1.5: it is noted that the proposed vessel/core inlet temperature514.6°F is greater than the Analysis of Record (AOR) vessel/core inlet temperature of 506.6*F and theproposed RCS pressure 2317 psia (includes uncertainty) is greater than the AOR RCS pressure of 2 100psia.a) Please justify why the most limiting Loss of Coolant Accident (LOCA) short term Mass andEnergy (M&E) releaseand containment response for the proposed minimum vessel/core inlettemperature 514.601 and pressure of 2317 psia (includes uncertainty) is bounded by the LOCAshort term M&E release and containment response in the AOR vessel/core inlet temperature of506.6°F and RCS pressure of 2100 psia.WEC Response -Short term M&E releases are strongly influenced by two key inputs;RCS temperature and RCS pressure. Short term blowdown releases are linked directly tocritical mass flux, which is maximized with increasing pressure and decreasingtemperature. The direction of conservatism is high for RCS pressure, and low for RCStemperature (increased fluid density increases short term break flow). The analysis ofrecord shoit term mass and energy releases were calculated to be bounding for both D.C.Cook Unit I and Unit 2, and the vessel/core inlet temperature of 506.6°F and RCSpressure of 2317 psia are thus bounding values.b) Please explain why subtracting (instead of adding) uncertainty from the realistic value would givea conservative input value of 514.6'F for the vessel/core inlet temperature and would giveconservative results for the M&E release.* WEC Response -Due to the increased density of the RCS fluid, low RCS temperaturesand high RCS pressures are the limiting condition for short term LOCA mass and energyreleases.SCVB RAI-2Reference 1, Enclosure 6, Section 5.4.2.2:a) Describe the most limiting LOCA break in the AOR from the containment response standpoint.Explain why the most limiting break in the AOR would also be the most limiting break for theproposed RCS NOP/NOT conditions.* WEC Response -The most limiting break in the AOR, which is bounding for D.C. CookUnit 1 and Unit 2, is the double ended pump suction break with a loss of offsite powerand minimum safeguards assumptions (i.e. one train of emergency diesel generatorfailure to start). This scenario maximizes mass and energy releases, while minimizing thecontaimnent active heat removal via containment spray. The analysis of recordconsidered the most limiting conditions for both Unit 1 and Unit 2, and as such wasperformed at nominal T.,, and RCS pressure plus any applicable uncertainties. Analyzingat NOP/NOT conditions will not result in a more limiting break location or limitingsingle failure.NP-2 LTR-PL-14-22, NP-Attachmentb) The UFSAR Section 14.3.4.1.3.1.3 input assumption 7 states that the air recirculation fan iseffective 132 seconds after the high-I contaimnent pressure bistable signal is actuated. Pleaseexplain the basis for changing this time to 300 seconds.WEC Response -The change in air recirculation fan activation time was driven by theLOCA Peak Clad Temperature (PCT) analysis. The effect of the delay was evaluated forthe LOCA containment integrity case, and shown to have a negligible impact on the peakcalculated contaimnent pressure.p) Please explain the basis for the containment spray actuation time of 315 seconds in the proposedevaluation. What is the containment spray actuation time in the AOR? In case the AOR sprayactuation time is different from 315 seconds, please justify that the change if the change is lessconservative.* WEC Response -The contaimnent actuation time in the AOR was 115 seconds. Thechange in containment spray activation time was driven by the LOCA PCT analysis. Theeffect of the delay was evaluated for the LOCA containment integrity case, and shown tohave a negligible impact on the peak calculated containment pressure.SCVB RAI-3Reference 1, Enclosure 6, Section 5.4.2.4 states:"For the contaimnent integrity analysis, this was completed by evaluating the effects of increased delaytimes for CTS actuation and containment air recirculation fan actuation on the LOCA containmentintegrity analysis."a) Explain the method of evaluation described in the above'statement.* WEC Response -Sensitivity studies on the effect of the delay in CTS and containment airrecirculation fan were completed using the licensed LOTIC 1 containment responsecomputer code (WCAP-8354-P-A). These sensitivity studies showed a negligible changein peak containment pressure.b) If the subject evaluation/analysis methodology is different from the currently used methodology,provide justification.* WEC Response -The methodology used for the evaluation was not different from thecurrently used methodology for ice condenser containment pressure calculations. Thismethodology is documented in WCAP-8354-P-A.NP-3 LTR-PL-14-22, NP-AttachmentSCVB RAI-4Reference 1, Enclosure 6, Section 5.4.2.5;a) The AOR vessel/core inlet temperature stated in the table is 552.5°F. Please explain why this isdifferent firom the AOR vessel/core inlet temperature of 506.6°F stated in Section 5.4.1.5.e. WEC Response -Section 5.4.1.5 considered short term mass and energy releases andcontainment response, while Section 5.4.2.5 considered long term mass and energyreleases and containment response. The direction of conservatism in the RCS temperatureis low for short term M&Es, and high for long term M&Es.b) Please explain why the RCS pressure of 2100 psia in the AOR and 2250 psia (2317 psia includinguncertainty) in the proposed change in NOP/NOT are not included for comparison as key inputparameters.* WEC Response -The analysis of record for long term mass and energy releases andcontainment response was bounding for D.C. Cook Unit I and Unit 2. Therefore, theAOR considered an RCS pressure of 2317 psia and was effectively completed atNOP/NOT conditions relative to D.C. Cook Unit 1, and thus not included as a parameterfor comparison in Section 5.4.2.5.c) Please justify why the AOR is bounding for both M&E release and containment response eventhough the proposed RCS pressure of 2317 psia (uncertainty included) is significantly greaterthan the AOR RCS pressure of 2100 psia.* WEC Response -The analysis of record for D.C. Cook Unit I and Unit 2 is a singlebounding analysis. This analysis was performed considering an RCS pressure of 2317psia.d) Aside froim the key parameters stated in the table, what are other input parameters for long termM&E analysis that differ in the AOR and the proposed analysis. Please provide these values.* WEC Response -An additional parameter not listed in Section 5.4.2.5 is core power. TheAOR core power was 3481 MWt, as compared to the NOP/NOT core power of 3327MWt.e) Aside from the key parameters stated in the table, what are other input parameters, for long termcontainment gas temperature response for equipment environmental qualification (EEQ) thatdiffer in the AOR and the proposed analysis. Please provide these values.* WEC Response -Westinghouse provides long term containment response results to AEPfor EEQ consideration. The long term containment temperature is largely a function ofdecay heat and containment heat removal systems. The increased delay time in the CTSand containment air recirculation fans were shown to have a negligible effect on thecontainment pressure/temperature transient. The decay heat resulting from the initial corepower in the AOR of 3481 MWt is bounding relative to the Unill core power of 3327MWt.NP-4 LTR-PL-14-22, NP-AttachmentSCVB RAI-5Reference 1, Enclosure 6, Section 5.4.2.6 states:"The evaluation of the long-term LOCA M&E andpeak containment pressure is predicated upon thecontinued application of the operability assessment supporting NSAL-11-5 (Reference 3), in confimctionwith the AOR."a) Is the AOR based on corrected M&E release, resolving the issues identified in NSAL-1 1-5(Reference 2)?* WEC Response -The AOR is not based on corrected M&E releases relative to the issuesidentified in NSAL-11-5.NP-5 LTR-PL-14-22, NP-AttachmentReferences Cited in RAI Transmittal1) Letter from I&M to NRC dated October 8, 2013, "Donald C. Cook Nuclear Plant Unit IDocket No. 50-315 License Amendment Request Regarding Restoration of NormalReactor Coolant System Operating Pressure and Temperature Consistent WithPreviously Licensed Conditions" (ADAMS Accession No. M L13283A121).2) NSAL-1 1-05, "Westinghouse LOCA Mass and Energy Release Calculation Issues," July26, 2011 (ADAMS Accession No. ML1 3239A479).NP-6 LTR-PL-14-22, NP-AttachmentSCVB RAI-4Reference 1, Enclosure 6, Section 5.4.2.5;f Please explain why the NOP/NOT core stored energy of[ ]'c is less thanthe AOR core stored energy of[ f, even though the average proposed RCStemperature off I`.C is greater than the average RCS AOR temperature off ]OCWestinghouse Response:The D. C. Cook Unit I Analysis of Record (AOR) core stored energy value, [ ]a'cwas previously calculated using [TheNormal Operating Pressure / Normal Operating Temperature (NOP/NOT) core stored energy value for D. C.Cook Unit 1, [ ]",, was calculated using []". So while a higher Reactor Coolant System (RCS) temperature would increase the overall corestored energy, [ ]PC.NP-7 LTR-PL-14-22, NP-AttachmentSCVB RAI-6"Reference 1, Enclosure 6, Section 5.5.1.4:Explain the basis for selecting the break sizes 1.4 ft2, 0.865 ft2, 0.857fet, 0.834fte, 0.808ft2, and 1.Oft2and their corresponding power levels at which the W&E release analysis is performed."ResponseThe safety analysis methodology documented in Reference 1 describes the basis for the choice ofthe break sizes and the initial power levels analyzed for the SLB M&E releases inside containment.Section 2.1 of Reference 1 discusses the break spectrum for the SLB M&E safety analysis. Thisdescription includes four plant power levels of 102%, 70%, 30%, and 0% of nominal full load andthree different break sizes: a full double-ended rupture (DER), a small DER, and a small split rupture.The discussion in Section 5.5.1.2 of Enclosure 6 presents the D. C. Cook Unit 1 plant-specific breakspectrum based on the approved methodology. The break size of the full DER is defined as the 1.4 ft2cross-sectional area of the flow restrictor integral with the discharge nozzle of the faulted steamgenerator. This is the break size in the forward-flow direction. The area of the steam pipe defines thebreak size in the reverse-flow direction; this area is modeled as 4.7465 ft2 for the initial blow down.The 1.0 ft2 small DER is analyzed only at.0% power initial conditions assuming no entrainment to showthat this break size is more limiting than the full DER at 0% power with entrainment (see the responseto SCVB RAI-7).Each of the split rupture break areas has been determined as the largest cross-sectional area that doesnot produce a steamline isolation signal from the primary protection equipment nor results in waterentrainment in the break effluent as discussed in Section 2.1 of Reference 1. These areas weredetermined for each initial power level based on the D. C. Cook Unit 1 plant-specific values for thesecondary-side protection system setpoints.Reference(s)1. WCAP-8822 (Proprietary) and WCAP-8860 (Nonproprietary), "Mass and Energy ReleasesFollowing a Steam Line Rupture:' September 1976; WCAP-8822-S1-P-A (Proprietary) andWCAP-8860-S1-A (Nonproprietary), "Supplement 1- Calculations of Steam Superheat InMass/Energy Releases Following a Steam Line Rupture," September 1986.SCVB RAI-7"Reference 1, Enclosure 6, Section 5.5.1.5 states:'All of the analyzed breaks conservatively assumed dry saturated steam releases (noentrainment) except the full DER at 0 percent initial power. As a result, the small DER withdry saturated steam release was analyzed at 0 percent power, represented by a 1.oft2 break(smaller than the area of a single integralflow restrictor) from the faulted-loop SG and a1.Oft2 breakfor the reverse-flow blowdown from the intact-loop SGs.'Please explain why entrainment was assumed in the full DER at zero percent initial power, and explainwhy as a result the small DER with dry saturated steam release was analyzed at zero-percent power."ResponseCalculations have been performed and documented for the full DER at 0 percent power with noentrainment similar to the other full DER SLBs identified in Section 5.5.1.4 of Enclosure 6. However,the containment response for D. C. Cook Unit I with the 0 percent power no entrainment M&E releasesNP-8 LTR-PL-14-22, NP-Attachmentproduced a peak temperature that exceeded the 324.7=F limit noted in Section 5.5.2.3 of Enclosure 6.The full DER at 0 percent power with entrainment was assumed In order to reduce the energy contentof the steam effluent during the initial blowdown. The peak containment temperature for this break isless than 321°F as shown in Table 5.5.2-1 of Enclosure 6. However, to satisfy the requirement thatthe maximum release rate has been considered in the M&E releases as stipulated in Reference 2,the 1.0 ft2 small DER has been analyzed with no entrainment to show that that the peak containmenttemperature does not exceed the limit value as shown in Table 5.5.2-1 of Enclosure 6.Reference(s)2. NUREG-0800, Standard Review Plan. Section 6.2.1.4, "Mass and Energy Release Analysis forPostulated Secondary System Pipe Ruptures," Revision 2 -March 2007.SCVB RAI-8"Reference 1, Enclosure 6, Table 5.5.1-1;Explain what is meant by the title of the fourth column 'Rod Motion (sec). Describe its method ofcalculation, and how it affects the M&E release."ResponseRod Motion is the nomenclature used in the table of the sequence of events to indicate the time ofreactor trip. Following a finite delay after the Reactor Trip Signal (third column in Table 5.5.1-1 ofEnclosure 6), the control and shutdown rods are released from the grippers and begin to fall into thecore. The protection system signals that actuate the reactor trip are the low steam pressure setpointfor the full DERs and the high-1 containment pressure setpoint for the small DER and split ruptures asdescribed In Section 5.5.1.5 of Enclosure 6.The method of calculation is to input the D. C. Cook Unit 1 plant-specific protection setpoints and delaysinto the protection model for the SLB safety analysis. As the protection functions are actuated in thetransient SLB M&E analysis, reactor trip and rod motion occurs following the applicable delays for thespecific protection function. It Is not the motion of the rods that is significant to the analysis of the SLBM&E releases, but the occurrence of the reactor trip, which reduces the core power and thus the long-term energy of the reactor coolant system.NP-9 LTR-PL-14-22, NP-AttachmentSCVB RA- 12) NUREG-0800, Standard Review Plan 6.2.1.5 describes the minimumcontainment pressure analysis for the ECCS performance capability. RegulatoryGuide (RG) 1.157, Section 3.12.1 provides guidance for calculating thecontainment pressure response used for evaluating cooling effectiveness duringthe post-blowdown phase of a LOCA. The RG states that the containmentpressure should be calculated by including the effects of containment heat sinksand operation of all pressure-reducing equipment assumed to be available. Usingthe above guidance please describe the impact of the changes in M&E input onthe minimum containment pressure analyses for ECCS performance during aLOCA and MSLB accident.LOCA Response:A conservatively low containment pressure for the Return to Reactor Coolant System (RCS)Normal Operating Pressure (NOP) / Normal Operating Temperature (NOT) large-break loss-of-coolant accident (LBLOCA) evaluation was calculated in accordance with Sections 11-4-11 and12-3-4 of the generically approved ASTRUM evaluation methodology (Reference 1). Thepressure was calculated using the generically approved LOTIC2 containment pressure model(References 2 and 3) and mass and energy (M&E) release from WCOBRA/TRAC at NOP/NOTconditions.The modeling assumptions related to the containment heat sinks and the operation of allpressure-reducing equipment was calculated consistent with References 2 and 3. Themodeling of the air recirculation fan delay time, the containment spray initiation delay time, andthe containment spray flow were updated as shown in Table 5.1.1-3 of Enclosure 6 toReference 4. The M&E release from WCOBRAITRAC captures the effect of the changes to theoperating conditions and plant configuration listed in Table 5.1.1-3 of Enclosure 6 to Reference4. Input assumptions not affected by the change in operating conditions and plant configurationwere unchanged from the D.C. Cook Unit I Analysis-of-Record (AOR) (Reference 5). Thecombined effect of these changes was an increase to the predicted minimum containmentbackpressure.References:1. WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Usingthe Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.2. WCAP-8354-P-A, Revision 0, Supplement 1, "Long Term Ice Condenser Containment Code-LOTIC Code," April 1976.3. WCAP-8339, Revision 0, "Westinghouse Emergency Core Cooling System Evaluation Model-Summary," June 1974.4. Letter from J. P. Gebbie (i&M) to NRC Document Control Desk, "Donald C. Cook NuclearPlant Unit 1, Docket No. 50-315, License Amendment Request Regarding Restoration ofNormal Reactor Coolant System Operating Pressure and Temperature Consistent WithPreviously Licensed Conditions," October 8, 2013 (Agencywide Documents Access andManagement System (ADAMS), Accession No. ML13283A121).5. Letter from T. A. Beltz (NRC) to M. W. Rencheck (I&M), "Donald C. Cook Nuclear Plant, Unit1 -Issuance of Amendment to Renewed Facility Operating Licensing Regarding Use of theWestinghouse ASTRUM Large Break Loss-of-Coolant Accident Analysis Methodology (TACNo. MD7556),'.' October 17 2008 (Agencywide Documents Access and Management System(ADAMS), Accession No. ML082670351).NP-10 LTR-PL-14-22, NP-AttachmenlRequest for Additional Information SNPB RAM-I:Table 2.1-1 of WCAP-17762-NP contains NSSS design parameters which, as described in Section2.1.2, "are used as the basis for the design transients and for the systems, structures, components,accidents and fuel analyses and evaltations." Each of the eight cases listed in Table 2.1-1 has adifferent steam outlet pressure listed, rangingfirom 618 to 851 psia.The description of the NSSS design transient evaluations in Section 3.1, how'ever, states that,consistent with the measurement uncertainty recapture (MUR) uprate license amendment issuedby the NRC on December 20, 2002 (ADAMS Accession No. ML0234 70126), "the full power steampressure will continue to be limited to a minimum of 679 psia (administratively limited to 690 psiafor conservatism). " These limits are higher than the steam outlet pressures listed in Table 2.1-1for Cases 1, 2, and 6.Please clarify how the limits fr'om Section 3.1 listed above interact with the design parametersfi'om Table 2.1-1for the design transient evaluations.Response:The NSSS design parameters in Tables 2.1-1 and 2.1-2 of WCAP-17762-NP provide a range ofmaximum and minimum design conditions based on the program-specified full powerthermal design conditions (e.g., NSSS power, feedwater temperature, desired Tavg range, thermaldesign flow, RCS pressure, and Steam Generator Tube Plugging (SGTP) range) to be used byvarious downstream analyses. The NSSS Design Transients in Section 3.1, which were providedas input to the structural design and fatigue/stress analyses for the NSSS components, bound alleight cases in Tables 2.1-1 and 2.1-2. However, to ensure that the steam generator primary tosecondary Ap limit is not exceeded during the limiting transients, the minimum full power steampressure is limited to 679 psia.In general, any limitations to the NSSS design parameters are provided in the section thatdocuments the basis for the restriction. The limitations are not cross-referenced in the NSSSDesign Parameter section such that all analyses start with the same bounding range of conditions.NP-1I LTR-PL-14-22, NP-AttachmentSNPB RAI-2Section 5.1.1 of WCAP-1 7762-NP describes the best-estimate large break loss-of-coolant accident(BE LBLOCA) analysis for the DC Cook Unit I return to NOP/NOT program. This analysis estimatesthe impact of fuel thermal conductivity degradation (TCD) by using a method found acceptable byNRC staff in a letter dated March 7, 2013 (ADAMS Accession No. ML 13077A 137). In this method,margin to the 10 CFR 50.46(b)(1) peak cladding temperature limit of 2200'F is recaptured by adjustinginput parameters, including accounting for peaking factor bumdown in the fuel.Table I of the previous TCD evaluation for DC Cook Unit 1, from March 19, 2012 (ADAMS AccessionNo. ML12088A 104), included burnup-dependent limits for both heat flux hot channel factor, Fq, andthe enthalpy rise hot channel factor, FAH. In WCAP-17762-NP, Table 5. 1. 1-1 presents the FQ and FAHburndown as a function of rod burnup used in the LBLOCA analysis. The AOR (available in ADAMS atAccession No. ML080090268) accounts for neither TCD nor peaking factor bumdown, and providessingle limits for FQ and FAH.If peaking factor burndown is required to recapture margin to the PCT limit of 2200°F for the return toNOP/NOT analysis, how will this be addressed in the CNP Unit I Core Operating Limits Report and/ortechnical specifications?Westinqhouse InputThe following peaking factor changes were included in the Normal Operating Pressure (NOP)/NormalOperating Temperature (NOT) project for D. C. Cook Unit 1:0 reduction in transient Fa (heat flux hot channel factor), including uncertainties, from 2.15 to2.09a reduction in steady-state FQ, without uncertainties, from 1.70 to 1.65* reduction in FAH (enthalpy rise hot channel factor), including uncertainties, from 1.545 to 1.530, corresponding reduction in hot assembly average power, including uncertaintiesOf the fuel peaking factor design values, the transient FQ and FAH parameters have specific limitsspecified in the Core Operating Limits Report (COLR). Additionally, the Engineering procedurecontrolling the process for Reactor Core Design is being updated to include the peaking factorburndown values to ensure compliance with the revised peaking factor limits in future core designs.Verification of the peaking factor limits is performed as part of the normal reload design process foreach core reload. Steady state FQ limits are verified by comparing predicted steady-state peakingfactors at full power conditions against the steady-state burnup dependent F0 peaking factor limits.Transient FQ and FAH limits are verified by comparing the predicted power distributions during normaloperation and the operational transients against the applicable burnup-dependent limits. Transientpower distributions are generated based on the methodology described in Reference 1. Predictedpower distribution used in reload analyses are based on core models developed using the NRC-approved ANC code described in Reference 2. Each reload cycle, these limits are analyticallyconfirmed and verified. If the analytical verification produces unacceptable results, then the core iseither redesigned or a Large Break Loss of Coolant Accident (LBLOCA) analysis re-assessment forNP-12 LTR-PL-14-22, NP-Attachmentthermal conductivity degradation (TCD) is performed with revised peaking factor input. Theacceptability of analysis results is based on confirming that the reactor core is operating as designed.The plant-specific burnup-dependent peaking factor limits used in the LBLOCA TCD evaluations wereestablished with consideration of current operating cycles. The reduced peaking factors at high burnupconditions (or burndown limits) constitute a core design constraint and will be confirmed during thereload design process. The peaking factor burndown limits defined to support the LBLOCA peakcladding temperature (PCT) evaluations reflect the physical phenomenon of lower rod power as thefuel rod becomes less reactive, and are within the practical application of the core design. The plant-specific burndown limits were developed based upon review of a neutronic model's peaking factorbehavior versus fuel rod burnup for actual cycle-designs for each plant. The burndown limit lines forthe higher burnup fuel were defined such that there was slightly more margin (or "white space") to thepredictive peaking factor data, as compared to that defined for the low burnup fuel (where noburndown credit is taken). By preserving more core design margin in the higher burnup region, thelimit lines are more conservative for the purposes of a LBLOCA PCT evaluation. From a TechnicalSpecification Surveillance perspective, this results in the lower burnup fuel to be more limiting,implicitly enveloping the higher burnup fuel. Therefore, no further Technical Specification Surveillanceactions or COLR changes are required; the burndown credits supporting the LBLOCA PCTevaluations can be confirmed analytically as part of the normal core design process.Reactivity and power distribution measurements are performed periodically during the cycle asrequired by Technical Specifications (TS) 3.1.2 (Core Reactivity), TS 3.2.1 (Heat Flux Hot ChannelFactor (FQ(Z))), and TS 3.2.2 (Nuclear Enthalpy Rise Hot Channel Factor (FrNAH)) to verify that corereactivity and peaking factors are within their respective design limits. Measured power distributionsand core reactivity are also compared against predicted power distributions and core reactivity. Thesecomparisons, when coupled with startup physics testing results following refueling, are used to verifythe core is operating as designed. This confirmation provides verification that the LBLOCA accidentanalysis input is within the specified limits. If the core is determined not to be operating as designed,an evaluation would be performed to assess analysis margins, understand the reasons for thedeviation, and make appropriate adjustments on a case-by-case basis to plant operations or setpointsto ensure operation remains within LBLOCA analysis limits.References:1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.NP-13 LTR-PL-14-22, NP-AttachmentSNPB RAI-3Sections 6.2.3 and 6.2.4 of WCAP-17762-NP discuss the evaluation offiuelperfornance at the Cook Unit Irehtrn to ANOP/NOT conditions. Section 6.2.3provides the acceptance criteria for the fitel, and Section 6.2.4provides the results of the evaluations with respect to these criteria.a. It is stated in Section 6.2.4 that "fnlo explicit PAD calculations were used to evaluate the fuel roddesign criteria at NOP/NOT conditions. " Discussion in several of the evaluations notes that the effectsof returning to NOP/NOT at Cook Unit I will be offset by available margin. Was a quantitative marginassessinent petformed?b. It is also noted in Section 6.2.4 that "it]he PAD code with USNRC-approved models.., for in-reactorbehavior is used to calculate the fitel rod peiformance over its irradiation history. " The NRC-approvedcurrent version of the PAD code, PAD 4. 0, does not include approved method for modeling TCD.In evaluating the fitel acceptance criteria for the return to NOP/NOT conditions, iwas considerationgiven to the effects of TCD onfitelpeiformance? If a quantitative margin assessment was peiformed perSNPB-RAI-3.a, did this margin assessment include the effects of TCD?Westinghouse Response:While no explicit, quantitative calculations were performed to evaluate the fuel rod design criteria, a qualitativeassessment was done as part of the Engineering Report (Reference 2) to compare the estimated impact ofNOP/NOT operation with the adequate available margin for recent cycles of D. C. Cook Unit 1. It wasdetermined that the available margin for D. C. Cook Unit I is sufficient to offset both the estimated impacts oftransition to NOP/NOT operation, as well as the maximum impacts of thermal conductivity degradation (TCD),which have been calculated generically. Additionally, all fuel rod design criteria are evaluated on a cycle-specific basis based on models developed by Nuclear Design, as described in Section 6.1 of Reference IEnclosure 6, so they are ensured to remain met each cycle.ReferencesI. Nuclear Regulatory Commission Agencywide Documents Access Management System (ADAMS)Accession No. MLI3283A122, "WCAP-17762-NP, Revision 1, 'D. C. Cook Unit 1 Return to ReactorCoolant System Normal Operating Pressure/Normal Operating Temperature Program -LicensingReport,' and Enclosures 7 through 9."2. Westinghouse Topical Report WCAP-17761-P, "D. C. Cook Unit 1 Return to Reactor Coolant SystemNormal Operating Pressure/Normal Operating Temperature Program -Engineering Report."NP-14 LTR-PL-14-22, NP-AttachmentSupplemental InformationPer Mr. Don Hafer's request, the following non-proprietary Figure 5.1.1-5 from the D.C. CookUnit 1 NOP/NOT engineering report (WCAP-1 7761-P [21) is to be provided with AEP's responseto EEEB RAI-1.Original TCD Evaluation..... Return to ,OP/NOT Eval Iation26-24 .....................................-' ' 2 2 -i .......................................23coCLo _ 2 0 .................................................................1610 100 200 300 400 500Time After Break (s)NP-15