ML12205A358

Crystal River, Unit 3 - Attachment D to 3F0712-03, Technical Report, ANP-3114(NP), Rev. 0, CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies.

ML12205A358
Person / Time
Site:	Crystal River
Issue date:	05/31/2012
From:	AREVA NP, Progress Energy Florida
To:	Office of Nuclear Reactor Regulation
References
TAC ME6527, 3F0712-03 ANP-3114(NP), Rev. 0
Download: ML12205A358 (39)

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FLORIDA POWER CORPORATIONCRYSTAL RIVER UNIT 3DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72ATTACHMENT DTECHNICAL REPORT ANP-3114(NP):CR-3 EPU -FEEDWATER LINE BREAK ANALYSISSENSITIVITY STUDIES,REVISION 0(NON-PROPRIETARY)

ControfledANP-3114(NP)Revision 0CR-3 EPU -Feedwater Line Break Analysis Sensitivity StudiesMay 2012AREVA Controlled DocumentANP-3114(NP)Revision 0May 2012CR-3 EPU Feedwater Line Break Analysis Sensitivity StudiesCopyright © 2012AREVA Inc.All Rights Reserved Controled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 3Record of RevisionRevision Pages/Sections/ParagraphsNo. Changed Brief Description I Change Authorization000 All Initial Release Controlled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 4Table of Contents1 .0 In tro d u c tio n ........................................................................................................................................ 82.0 Analytical M ethodology ............................................................................................................ 93.0 Analysis Inputs ................................................................................................................................ 104 .0 R e s u lts ............................................................................................................................................ 1 34.1 Base Case for Sensitivity Studies .......................................................................................... 134.2 RCS Pressure Sensitivity ...................................................................................................... 144.3 RCS Flow Sensitivity ................................................................................................................... 154.4 Power Level Sensitivity ........................................................................................................ 164.5 Tim e in Cycle Sensitivity ............................................................................................................. 184.6 Steam Generator Inventory (Tave, SG Pressure, and SG Level) Sensitivity ............................. 204.7 Pressurizer Level Sensitivity ....................................................................................................... 244.8 Loss of Offsite Power ........................................... I .................................................................. 254.9 Break Size Sensitivity ............................................................................................................ 264.10 Break Location Sensitivity Study ................................................................................................. 284.11 Evaluation of Continued M ain Feedwater to Unaffected SG ...................................................... 304.12 Lim iting Feedw ater Line Break Conditions and Results ............................................................. 315.0 References ...................................................................................................................................... 38 Controlled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 5List of TablesTable 4-1: RCS Pressure Sensitivity Results ......................................................................................... 14Table 4-2: RCS Flow Sensitivity Results ............................................................................................ 15Table 4-3: Steady-State Results for Power Level Sensitivity ............................................................... 17Table 4-4: Power Level Sensitivity Results ........................................................................................... 17Table 4-5: Time in Cycle Sensitivity Results ........................................................................................ 19Table 4-6: SG Inventory Sensitivity Results ......................................................................................... 21Table 4-7: Pressurizer Level Sensitivity Results ................................................................................. 24Table 4-8: LOOP Evaluation Results .................................................................................................... 25Table 4-9: Break Size Sensitivity Results ............................................................................................. 27Table 4-10: Break Location Sensitivity Study ............................................................................................ 29Table 4-12: Effect of Continued MFW to Unaffected SG ........................................................................... 30Table 4-13: Sequence of Events for Limiting FWLB Transient ............................................................ 32Table 4-14: Results for Limiting FWLB Transient ................................................................................. 32 Controfled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 6List of FiguresFigure 4-1: Time to Reactor Trip versus Initial Steam Generator Inventory .............................................. 22Figure 4-2: Peak RCS Pressure versus Initial Steam Generator Inventory ............................................... 23Figure 4-3: Feedwater Line Break Locations ............................................................................................. 29Figure 4-4: FWLB RCS Peak Pressure versus Time ................................................................................. 33Figure 4-5: FWLB RCS Temperatures versus Time .................................................................................. 34Figure 4-6: FWLB Pressurizer Level versus Time ..................................................................................... 35Figure 4-7: FWLB SG Pressures versus Time .................................................................................. : ........ 36Figure 4-8: FWLB SG Levels versus Time .......................................................................................... 37 Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 7NomenclatureAcronym Definitionp3eff Delayed Neutron FractionBOC Beginning of CycleCR-3 Crystal River Unit 3DTC Doppler Temperature CoefficientEFIC Emergency Feedwater Initiation and ControlEFW Emergency FeedwaterEOC End of CycleEPU Extended Power UprateFOGG Feed Only Good GeneratorFWLB Feedwater Line BreakHZP Hot Zero PowerICS Integrated Control SystemLAR Licensing Amendment RequestLOCA Loss of Coolant AccidentLOFW Loss of FeedwaterLOOP Loss of Offsite PowerMFW Main FeedwaterMSSV Main Steam Safety ValvesMTC Moderator Temperature CoefficientNRC Nuclear Regulatory CommissionPORV Pilot Operated Relief ValvePSV Pressurizer Safety ValveRCP Reactor Coolant PumpRCS Reactor Coolant SystemRPS Reactor Protection SystemSDM Shutdown MarginSG Steam GeneratorSRP Standard Review PlanTave Average RCS TemperatureTSV Turbine Stop Valves ControDled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 81.0 INTRODUCTIONSection 2.8.5.2.4 of the Crystal River Unit 3 (CR-3) extended power uprate (EPU) licensing amendmentrequest (LAR) describes the feedwater line break (FWLB) evaluations that were performed to support thepower uprate. The evaluations that were performed are consistent with approved methodology [2] andthe current licensing basis of the plant, except that the licensing basis is being revised for EPU to use theStandard Review Plan (SRP) Section 15.2.8 reactor coolant system (RCS) pressure limit of 120% of thedesign pressure, or 3000 psig. Since the CR-3 licensing basis is being revised to use the SRP pressurelimit, additional sensitivity studies are performed to address requirements of SRP 15.2.8. The sensitivitystudies documented in this report are for the following parameters:* Initial RCS Pressure" Initial RCS Flow* Initial Power Level* Time in Cycle* Initial RCS Average Temperature (Tave)" Initial steam generator (SG) Pressure* Initial SG Level* Initial Pressurizer Level* Loss of offsite power (LOOP)* Break Size* Break Location Controlled DocumentAARE VAANP-31 14(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 92.0 ANALYTICAL METHODOLOGYThe thermal-hydraulic analysis of the FWLB at the CR-3 EPU power level is performed using theRELAP5/MOD2-B&W computer program (Reference [1]). The code simulates RCS and secondarysystem operation. The reactor core model is based on a point kinetics solution with reactivity feedbackfor control rod assembly insertion, fuel temperature changes, moderator temperature changes, andchanges in boron concentration. The RCS model provides for heat transfer from the core, transport of thecoolant to the steam generators, and heat transfer to the steam generators. The secondary modelincludes a detailed depiction of the main steam system, including steam relief to the atmosphere throughthe main steam safety valves (MSSVs) and simulation of the turbine stop valves (TSVs). The secondarymodel also includes the delivery of feedwater, both main and emergency, to the steam generators.The RELAP5/MOD2-B&W code has been approved by the Nuclear Regulatory Commission (NRC) foruse in non Loss of Coolant (LOCA) safety analyses (Reference [2]). The FWLB analysis documented inSection 2.8.5.2.4 of the CR-3 EPU LAR is consistent with Reference [2]. The methodology in Reference[2] dictates the input boundary conditions for many plant parameters, such as the RCS flow, Tave, andthe RCS initial pressure. The intent of the evaluations in this report is to determine the effect of changinginitial conditions on the FWLB transient to support compliance with SRP Section 15.2.8. Therefore, theanalyses in this document include input conditions that are not consistent with Reference [2]. However,the conditions specified in Reference [2] are included or bounded by the range of conditions considered inthe sensitivity studies.

Controfled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 103.0 ANALYSIS INPUTSMany of the input parameters in the FWLB sensitivity studies are consistent with the FWLB analysiswithout pressurizer spray performed to support Section 2.8.5.2.4 of the CR-3 EPU LAR. The inputparameters that are consistent are listed below:* The maximum power level considered is the targeted EPU power level of 3026.1 MWt (100.4% of3014 MWt). A conservative reactor coolant pump (RCP) heat of 16.4 MWt is included when theRCPs are operating.* The nominal average RCS temperature (Tave) is 582 *F, consistent with the increase in Tave thatis planned in conjunction with the EPU.* The nominal hot leg pressure is 2170 psia.* The minimum RCS flow rate considered is 374,880 gpm.* The nominal indicated pressurizer level plus uncertainty is 240 inches.* Two pressurizer safety valves (PSVs) were modeled with a nominal lift setpoint of 2500 psig, plus3% lift tolerance, and 0% accumulation. A blowdown of 4% was also considered.* Pressurizer heaters and pressurizer spray were not modeled.* The Pilot Operated Relief Valve (PORV) was not modeled for the FWLB analysis." Reactor trip was modeled to occur based on the reactor protection system (RPS) high RCSpressure trip function. The high RCS pressure trip setpoint includes the effects of elevatedpressure that may exist inside containment post-FWLB.* The control rod worth modeled in the analysis is the minimum rod worth that will meet theshutdown margin (SDM) requirement.* After reactor trip, the core heat generation rate was conservatively based on 1.0 times the ANS1971 decay heat standard for fission plus heavy actinides." Beginning-of-cycle (BOC) typical reactivity coefficients (Doppler temperature coefficient (DTC) -1.30 x 10-5 Ak/k/OF and moderator temperature coefficient (MTC) = 0.0 x 10-4 Ak/k/°F) wereevaluated.* The MSSVs were modeled to lift at the nominal setpoints plus 3% lift tolerance and 3%accumulation. A nominal blowdown of 5% was used." A single failure of one train of the Emergency Feedwater Initiation and Control (EFIC) system wasassumed as the worse case single failure. Therefore, only one of the two emergency feedwater(EFW) pumps was assumed to be available to provide flow to the SGs. The EFW temperaturewas modeled as 120 *F. The EFIC system contains Feed Only Good Generator (FOGG) logic,thus all EFW was provided to the unaffected SG.* Steam generator tube plugging of 5% was modeled.

Controlled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 11* No operator actions were credited." No Integrated Control System (ICS) actions were credited.Key input parameters modeled in the sensitivity studies that are different from the FWLB analyses inSection 2.8.5.2.4 of the CR-3 EPU LAR include:* The sensitivity studies considered power levels ranging from 60% to 100.4% of the EPU ratedthermal power level (3014 MWt). The sensitivity demonstrates that the peak RCS pressure ismuch less at lower power levels, so evaluations below 60% were not required.* The sensitivity studies considered a Tave range of 582 +/- 3 *F, allowing for an uncertainty of upto 3 'F.* The sensitivity studies considered hot leg pressure ranging from the minimum allowed value of2078.7 psia to a maximum value of 2244.7 psia, which is 25 psi above the setpoint for openingthe pressurizer spray valves to allow for uncertainty." The sensitivity studies considered RCS flow ranging from the minimum DNB flow of 374,880 gpmto a maximum flow of 398,850 gpm. The maximum flow value is based on no tube plugging, a2.5% uncertainty, and the conservative assumption that all of the SG tube walls are at theminimum tube wall thickness.* The sensitivity studies considered initial pressurizer level ranging from 240 inches to 290 inches,which is the maximum pressurizer level currently allowed for CR-3. A higher pressurizer level ismore limiting, so levels below 240 inches were not evaluated.* The sensitivity studies evaluated the effect of time in cycle on the FWLB transient including end ofcycle (EOC) typical reactivity coefficients (DTC = -2.00 x 10-5 Ak/k/0F and MTC = -4.0 x 10-4Aklk/°F).* The sensitivity studies evaluated initial SG levels ranging from 50 %OR to 95 %OR.* The sensitivity studies evaluated initial turbine header pressures in the range of 930 +/- 50 psia.* The sensitivity studies evaluated the FWLB transient with and without a LOOP. When modeled,the LOOP occurs at the time of turbine trip because a LOOP at the start of the event is lesslimiting.* The FWLB analyses in Section 2.8.5.2.4 of the CR-3 EPU LAR considered a high RCS pressurereactor trip setpoint of 2445.45 psia. This trip setpoint is based on the conservative assumptionthat the containment pressure is at the maximum allowed pressure (55 psig) by the time the tripsetpoint is reached following a FWLB. A more realistic evaluation of containment pressurefollowing a secondary side pipe break demonstrated that the containment pressure increases byless than 29.1 psi by the time the high RCS pressure trip setpoint is reached in a FWLB transient.Therefore, the sensitivity studies are based on a high RCS pressure setpoint of 2420 psia.* The FWLB analyses in Section 2.8.5.2.4 of the CR-3 EPU LAR conservatively modeled a1.0 %Ak/k SDM requirement at hot zero power (HZP). The Modes 1 and 2 minimum SDMrequirement is being increased to 1.3 %Aklk for the CR-3 EPU. Therefore, the sensitivity studiesare based on the EPU SDM requirement of 1.3 %Ak/k at HZP.

Controfled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 12The FWLB analyses in Section 2.8.5.2.4 of the CR-3 EPU LAR conservatively modeled the EFWflow as 550 gpm (total) occurring with a 60-second delay after the low SG level initiation setpoint.The CR-3 EPU loss of feedwater (LOFW) analysis in Section 2.8.5.2.3 of the CR-3 EPU LARrequires a minimum EFW flow of 660 gpm and a 40-second delay. Therefore, the sensitivitystudies are based on a minimum EFW flow of 660 gpm and a 40-second delay after the low SGlevel initiation setpoint for consistency with the CR-3 EPU LOFW analyses.

Controlled DocumentAARE VAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 134.0 RESULTS4.1 Base Case for Sensitivity StudiesAs identified in Section 3.0 of this report, several of the key input assumptions in the FWLB analysis inSection 2.8.5.2.4 of the CR-3 EPU LAR contained excess conservatism. These assumptions, as well assome details in the model, were modified prior to performing the sensitivity studies. In particular, thefollowing input assumptions were changed:1. The minimum EFW flow of 550 gpm with a 60-second delay time after the low SG level initiationsetpoint is reached was changed to a minimum EFW flow of 660 gpm with a 40-second minimumdelay time. This change makes the FWLB analysis consistent with the requirements of theLOFW transient. Note that the minimum EFW flow is based on the limiting single failureassumption that only one of the two EFW pumps is available to provide flow to the SGs.2. The SDM requirement was increased from 1.0 %Ak/k to 1.3 %Ak/k at HZP with the maximumworth control rod removed from the core. The Modes 1 and 2 minimum shutdown marginrequirement is being increased to 1.3 %Ak/k for the EPU.3. The RPS high RCS pressure trip setpoint was reduced from 2445.45 psia to 2420 psia.Evaluations determined that a setpoint of 2420 psia continues to conservatively account forelevated pressure that may exist inside containment post-FWLB at the time that the reactor tripsetpoint is reached.4. More detailed calculations of the form loss in the pressurizer surge line as a function of surge lineflow rate were incorporated into the FWLB model.5. A more detailed calculation of the reverse form loss through the MFW nozzles was performed.The results of the calculation were incorporated into the FWLB model.6. A preliminary sensitivity study of initial SG inventory determined that the peak RCS pressure ishighest for an initial SG level of 80 %OR. Therefore, the initial SG level was changed to 80 %OR.The peak RCS pressure for the FWLB analysis in Section 2.8.5.2.4 of the CR-3 EPU LAR is 2896.20psia. The FWLB transient described in Section 2.8.5.2.4 of the CR-3 EPU LAR was re-evaluated with thechanges described above. The resulting peak RCS pressure was 2829.42 psia. Unless otherwisestated, the model resulting in a peak RCS pressure of 2829.42 psia is used as the base case forsensitivity studies described in this report. The base case targets the following initial conditions:* 2170 psia hot leg pressure* Minimum RCS flow of 374,880 gpm* 100.4% of 3014 MWt0 Beginning of cycle typical parameters (MTC = 0.0 x 10-4 Ak/k/OF, DTC = -1.30 x 10-5 Ak/k/OF,Peff = 0.0070, Fuel Temperature -1400 OF)* Taveof582 °F* Pressurizer indicated level of 240 inches* SG level of 80 %ORa No LOOP* Double-ended guillotine break in the main feedwater line shortly before the line splits into twosmaller branches that feed the SG (see Figure 4-3).

Controfled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 144.2 RCS Pressure SensitivityThis section presents the results of the RCS pressure sensitivity study. The RCS pressure sensitivitystarts from the base case (nominal RCS pressure) and changes the initial hot leg pressure. The peakRCS pressure and reactor trip times from the RCS pressure sensitivity study are presented in Table 4-1.The RCS pressure sensitivity study demonstrates that the time of reactor trip during a FWLB transient isdirectly affected by the initial RCS pressure. For cases evaluated near the nominal hot leg pressure (i.e.2170 psia +/- 30 psia), the change in the time of reactor trip does not significantly affect the increase inRCS pressure that occurs after reactor trip. Consequently, the peak RCS pressure for each caseanalyzed near the nominal hot leg pressure is within 5 psi of the other cases.In cases that are initialized further away from the nominal hot leg pressure (i.e. 2078.7 psia and 2244.7psia), the change in the time of reactor trip is more significant. The transient progression during the timeleading up to reactor trip for these cases leads to a lower peak RCS pressure than the case evaluated atthe nominal hot leg pressure.Based on the RCS pressure sensitivity study, setting the initial RCS pressure to the nominal value isappropriate for evaluating the FWLB transient.Table 4-1: RCS Pressure Sensitivity Results Controlled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 154.3 RCS Flow SensitivityThis section presents the results of the RCS flow sensitivity study. The RCS flow sensitivity starts fromthe base case (minimum RCS flow) and changes the initial RCS flow to the maximum value. Establishingsteady-state conditions at higher RCS flow will result in a change in Tave, the SG pressure, the SGinventory, or a combination of these parameters. The RCS flow sensitivity maintains a consistent SGinventory for all cases and evaluates one case where only Tave changes and one case where only theSG pressure changes. The peak RCS pressure and reactor trip times from the RCS flow sensitivity studyare presented in Table 4-2.The RCS flow sensitivity study shows that the RCS flow does not significantly change the hot legpressure during the transient, and therefore does not affect the time of reactor trip. However, since thehot leg pressure is the same and the RCS flow is higher, the AP from the hot leg to the location of thepeak RCS pressure (i.e. at the bottom of the reactor vessel) increases. Consequently, the peak RCSpressure is [ ] higher for cases evaluated at the maximum RCS flow.Based on the RCS flow sensitivity study, the peak RCS pressure for a FWLB transient should beevaluated assuming maximum RCS flow. This would not be conservative for evaluating departure fromnucleate boiling (DNB). However, as explained in Section 2.8.5.2.4.2 of the CR-3 EPU LAR, since thereactor coolant pumps (RCPs) remain operating, the RCS fluid remains subcooled, and the core powerremains less than 112% throughout the FWLB analysis, it is concluded that the minimum DNB ratio wouldremain above the applicable correlation limit without performing an explicit analysis for FWLB. Therefore,the RCS flow for the FWLB transient should be biased to challenge the peak RCS pressure acceptancecriteria.Table 4-2: RCS Flow Sensitivity Results Controled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 164.4 Power Level SensitivityThis section presents the results of the power level sensitivity study. The base case is evaluated at100.4% of the proposed EPU power level. The power level sensitivity study compares the base case toevaluations performed at 95.4 %FP, 80 %FP, and 60 %FP. Table 4-3 documents the steady-stateconditions achieved for each power level. The peak RCS pressure and reactor trip times from the FWLBpower level sensitivity study are presented in Table 4-4.The FWLB results show that as the power level decreases, the lower initial SG inventory results in anearlier reactor trip. The earlier reactor trip in conjunction with a lower initial core power level causes thepeak RCS pressure to be significantly lower as the power level decreases. The peak RCS pressure isless than 110% of the RCS design pressure (< 2764.7 psia) for initial power levels of 80 %FP and below.Given the significant reduction in peak RCS pressure observed at 80 and 60 %FP, it was concluded thatpower levels <60 %FP did not need to be evaluated.CR-3 is allowed to operate with a positive MTC at lower power levels. As discussed in Section 2.8.5.4.1.2of the CR-3 EPU LAR, the most positive MTC that will be allowed for EPU core designs is +7.5 x 10-5Ak/k/°F. The FWLB power level sensitivity study included evaluations at 95.4 %FP, 80 %FP, and 60 %FPwith an MTC of +7.5 x 10`5 Ak/k/°F. A positive MTC resulted in a higher peak RCS pressure for eachpower level analyzed; however, the base case (100.4 %FP, 0.0 x 10-5 Aklk/°F) remained the limiting case.Based on the power level sensitivity study, the FWLB transient should be evaluated at 100.4 %FP.

Controfled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivitv StudiesANP-3114(NP)Revision 0Page 17Table 4-3: Steady-State Results for Power Level SensitivityTable 4-4: Power Level Sensitivity Results Controfled DocumentAARN EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 184.5 Time in Cycle SensitivityThis section presents the results of the time in cycle sensitivity study. The base case is evaluated atconservative BOC conditions. The MTC (a constant 0.0 x 10-5 Ak/k/OF) is the most positive MTC allowedat full power. The DTC is a value (-1.3 x 10-5 Ak/k/°F) that is more positive than expected for EPU fuelcycle designs (see Table 2.8.2-2 of the CR-3 EPU LAR). The fuel temperature (-1400 OF) isrepresentative of BOC conditions at the EPU power level. Finally, the effective delayed neutron fraction(Peff) is a BOC typical value (0.0070). The rod worth used in the base case is determined to be theminimum required rod worth needed to meet the SDM definition for the modeled MTC, DTC, and fueltemperature. The time in cycle sensitivity study evaluates the effects of changing MTC, DTC, fueltemperature, and P3eff on the FWLB transient. Table 4-5 presents the key results from the FWLB time incycle sensitivity study.The base case utilizes an MTC that does not change with moderator temperature or density. Therefore,the time in cycle sensitivity analysis compared the base case to cases using constant MTC values of-10.0 x 10-5 Ak/k/°F and -20.0 x 10-5 Ak/k/°F. When a constant MTC is used, the thermal power level ofthe core does not change significantly prior to reactor trip. Consequently, the time of reactor trip isapproximately the same for all three cases evaluated with a constant MTC. Although the pre-triptransient behavior is approximately the same in all three cases, the results show that a more negativeMTC resulted in a lower peak RCS pressure. This occurs because a more negative MTC requires morerod worth in order for the SDM requirement to be met. The additional rod worth results in a fasterreduction in the core thermal power shortly after trip, which in turn reduces the peak RCS pressure. Thetrend versus MTC is clear, so evaluations were not performed for constant MTCs that are more negativethan -20.0 x 10-5 Ak/k/°F.The time in cycle sensitivity study also includes cases that allow the moderator reactivity feedback tochange with moderator density. The sensitivity study considered five reactivity versus moderator densitycurves. The MTCs from the curves at initial hot full power conditions are 0.0 x 10.5 Ak/k/0F, -10.0 x 10-5Ak/k/0F, -20.0 x 10-5 Ak/k/0F, -30.0 x 10-5 Ak/k/°F, and -40.0 x 10.5 Ak/k/°F. Table 2.8.2-2 of the CR-3EPU LAR lists the expected EPU most negative MTC limit as -37.5 x 10-5 Ak/k/°F, so a -40 x 10.5 Ak/k/°FMTC bounds typical EPU cycles. The sensitivity study performed using different reactivity versusmoderator density curves demonstrates that as the MTC becomes more negative, the peak thermalpower increases and the time to reactor trip decreases. However, in all of the cases the peak thermalpower remained well below 112 %FP. The increase in thermal power that occurs prior to reactor trip ismore than offset by the increased rod worth that is needed to meet the SDM definition at more negativeMTCs. Therefore, as the MTC becomes more negative the peak RCS pressure decreases.The time in cycle sensitivity study included a case to evaluate the effect of DTC. The case is evaluatedwith an MTC that is representative of EOC (i.e. -40 x 10-5 Ak/k/°F). The DTC sensitivity study considersvalues of -1.30 x 10.5 Ak/k/°F and -2.00 x 10.' Ak/k/IF, which bound the typical EPU values in Table2.8.2.2 of the CR-3 EPU LAR. The DTC comparison shows that a more negative DTC causes themaximum thermal power to decrease because of the additional negative reactivity feedback.Furthermore, the more negative DTC requires more rod worth to meet the minimum SDM requirement.These two factors result in a lower peak RCS pressure for the case evaluated with a more negative DTC.The effect of fuel temperature is evaluated with reactivity coefficients that are representative of EOC. Thefuel temperature sensitivity reduces the BOC typical fuel temperature (-1400 OF) to a fuel temperaturethat is representative of EOC conditions (-1080 OF). The fuel temperature is changed by increasing thefuel gap thermal conductance. The fuel temperature comparison demonstrates that the increased fuelgap thermal conductance associated with EOC fuel temperatures results in a slightly higher peak thermal Controlled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 19power. In addition, a lower initial fuel temperature means that less rod worth is needed to meet theminimum SDM requirement because the initial fuel temperature is closer to the HZP temperature.However, the lower initial fuel temperature also results in less heat stored in the fuel that must beremoved following reactor trip. Consequently, the peak RCS pressure is significantly less for casesevaluated at EOC typical fuel temperatures.The effect of P3eff is evaluated with an MTC, DTC, and fuel temperature that are representative of EOC.The P3eff sensitivity study considers values of 0.0045 and 0.0070, which bound the typical EPU values inTable 2.8.2.2 of the CR-3 EPU LAR. The P3eff comparison shows that smaller values of 3eff result in afaster positive reactivity response prior to trip, which increases the maximum thermal power. However,smaller values of 13eff also result in a faster negative reactivity response as the control rods are inserted.The effect of P3eff on the post-trip portion of the transient has more effect on the peak RCS pressure,,which occurs a few seconds after reactor trip. Consequently, the peak RCS pressure is significantly lessfor cases evaluated at EOC typical values of P3eff.The sensitivity studies on time in cycle demonstrated that the peak RCS pressure in a FWLB transient ishigher when BOC typical values are modeled for MTC, DTC, fuel temperature, and Peff.Table 4-5: Time in Cycle Sensitivity Results Controled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 204.6 Steam Generator Inventory (Tave, SG Pressure, and SG Level) SensitivityThis section presents the results of the SG inventory sensitivity study, which varies RCS Tave, SGpressure, and SG level to achieve different initial SG inventories. The SG inventory sensitivity studystarts from the base case. The RCS flow is set to a maximum value because the evaluationssummarized in Section 4.3 determined that a maximum RCS flow results in a higher peak RCS pressurethan a case at a minimum DNB flow. The initial RCS pressure, initial power level, and time in cycleevaluated in the SG inventory sensitivity study match the base case because the values used in the basecase were shown to be the limiting conditions. The RCS Tave, SG pressure, and SG level are variedwithin the following spectrum of initial conditions:" RCS Tave in the range of 582 'F +/- 3 *F" Turbine header pressure in the range of 930 psia +/- 50 psi" SG level ranging from 50 %OR to 95 %ORAlthough the above ranges are considered, every possible combination cannot realistically be achieved.For a given Tave, the minimum level that can be achieved in the SG is affected by the orifice plate in theSG downcomer and the initial SG pressure. Therefore, the minimum SG levels considered for a givenRCS Tave and SG pressure combination is either 50 %OR or the minimum level when the orifice plate isfully opened. Similarly, the maximum SG pressure that can be modeled for a given RCS Tave is thepressure that results in a SG level of 95 %OR when the orifice plate is fully open.Table 4-6 summarizes the steady-state SG inventory and the key FWLB transient results for each of the34 cases that were evaluated to consider the effects of RCS Tave, SG pressure, and SG level on theFWLB transient. The results clearly show that as the SG inventory increases, the time to reactor tripincreases. This is demonstrated by Figure 4-1. The longer time to reactor trip means that cases startingwith more initial SG inventory also discharge more inventory out of the break by the time reactor trip isreached. This behavior leads to relatively small differences in the affected SG inventory when reactor tripoccurs. Since each case has approximately the same inventory in the affected SG at the time of reactortrip, the RCS pressure increase shortly after reactor trip is close to the same for each case considered.Furthermore, since the reactor trip is initiated by a high RCS pressure trip function, each case is at thesame hot leg pressure when the reactor trip occurs. Consequently, the peak RCS pressure in a FWLBtransient is not sensitive to RCS Tave, SG pressure, SG level, or SG inventory. This is demonstrated bythe similarity in the peak RCS pressure results in Table 4-6 and Figure 4-2. The highest peak RCSpressure for any of the cases considered is 2842.15 psia at an RCS Tave of 585 'F, a turbine headerpressure of 940 psia, and a SG level of 80 %OR.Although the peak RCS pressure is not sensitive to changes in RCS Tave, SG pressure, SG level, or SGinventory, these parameters will affect the mass and energy released to containment. Cases with higherinitial SG inventory would release more mass from the SG side of the break and would delay reactor tripallowing more mass from the MFW side of the break to reach containment. Conversely, minimizing theinitial SG inventory will minimize the mass and energy released to containment.

Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 21Table 4-6: SG Inventory Sensitivity Results Controled DocumentAAR EVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 22Figure 4-1: Time to Reactor Trip versus Initial Steam Generator Inventory Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 23-Figure 4-2: Peak RCS Pressure versus Initial Steam Generator Inventory Controled DocumentAAR EVAANP-31 14(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 244.7 Pressurizer Level SensitivityThis section presents the results of the pressurizer level sensitivity study. The evaluations summarized inSections 4.1 through 4.6 are based on an indicated pressurizer level of 240 inches, which is the nominalpressurizer level of 220 inches plus 20 inches for uncertainty. The pressurizer level sensitivity studyevaluates pressurizer levels as high as 290 inches, which is the maximum pressurizer level currentlyallowed for CR-3.Six of the cases evaluated in Section 4.6 were selected for evaluation in the pressurizer level sensitivitystudy. The cases chosen were selected to span the ranges of conditions evaluated in Section 4.6. Thecases chosen also include the cases in Table 4-6 that had the highest peak RCS pressure. The peakRCS pressure and reactor trip times from the pressurizer level sensitivity study are presented inTable 4-7.The pressurizer level sensitivity study demonstrated that for each case, a higher initial pressurizer levelcauses the RCS pressure to respond more quickly to changes in the RCS temperatures because of thesmaller available steam space. Consequently, the reactor trip occurs sooner and the peak RCS pressureis higher when the initial pressurizer level is higher. The highest peak RCS pressure for any of the casesconsidered with a pressurizer level of 290 inches is 2861.20 psia.In addition to evaluating the peak RCS pressure, the pressurizer level sensitivity study also evaluated thetemperature of liquid in the pressurizer for those cases that predicted liquid relief out of the PSVs. Thepressurizer sensitivity study verified that if liquid is passed by the PSVs, the liquid remains above 600 OF.The sensitivity study on pressurizer level demonstrated that a higher initial pressurizer level is morelimiting for the FWLB transient.Table 4-7: Pressurizer Level Sensitivity Results ControDled DocumentAAREVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 254.8 Loss of Offsite PowerThis section presents the results of the LOOP sensitivity study. The evaluations in Sections 4.1 through4.7 are based on offsite power remaining available throughout the FWLB transient. The LOOP evaluationdetermines if the FWLB transient results are more limiting if offsite power is lost during the transient.If the LOOP occurs at the time that the break opens, the loss of offsite power will result in a reactorscram, which would reduce the energy added to the RCS by the core. Therefore, a LOOP at eventinitiation is less limiting than if a LOOP does not occur. On the other hand, if the LOOP is modeled tooccur at the time of turbine trip, which is modeled to occur coincident with reactor trip, the LOOP will notaffect the energy added to the RCS by the core. Furthermore, a LOOP at turbine trip causes the RCSflow to decrease just as the RCS pressure is approaching the maximum value. Therefore, a LOOP atturbine trip is evaluated as the limiting time for a LOOP to occur.The LOOP evaluation considers the six cases from Section 4.7, but with a LOOP modeled at turbine trip.The results with and without LOOP for each of the six cases are summarized in Table 4-8.The LOOP evaluation demonstrates that in each case, the peak RCS pressure is higher when a LOOPoccurs at turbine trip. The highest peak RCS pressure (2878.38 psia) continues to occur at the sameconditions as the case with no LOOP (Tave of 585 OF, a turbine header pressure of 940 psia, and a SGlevel of 80 %OR). The LOOP evaluation also confirmed that liquid passed by the PSVs remains above600 OF.Based on the LOOP evaluation, the FWLB transient should be evaluated with a LOOP occurring at thetime of turbine trip.Table 4-8: LOOP Evaluation Results Controlled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 264.9 Break Size SensitivityThis section presents the results of the break size sensitivity study. The evaluations in Sections 4.1through 4.8 are based on a double-ended guillotine break of the largest main feedwater (MFW) pipe nearthe junction where the largest pipe splits into two smaller pipes that continue to the SG (see Figure 4-3).The break size sensitivity study evaluates the effect of smaller break sizes at the same location. Thebreak size sensitivity study did not change the area of the MFW side of the break. Therefore, all MFW islost to the affected SG. Only the area on the SG side of the break was reduced to determine the breaksize sensitivity.The break size sensitivity study was performed at the same conditions as the limiting case identified inSection 4.8, namely:0 Nominal RCS Pressure* Maximum RCS Flowa 100.4 %FP* BOCa 585 'F RCSTave* 940 psia Turbine Header Pressure& 80%ORSG Level* 290 Inch Indicated Pressurizer Level* LOOP occurring at Turbine TripThe model was updated to include a detailed MFW system model from the booster pumps to the SGs.The MFW system model was set to isolate MFW to the unaffected steam generator at the start of thetransient for consistency with previous FWLB evaluations. Overall, the model with a simplified MFWsystem and the model with a detailed MFW system produced equivalent FWLB transient results for thesame break size. Incorporating the detailed MFW system resulted in a slightly different peak RCSpressure (2875.31 psia versus 2878.38 psia).The CR-3 MFW system has three pipe sizes between the isolation check valve and the SGs. The largestmain feedwater pipe is 18" SCH 80 (1.418 ft2). The largest main feedwater pipe branches into two 14"SCH 80 (0.8522 ft2) pipes. The two 14" SCH 80 pipes then feed into a total of 32 riser pipes, each ofwhich are 3" SCH 80 (0.04587 ft?). These three pipe areas are considered as the area of the SG side ofthe break along with arbitrarily selected pipe areas to fully capture the behavior of peak RCS pressureversus break size. The results of the break size sensitivity study are shown in Table 4-9.The break size sensitivity study demonstrates that the peak RCS pressure is effectively the same forbreak sizes ranging from 0.8522 ft2 (the area of the 14" SCH 80 pipes in the MFW system) to 1.418 ft2(the area of the 18" SCH 80 pipe in the MFW system). For break sizes less than 0.8522 ft2, the peakRCS pressure steadily decreases. Note that the highest peak RCS pressure was calculated for a 1.100ft2 break, but the difference in peak RCS pressure between a 1.418 ft2 break and a 1.100 ft2 break is only2.6 psi. Furthermore, the CR-3 MFW system does not actually have a pipe with a 1.100 ft2 area betweenthe isolation check valves and the SG. Therefore, evaluations based on the maximum pipe area of1.418 ft2 represent the limiting break area.Based on the break size sensitivity study, the limiting break area for evaluating peak RCS pressure is thefull area of the largest pipe in the CR-3 MFW system between the isolation check valves and the SGs(1.418 ft2).

Controled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivitv StudiesANP-3114(NP)Revision 0Page 27Table 4-9: Break Size Sensitivity Results Controlled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 284.10 Break Location Sensitivity StudyThis section presents the results of the break location sensitivity study. This break location sensitivitystudy evaluates the effects of moving the double ended guillotine break to different pipe locations. Thefollowing locations were considered (see Figure 4-3):" The limiting break from Section 4.9, which is a break in the main feedwater pipe in SG B. Thebreak is located shortly before the tee connecting the 18" main feedwater line to the two 14" sidebranches." A break in the same location as the limiting case from Section 4.9, but occurring on SG A insteadof SG B.* A break in the same pipe as the limiting case from Section 4.9, but occurring immediately afterthe isolation check valve.* A break in one of the side branches of SG B at the tee that connects the main feedwater line tothe side branches. The break is modeled as a double-ended guillotine break of a pipe with a1.418 ft2 area.* A break in one of the side branches of SG B near the entrance to the SGs. The break is modeledas a double-ended guillotine break of a pipe with a 0.8522 ft2 area.The peak RCS pressure and reactor trip times for each of the identified break locations are summarizedin Table 4-10.The results in Table 4-10 show that a break in SG B is clearly more limiting than a break in SG A.The peak RCS pressure for the four break locations in SG B all result in effectively the same peak RCSpressure, but different times for reactor trip. The reactor trip time indicates how quickly the affected SGdischarges inventory out of the break. The earliest reactor trip time occurs when the break is in thebranch side of the tee connecting the main feedwater line to the side braches. This location allows theSG inventory to reach both sides of the 1.418 ft2 double-ended guillotine break. Since the total effectivebreak area is larger, the reduction in inventory occurs more quickly. However, as mentioned above, thepeak RCS pressure is effectively the same as the peak RCS pressure when the break is located in themain feedwater line before the tee. Furthermore, for both break locations all of the affected SG inventorywill be discharged to containment by 60 seconds and the RCS conditions at that time are approximatelythe same. Therefore, the FWLB transient results are effectively the same for both break locations.A break in one of the side branches near the SG also allows the SG inventory to reach both sides of thedouble-ended guillotine break, but the break area in both directions is only 0.8522 ft2.The net effect isthat the reactor trip time is slightly (0.188 sec) longer than the case with a break in the main feed linebefore the tee. The calculated peak RCS pressure in both cases differs by only 0.10 psi, which is aninsignificant difference.A break in the main feed line just downstream of the isolation check valves requires the longest time toreach the reactor trip setpoint. The peak RCS pressure is slightly less than the case with the break in themain feedwater line right before the tee, but the difference is negligible (2.12 psi).The break location sensitivity study confirms that the limiting break location for evaluating peak RCSpressure is a double-ended break in the main feedwater pipe in SG B, located shortly before the teeconnecting the 18" main feedwater line to the two 14" side branches.

Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 29Figure 4-3: Feedwater Line Break LocationsIsolation Check Valve18"14"14"SGTable 4-10: Break Location Sensitivity Study Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 30-4.11 Evaluation of Continued Main Feedwater to Unaffected SGThe CR-3 EPU FWLB evaluations conservatively model all feedwater to the unaffected SG as being lostat the start of the event. This conservative approach simplifies the analysis because the effect of controlsystems on the main feedwater pump speed during the transient can be ignored. This section presentsan evaluation that demonstrates the conservatism added to the analysis by not crediting feedwater.The evaluations with continued feedwater assume that the main feedwater pump speed is maintained atthe initial condition value. EFIC actuation isolates feedwater or main feedwater is runback following areactor trip. Once main feedwater is isolated or runback is.started, the analysis conservatively assumesthat all main feedwater is stopped to the unaffected SG in 3.2 seconds for consistency with the LOFWanalysis (Section 2.8.5.2.3.2 of the CR-3 EPU LAR).The detailed feedwater model includes the cross connect piping that exists between the booster pumpsand the main feedwater pumps. Therefore, the evaluations with continued feedwater allow flow from thebooster pump in loop A to feed the main feedwater pump on the broken loop if the transient predicts thisbehavior. During normal operation, the CR-3 main feedwater system does not have a cross connectbetween the loops downstream of the main feedwater pumps.The FWLB evaluations with continued feedwater were performed for several break sizes. The peak RCSpressure and time of reactor trip are summarized in Table 4-11. Table 4-11 also contains the equivalentresults when no credit is taken for continued MFW to the unaffected SG.Table 4-12: Effect of Continued MFW to Unaffected SGThe results show that for each break size, the time to reactor trip is longer and the peak RCS pressure islower when MFW is modeled to reach the unaffected SG. For the limiting break size, the peak RCSpressure is 2875.31 -2828.24 = 47.07 psi lower when MFW is allowed to reach the unaffected SG. Theeffect is less at smaller break sizes, but the reduction in peak RCS pressure is still significant.The above evaluations demonstrate that significant conservatism is included in the FWLB evaluationsbecause of the assumption that all feedwater is lost to the unaffected SG at the start of the event. Themargins calculated above do not consider the effect of control systems changing the main feedwaterpump speed during the transient. Nevertheless, any flow reaching the unaffected SG would be a benefit.

Controfled DocumentAAR EVAANP-3114(NP)CR-3 EPU Feedwater Line Break Analysis Revision 0Sensitivity Studies Page 314.12 Limiting Feedwater Line Break Conditions and ResultsThe following targeted conditions were found to result in the highest peak RCS pressure during a FWLBtransient.* Nominal RCS pressure (2170 psia hot leg pressure)* Maximum RCS flow (398,850 gpm)a 100.4 %FP* BOC (0.0 x 10- MTC, -1.3 x 10" Ak/k/°F DTC, -1400 'F, 0.0070 Peff)* 585 'F RCSTave* 940 psia turbine header pressurea 80 %OR SG level* 290 inch indicated pressurizer level* LOOP occurring at turbine trip0 Double ended guillotine break located in the main feedwater line of SG B right before the teeconnecting to the two side branches.The sequence of events from this case is presented in Table 4-13. The results for the limiting case arepresented in Table 4-14. Figures 4-4 through 4-8 plot key transient results for the limiting case.

Controfled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 32Table 4-13: Sequence of Events for Limiting FWLB TransientEvent Time(sec)Transient initiated 0.0MFW to both SGs interrupted 1.OE-6Pressurizer spray begins N/APeak thermal power occurs 7.9RPS high RCS pressure trip actuated 10.0Control rods begin to insert 10.6Turbine stop valves (TSVs) begin to closeLOOPEFIC actuated on low SG-B level 11.5Initial PSV lift occurs -12.5Peak RCS pressure occurs 14.7EFW flow begins 51.5Affected SG depressurization complete -55.0Final PSV closure occurs -200.0Peak Tave occurs -390.0Transient terminated 600.0Table 4-14: Results for Limiting FWLB TransientParameter ValuePeak RCS Pressure (psia) 2878.38Peak thermal power (%RTP) 100.50Peak Tave (°F) 616.27 Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 33Figure 4-4: FWLB RCS Peak Pressure versus Time0U.P4.30002900280027002600250024002300220021002000 Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 34Figure 4-5: FWLB RCS Temperatures versus Time65064063062061060020 590.-o.580570560550540530100200 300400Time (s)500600 700 800 Controiled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivitv StudiesANP-3114(NP)Revision 0Page 35Figure 4-6: FWLB Pressurizer Level versus Time44424000.rE00)N3836343230282624400Time (s)

Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 36Figure 4-7: FWLB SG Pressures versus TimeI-00.400Time (s)

Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 37Figure 4-8: FWLB SG Levels versus TimeU-(-)Time (s)

Controlled DocumentAAREVACR-3 EPU Feedwater Line Break AnalysisSensitivity StudiesANP-3114(NP)Revision 0Page 385.0 REFERENCES1. BAW-10164PA-06, "RELAP5/MOD2-B&W- An Advanced Computer Program for Light WaterReactor LOCA and Non-LOCA Transient Analysis."2. BAW-1 01 93PA-00, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed PressurizerWater Reactors."