ML16057A812
| ML16057A812 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/25/2016 |
| From: | Sartain M D Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards |
| References | |
| 16-011A, CAC MF6251 | |
| Download: ML16057A812 (76) | |
Text
Dominion Nuclear Connecticut, Inc. 5000 Dominion Boulevard, Glen Alien, VA 23060 DominuIionWeb Address: www.dom.comFebruary 25, 2016U.S. Nuclear Regulatory Commission Serial No. 16-011lAAttention: Document Control Desk NLOS/WDC R0Washington, DC 20555 Docket No. 50-423License No. NPF-49DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDINGLICENSE AMENDMENT REQUEST TO ADOPT DOMINION CORE DESIGN AND SAFETYANALYSIS METHODS AND TO ADDRESS THE ISSUES IDENTIFIED IN WESTINGHOUSEDOCUMENTS NSAL-09-5, REV. 1, NSAL-15-1, AND 06-1C-03 (CAC NO. MF6251)By letter dated May 8, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted a licenseamendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposedamendment would revise the Technical Specifications (TS) to enable use of the Dominionnuclear safety analysis and reload core design methods for MPS3 and address the issuesidentified in three Westinghouse communication documents. In a letter dated January 8,2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additionalinformation (RAI) to DNC related to the LAR. The RAI contained 18 questions. In a letterdated January 28, 2016, DNC responded to RAI Questions RAI-1 through RAI-8, RAI-13, andRAI-18. Attachment I is the DNC response to RAI Questions RAI-9 through RAI-12 and RAI-14 through RAI-16. DNC plans to submit the response to the remaining RAI question, RAI-17,by March 31, 2016.During preparation of the RAI responses, Dominion identified several minor discrepancieswithin Attachment 5 of the May 8, 2015 LAR. In response to RAI-9, DNC has updated theanalyses discussed in the RETRAN benchmark information originally provided in Attachment5. The RETRAN benchmark information is not used in any analysis of record. Therefore, thediscrepancies do not impact the no significant hazards consideration determination provided inthe May 8, 2015 LAR. Attachment 2 provides an update to Attachment 5 of the May 8, 2015LAR which corrects the discrepancies and includes the revised results in response to RAI-9.If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.Sincerely,-' ::Vicki, L.:di Hull:.~NOTARY PUBILICMark D. Sartain I Commonwe~alth of virgin~iaVice President -Nuclear Engineering My om....o ...........3.l201COMMONWEALTH OF VIRGINIA)COUNTY OF HENRICO)The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark 0. Sartain,who is Vice President -Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is dulyauthorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true tothe best of his knowledge and belief.Acknowledged before me this 2016.My Commission Expires: S Notary Public
~Serial No. 16-011ADocket No. 50-423~Page 2 of 2Commitments made in this letter: NoneAttachments:1. Response to Request for Additional Information Regarding License AmendmentRequest to Adopt Dominion Core Design and Safety Analysis Methods and toAddress the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1,NSAL-15-1, and 06-1C-03 (CAC No. MF6251) -RAI Questions RAI-9 throughRAI-12 and RAI-14 through RAI-162. RETRAN Benchmarking Information -Updatedcc: U.S. Nuclear Regulatory CommissionRegion I2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713R. V. GuzmanSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North, Mail Stop 08-C211555 Rockville PikeRockville, MD 20852-2738NRC Senior Resident InspectorMillstone Power StationDirector, Radiation DivisionDepartment of Energy and Environmental Protection79 Elm StreetHartford, CT 06106-5127 Serial No. 16-O11ADocket No. 50-423ATTACHMENT IRESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDINGLICENSE AMENDMENT REQUEST TO ADOPT DOMINION CORE DESIGN ANDSAFETY ANALYSIS METHODS AND TO ADDRESS THE ISSUES IDENTIFIED INWESTINGHOUSE DOCUMENTS NSAL-09-5. REV. I. NSAL-15-I, AND 06-1C-03(CAC NO. MF6251)RAl QUESTIONS RAI-9 THROUGH RAl-12 AND HAl-14 THROUGH RAl-16DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 t.Serial No. 16-011IADocket No. 50-423Attachment 1, Page 1 of 11By letter dated May 8, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted alicense amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). Theproposed amendment would revise the Technical Specifications (TS) to enable useof the Dominion nuclear safety analysis and reload core design methods for MPS3and address the issues identified in three Westinghouse communication documents.In a letter dated January 8, 2015, the Nuclear Regulatory Commission (NRC)transmitted a request for additional information (RAI) to DNC related to the LAR. TheRAI contained 18 questions. In a letter dated January 28, 2016, DNC responded toRAI Questions RAI-1 through RAI-8, RAI-13, and RAI-18. This attachment providesDNC's response to RAI Questions RAI-9 through RAI-12 and RAI-14 through RAI-16.DNC plans to submit the response to the remaining RAI question, RAI-17, by March31, 2016.RAI -9 (SRXB): Reanalysis for the RETRAN Benchmarking CasesThe licensee indicated in an e-mail dated June 30, 2015 (ADAMS Accession No.ML 15349A808) that it identified a discrepancy between the MPS3 RETRAN BaseModel Pressurizer Shell Heat Conductor and the Dominion RETRAN topical report(TR). The MPS3 RETRAN base input deck models the pressurizer shell as a heatconductor, which differs from TR, VEP-FRD-41, which states that "Dominioncontinues to model the non-equilibrium wall as an adiabatic surface." Each of the fivebenchmarking cases supporting the LAR were reanalyzed with needed correction.Discuss the results of the reanalysis for the five benchmarking cases, and provide themodified graphs showing the changes and text for the affected cases, including theaffected loss of normal feedwater benchmark case.DNC ResponseThe results of the reanalysis of the five benchmarking cases originally submitted areincluded in an update to Attachment 5 of the May 8, 2015 LAR provided inAttachment 2 to this letter. The updated Attachment 5 is provided with the changesnoted by a change bar in the right hand margin of the affected pages. Most of theplots of key parameters are unchanged as there was no observable difference in theplotted values. Some selected plots for the Loss of Load event are revised becausethey exhibited differences in the period after the occurrence of peak reactor coolantsystem (RCS) pressure. The description of some items in input summary tables andresults tables have been clarified to conform with the details of the analyses. Thesame conclusions originally stated in Attachment 5 of the May 8, 2015 LAR continueto be supported by the updated benchmarking analyses.
Serial No. 16-011IADocket No. 50-423Attachment 1, Page 2 of 11RAI -10 (SRXB): Loss of Load (LOL) Benchmark Analysis -RCSPressurization RateOn page 9 of Attachment 5 to the LAR, the licensee indicated that "the Dominioncase trips slightly earlier than the FSAR [final safety analysis report] data because ofthe higher RCS [reactor coolant system] pressurization rate", and Table 4.1-2showed that for the L OL [loss of load/turbine trip] analysis, the calculated peak RCSpressure for the Dominion case is 2, 717.19 psia [pounds-per-square-inch, absolute],which is 12.22 psi lower than the peak pressure of 2, 729.41 psia for the FSAR case.Discuss the differences of the models, input parameters or assumptions used in theLOL analyses for the Dominion case and the FSAR case that will result in:(1) a higher RCS pressurization rate before the reactor trip for the Dominion caseand;(2) a lower peak RCS pressure for the Dominion case against the ESAR casediscussed above.DNC ResponseThe peak pressure value of 2,729.41 psia referenced in the RAI question above haschanged to 2,705 psia in the updated benchmarking analysis provided in Attachment2.The loss of load/turbine trip (LOL) transient causes a sudden reduction in steam flow,resulting in an increase in the steam generator (SG) secondary pressure and acorresponding increase in RCS temperature and pressure. During the early phase ofthe LOL benchmark analysis, the Dominion rate of RCS pressurization is similar tothe FSAR case. However, the Dominion case begins to pressurize slightly earlier asthe primary side fluid expands into the pressurizer. The slightly earlier pressurizationis attributed to differences in the SG primary-to-secondary heat transfer associatedwith the Dominion single-node steam generator (SNSG) secondary compared to theFSAR multi-node steam generator (MNSG) secondary. For the SNSG, thesecondary-side temperature corresponds to the saturation temperature for thesecondary-side pressure, and will therefore increase with any increase in pressure.The MNSG is subdivided into regions that may be either saturated or subcooled. TheMNSG tends to better maintain heat transfer during transient conditions due toincreased noding and modeling of dynamic effects for the liquid/vapor flow throughthe tube bundle. These effects are expected to be small for the rapid changesassociated with the LOL event, but do result in a slightly earlier heatup for the SNSGand associated increase in primary-side pressure. The subsequent rate ofpressurization is similar for both cases. However, for the Dominion case, becausethe pressure increase starts earlier, the reactor trip on high pressurizer pressureoccurs slightly earlier.
Serial No. 16-011ADocket No. 50-423Attachment 1, Page 3 of 11The peak RCS pressure, which occurs after the reactor trip, is closely related to theresponse of the pressurizer safety valves (PSV). It is noted here that the main steamsafety valves (MSSV) actuate after the time of peak RCS pressure and, therefore,are not a factor for peak RCS pressure. As shown in Figure 4.1-1 of updatedAttachment 5 (provided in Attachment 2 of this letter), the pressurizer pressure ismodulated over a narrow band by the PSVs for the FSAR case, achieving pressuresthat are slightly higher than the Dominion case which has a relatively flat pressureprofile when the PSVs open. Since the LOL event results in a very rapid pressureincrease exceeding 60 psi/sec, even small differences in PSV response (e.g., delays,opening profiles, etc.) can noticeably affect the peak pressure. These differences areeven more pronounced in the RCS cold leg and reactor vessel lower plenum wherepeak pressures exceed 2700 psia and are affected by differences in loop response(RCS loop, reactor vessel, and surge line loss coefficients, reactor coolant pumphead dynamics, etc.). Given the large and sudden increase in pressure from theinitial value of 2200 psia for the pressurizer, the differences in peak pressure arerelatively small.RAI -11 (SRXB): Locked Rotor (LR) Benchmark Analysis -Over-Pressurization and DNBOn page 14, the licensee stated for the LR analysis that, based on the datacomparison between the Dominion case and FSAR case, "both the initial under-prediction of the heat flux response, followed by an over-prediction during the rodinsertion is indicative of the fuel rod heat transfer being modeled differently in thevendor methods than in the Dominion model." It further stated that "the over-prediction of both nuclear power and heat flux will lead to conservative results at thelimiting point in the transient for both RCS over-pressurization and DNB during rodinsertion."Discuss the differences of the fuel rod heat transfer models used in the LR analysesfor the Dominion case and the FSAR case, and justify a higher peak RCS pressure of2,680.75 psia and a higher peak cladding temperature (PCT) of 1,760.0 0F predictedfor the Dominion case, compared to lower corresponding values of 2, 616.65 psia and1, 718.3 °F (shown in Table 4.2-2 and Table 4.2-4, respectively) for the FSAR case.DNC ResponseThe Dominion calculated RCS peak pressure value is 2,680 psia and the peakcladding temperature is 1,760 °F in the updated benchmarking analysis provided inAttachment 2.The Locked Rotor (LR) event is initiated by the instantaneous seizure of a reactorcoolant pump rotor resulting in a rapid reduction in flow through the affected RCSloop, reduction in heat transfer to the SG secondary with a resulting temperature and Serial No. 16-011IADocket No. 50-423Attachment 1, Page 4 of 11pressure increase of the RCS. As noted in the updated Attachment 5 Section 4.2Summary for the LR benchmark transient, the response differences between theDominion and FSAR cases are primarily attributable to loop friction losses and fuelrod modeling differences. Additional details relative to the higher peak ROS pressureand higher POT temperature observed for the Dominion cases are provided below.For the benchmark analysis, when the LR event occurs, flow begins to decreasethrough the faulted RCS loop and reactor vessel core as shown on updatedAttachment 5, Figures 4.2-3 and 4.2-5. About one second into the event, the reactortrips on low RCS flow, the unfaulted loops' reactor coolant pumps (RCPs) begin tocoast down, and flow reversal occurs in the faulted loop. It is during this intervalbetween one and two seconds, that the rate of pressure increase is reduced andmost of the pressure deviation develops between the FSAR and Dominion cases.The RCS cold leg and reactor vessel lower plenum pressures trend toward thepressurizer pressure as shown on the figure below where the Dominion pressurizerpressure response (which is not available for the FSAR case) is added to Attachment5, Figure 4.2-1, for reference. As shown on the figure, after the RCPs trip and flowdecreases, the pressure difference between the cold legs and the pressurizer isreduced. The differences between the FSAR and Dominion pressure responsessuggests that there are differences in the loss coefficients (RCS loops, reactorvessel, and surge line) and related RCP dynamics contributing to much of theobserved difference in peak pressure. It is noted that most of the difference developsbefore there is any appreciable reduction in reactor core power. After this timeinterval, which ends at about two seconds, the rate of pressure increase returns tonearly the same value as before the RCPs tripped and the difference between theFSAR and Dominion pressures remains relatively unchanged through the point ofpeak pressure. The opening response for the PSVs will also have an effect on peakpressure and become noticeable during the LR event when pressure is changingvery rapidly. Results of a sensitivity case in which the difference in peak pressurebetween the FSAR and Dominion cases was reduced, are discussed below.
Serial No. 16-011ADocket No. 50-423Attachment 1, Page 5 of 11Figure 4.2-1 SupplementLR -Reactor Vessel Lower Plenum Pressure2700.002600.002500.00! 2400.002:3007.0022007.0032100.00 40.001.00 2.00 3.00 4.00Time (sac)5.00Following the reactor trip, the control rods insert in approximately 1-4 seconds. Themost significant reductions in core power occur after about 2 seconds, after whichtime core power remains higher for the Dominion case until both peak ROS pressureand peak POT have occurred. This power response is affected by reactivityfeedback from fuel Doppler effects. For the Attachment 5 benchmark analysis to beconsistent with current FSAR methods, Doppler feedback effects for both cases weremodeled using a Doppler power coefficient (DPC). The Dominion method (VEP-FRD-41-P-A), however, uses a Doppler temperature coefficient (DTC) instead of aDPC. In order to better understand the difference in response between the FSARand Dominion cases, a sensitivity case was performed using the Dominion DTCmodel. The resulting power response is shown below as a supplement to Figure 4.2-6 from Attachment 5. As shown, the agreement in core power, particularly afterreactor trip, is much closer to the FSAR core response. In addition, the peak ROSpressure for the Dominion case was reduced from 2680 psia to 2666 psia and thesubsequent hot spot model case results in a reduction in peak POT from 1760°F to1739°F. It should be noted that the FSAR method also includes certain proprietarymodeling approaches that modify heat removal from the fuel rods which likelycontributed to the different response. Nevertheless, the sensitivity case indicates thatthe Dominion method produces results that compare favorably with the FSARresponse given the very rapid primary-side heatup associated with the LR event.
Serial No. 16-011ADocket No. 50-423Attachment 1, Page 6 of 11Figure 4.2-6 Supplement LR -Nuclear Power0.80Dom inion0.6z 0.400.20000.00 1 0(0 2.(00 30(0 4.00 5.00Tim~e (sec)
Serial No. 16-011IADocket No. 50-423Attachment 1, Page 7of 11RAI -12 (SRXB): Loss of Normal Feedwater Benchmark Analysis -Pressurizer Water Volume ResponseThe pressurizer water volume response shown in Figure 4.3-6 indicated thatDominion analysis predicts the same trends as the FSAR data, but calculates lowervalues in the period from 63 to 900 seconds, followed with a strong in-surge duringthe second heat-up period in the transient. The calculated maximum water volume of1588.96 ft3 is lower than the FSAR case of 1730.85 ft3.The licensee indicated thedeviations of the pressurizer water volume response are attributed to differences inthe main steam safety valves (MSSV) modeling, as well as differences in thepressurizer spray models.Discuss the differences in the MSSV model and pressurizer spray model used in theDominion analysis and the ESAR case and justify the deviations discussed above forthe pressurizer water volume response.DNC ResponseThe Dominion calculated maximum water volume is 1610 ft3 in the updatedbenchmarking analysis provided in Attachment 2.The loss of normal feedwater (LONF) event results in a reduction of SG secondary-side fluid mass and challenges the ability to adequately remove decay heat andstored energy from the primary side. For the benchmark analysis, the SG secondarypressurizes until steam release occurs through the MSSVs following a reactor trip onlow-low SG level. Auxiliary feedwater flow (AFW) flow is initially unable to adequatelyremove the primary-side energy resulting in a steady loss of SG fluid mass untildecay heat is sufficiently reduced, about 2000-2500 seconds into the event. Theprimary-side temperature and corresponding fluid volume respond to the balancebetween the generation of decay heat and the removal of energy through the SGsecondary via the MSSVs. The Dominion MSSV model includes the effect ofblowdown, which is not modeled for the FSAR case, resulting in a lower SG pressureand saturation temperature between approximately 70 and 1200 seconds, as shownin Figure 4.3-3. During this time, the difference between the SG pressure for theDominion case and the FSAR case is about 50 psi. This corresponds to a saturationtemperature difference of about 5°F, consistent with the primary-side temperaturedifference and associated pressurizer fluid volume early in this phase. For theDominion case, after 1200 seconds the MSSVs begin to cycle open and closed,raising the average pressure, but still remaining somewhat lower than the FSARcase. In addition, the lower SG pressures for the Dominion case result in slightlygreater AFW flow, which varies with SG back-pressure, also contributing todifferences in primary-side cooling. During the heatup phase, which lasts from about200 to 2500 seconds, the increase in primary-side temperature and related fluidvolume is greater for the Dominion case, but these values are lower at the start ofthis interval resulting in peak values that are lower.
Serial No. 16-011IADocket No. 50-423Attachment 1, Page 8 of 11Pressurizer sprays, heaters, and PORVs are assumed to function normally for theLONF event since this assumption yields more conservative results. During theheatup phase, the pressurizer liquid volume increases significantly for both cases;however, the resulting pressure is more suppressed for the FSAR case. TheDominion case results in opening of the pressurizer PORVs while for the FSAR case,pressurizer sprays alone are able to contain pressure. This could be attributable todifferences in the condensation or spray models and/or higher assumed spray flowrates. It is acknowledged that although a conservative spray model response cancontribute to higher pressurizer liquid volumes; this effect is considered to be small.RAI -14 (SRXB): Control Rod Bank Withdrawal at Power (RWAP) BenchmarkAnalysis -Higher Core Power Rate of IncreaseThe core power response in Figure 4.5-1 shows that for the RWAP 1 pcm/sec case,its rate of increase for the Dominion model is greater than the FSAR data. The fasterpower increase rate leads to the Dominion modeling tripping on high neutron flux atabout 73 seconds, and the lower power increase rate for the FSAR case results in areactor trip on an 0 TA T [Overtemperature delta T] signal at about 93 seconds.Discuss the differences of the nodding, input parameters, models or assumptionsused in analyses of the RWAP I pcm/sec case for the Dominion case and the ESARcase, and justify the greater increase of the core power rate observed in the analysisof the Dominion case.DNC ResponseThe Dominion case trips on high neutron flux at about 74 seconds in the updatedbenchmarking analysis provided in Attachment 2.The RWAP event involves the inadvertent addition of core reactivity caused by thewithdrawal of rod cluster control assembly (RCCA) banks. The RWAP benchmarkcase referenced in RAI-14, assumes a relatively slow reactivity insertion rate of 1pcm/sec, resulting in a steady increase in core power and related primary systemheatup. The power response, in turn, is affected by moderator and Doppler reactivityfeedback effects. For the benchmark analysis, moderator reactivity feedback isassumed to be zero for both cases. For Doppler reactivity, the Dominion model usesa DTC while the FSAR model uses a DPC with minimum reactivity feedbackconservatively assumed for both cases. Although fuel temperature and core powerboth increase in a similar manner prior to the reactor trip, the relative effect of thatincrease and the associated modeling of the feedback effect may vary in therespective Dominion and ESAR Doppler models. It is also noted that the reactor coremodel used in the FSAR case incorporates proprietary mechanisms to modify theremoval of heat from the core. Based on the core power response shown inAttachment 5 Figure 4.5-1, the differences result in greater reactivity feedback and aslower increase in power for the FSAR case.
Serial No. 16-011ADocket No. 50-423Attachmentl1, Page 9ofl11Selection of a 1 porn/second reactivity insertion rate is intended to be representativeof a slow insertion rate case. The dynamic effects of the RWAP event at a giveninsertion rate are different for the Dominion and FSAR methodologies as notedabove. The critical characteristic of the event analysis is determination of theminimum departure from nucleate boiling ratio (DNBR) which occurs for a casewhere reactor trip times for the overtemperature AT (OTDT) and high neutron flux tripsignals coincide. This point will occur at different reactivity insertion rates for the twomethodologies.RAI -15 (SRXB): RWAP Analysis -DNBR CalculationsThe results of MPS3 FSAR Chapter 15 non-LOCA analyses indicated the RWAPevent is the most limiting event in terms of the margin to the safety limit DNBR in thecategor'y of the anticipated operating occurrences (A QOs). Since the licensee alsoproposed to use the RETRAN and Dominion VIPRE-D method to perform DNBRcalculations for assessing the fuel integrity during AOOs and accidents, the RETRANbenchmarking analysis for both RWAP I pcmlsec and 100 pcm/sec cases should beperformed to include the results of the DNBR calculation by using the DominionVIPRE-D method. The requested information includes a comparison of DominionVIPRE-D analyses to the MPS3 FSAR analysis of record (A OR) showing that thecalculated DNBRs for both cases are compatible with the AOR and the allowablerange of the use of the NRC-approved DNBR correlation in VIPRE-D for theDominion method is not exceeded.DNC ResponseIn a December 15, 2015 teleconference between DNC and NRC staff, the NRCagreed it would be sufficient to provide the requested comparison for one reactivityinsertion rate case versus both cases as stated in the request.The Dominion RETRAN and VIPRE-D results for the 1 pcm/sec insertion rate RWAPcase are chosen to represent the Dominion DNBR evaluation of the RWAP transient.The VIPRE-D model input file for this case was developed in accordance withDominion VIPRE-D Topical DOM-NAF-2-P-A. The resulting transient DNBR plot wascompared to the DNBR plot shown in MPS3 ESAR Figure 15.4-9. Both the Dominionand FSAR DNBR plots show a comparable trend. The observed differences areprimarily due to the dynamic effects associated with the RETRAN benchmark results,as noted in the response to RAI-14. The core power rate of increase in the 1pcm/sec case for the Dominion RETRAN model is greater than the FSAR data suchthat the reactor trip occurs approximately 20 seconds earlier. The inverse effect ofpower on DNB is clearly observed in the transient DNBR plot below, and theminimum DNBR values for the Dominion and FSAR cases are observed to becomparable. In addition, the thermal-hydraulic conditions of the RWAP transientanalyzed were confirmed to be within the validation range of the NRC-approved
~Serial No. 16-011IADocket No. 50-423Attachment 1, Page 10 of 11DNBR correlations utilized in the VIPRE-D model (WRB-2M and ABB-NV) consistentwith the limitations on the use of DOM-NAF-2-P-A.Figure 4.5-7 Supplement RWAP -1 pcm/sec DNBR3.9 i-l--1pcrnFSARDNBRl--- 1 pcm Diora3.4 1.i,, 2.92!2;.4 I0 20 40 60 80 100 120TIME %SEC)RAt -16 (SRXB): Feedwater Line Break AnalysisMPS3 ESAR (2012 Version), Section 15.2.8 discussed the feedwater line break(FLB) analysis for both cases with and without offsite power available. The FSARAOR presented transient results including nuclear power, core heat flux, totalreactivity, pressurizer pressure, total RCS flow, feedwater break flow, looptemperature, intact loop temperature, and SG pressure. FSAR Figures 15.2-13 and15.2-19 indicated that a post-trip return-to-power will occur for the case with offsitepower available, and core will remain subcritical throughout the transient time ofseveral seconds for the case without offsite power available. Also, page 1 5-2-16indicated that the FLB is the most limiting event in the decrease in secondaryremoval category. The analysis of the FLB will use a broad scope of the models inRETRAN, including feedwater break flow model, RC pumps coastdowm model, SGheat transfer model, and reactivity feedback model.Perform the RETRAN benchmarking analysis for the FLB event for both cases withand without offsite power available. The information to be provided should showthat: the values of the plant parameters and assumptions used in the Dominion FLBanalysis are consistent with that used in the FSAR AOR; the results are compatiblewith the AOR; and there is no unexplainable thermal-hydraulic phenomenathroughout the transients.
Serial No. 16-011ADocket No. 50-423Attachment 1, Page 11 of 11DNC ResponseThe additional benchmarking analysis has been performed for the FLB event forcases with and without offsite power available. The discussion of the event analysis,including inputs and assumptions and results, as compared with the FSAR analysisare included in Section 4.6 of the update to Attachment 5 enclosed in Attachment 2.
Serial No. 16-011ADocket No. 50-423ATTACHMENT 2RETRAN BENCHMARKING INFORMATION -UPDATEDDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3
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2RETRAN Benchmarking Information
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2 Page 2 of 61TABLE OF CONTENTS1.0 INTRODUCTION AND SUMMARY ......................................................... 31.1 INTRODUCTION........................................................................................ 31.2 SUMMARY .............................................................................. ............... 32.0 MPS3 RETRAN MODEL...................................................................... 43.0 METHOD OF ANALYSIS ...................................................................... 84.0 BENCHMARKING ANALYSIS RESULTS................................................. 94.1 Loss OF LOAD/TURBINE TRIP............. ............... .......................................... 94.2 LOCKED ROTOR ..................................................................................... 144.3 Loss OF NORMAL FEEDWAThR ............................................................ ....... 244.4 MAIN STEAM LINE BREAK .............. .......................................................... 314.5 CONTROL ROD BANK WITHDRAWAL AT POWER................................................ 424.6 MAIN FEED WATER LINE BREAK .................................................................. 495.0 CONCLUSIONS................................................................................ 616.0 REFERENCES ......................... ....................................................... 61
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2Pae3olPage 3 of 611.0 Introduction and Summary1.1 IntroductionTopical report VEP-FRD-41-P-A, "VEPCO Reactor System Transient Analyses Using theRETRAN Computer Code," (Reference 1) details the Dominion methodology for Nuclear SteamSupply System (NSSS) non-LOCA transient analyses. This methodology encompasses the non-LOCA licensing analyses required for the Condition I, II, III, and IV transients and accidentsaddressed in the Final Safety Analysis Report (FSAR). The VEP-FRD-4 1-P-A methods are alsoused in support of reload core analysis. In addition, this capability is used to perform best-estimate analyses for plant operational support applications. The material herein supports theapplicability assessment of the VEP-FRD-4 1-P-A methods to Millstone Power Station Unit 3(MPS3) for the stated applications.1.2 SummaryThis attachment provides a description of the RETRAIN base model for MPS3 and results Ofbenchmarking analyses using this model. The MPS3 model was developed in accordance withthe methods in VEP-FRD-4 1-P-A, with certain noding changes noted below. This assessmentconfirms the conclusion that the Dominion RETRAIN methods, as documented in topical reportVEP-FRD-4 1-P-A, are applicable to MPS3 and can be applied to MPS3 licensing analysis forreload core design and safety analysis. Dominion analyses of MPS3 will employ the modeling inVEP-FRD-4 1-P-A, as augmented with the noding changes listed below. Thus, VEP-FRD-41-P-A, as augmented, is the Dominion methodology for analyses of non-LOCA NSSS transients forMPS3.The MPS3 RETRAN base model contains the following alterations in noding with respect to themodeling that is documented in VEP-FRD-4 1-P-A.a) The MPS3 model explicitly models the safety injection (SI) accumulators.b) The MPS3 model has separate volumes for the steam generator inlet and outlet plenums.c) The MPS3 model includes cooling paths between dowucomer and upper head.
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2Pae4olPage 4 of 612.0 MPS3 RETRAN ModelThe MPS3 RETRAN-3D Base Model and associated model overlays are developed usingDominion analysis methods described in the Dominion RETRAN topical report (Reference 1).The Dominion analysis methods are applied consistent with the conditions and limitationsdescribed in the Dominion topical report and in the applicable NRC Safety Evaluation Reports(SERs).The MPS3 Base Model noding diagram for a representative loop is shown on Figure 2-1.Volume numbers are circled, junctions are represented by arrows, and the heat conductors areshaded. This model simulates all four reactor coolant system (RCS) loops and has a single-nodesteam generator (SG) secondary side, consistent with Dominion methodology. The SG primarynodalization includes 10 steam generator tube volumes and conductors. There is a multi-nodeSG secondary overlay that can be added to the Base Model for sensitivity studies although noneof the analysis results presented herein utilize this overlay.In addition to the base MPS3 model, an overlay deck is used to create a split reactor vesselmodel to use when analyzing Main Steam Line Break (MSLB) events, consistent with Dominionmethodology. This overlay adds volumes to create a second, parallel flow path through theactive core from the lower plenum to the upper plenum such that RCS loop temperatureasymmetries can be represented. This noding is consistent with the method described in VEP-FRD-4 1-P-A. A noding diagram of the split reactor vessel is shown on Figure 2-2.The base MPS3 model noding is virtually identical to the Surry (SPS) and North Anna (NAPS)models with the exception of some minor noding differences listed as follows.a) The MPS3 model explicitly models the SI accumulators.b) The MPS3 model has separate volumes for the SG inlet and outlet plenums.c) The MPS3 model includes cooling paths between downcomer and upper head.The SI accumulators are part of the MPS3 model because injection from the accumulators occursin the current FSAR analysis for MSLB. The use of separate volumes for the inlet and outletshould have little effect on transient response since the fluid temperature in these volumes isgenerally the same as the connecting RCS piping. The cooling paths are included toappropriately model upper head T-cold conditions.The Dominion models, including the MPS3 model, have some differences compared to thevendor RETRAN model that was used to perform the current FSAR analyses. Table 2-1 and thesubsequent text discussion provide an overview of these differences. Additional details
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2Pae5olPage 5 of 61concerning differences between the Dominion MPS3 and FSAR RETRAN models are discussedin the benchmarking analyses in Section 4.A description of the Dominion RETRAN methodology is provided in Reference 1, wherespecific model details are discussed in Sections 4 and 5 of that reference.Table 2-1RETRAN Model Comnarison of Key CharacteristicsParameter Dominion FSARCode Version: RETRAN-3D in "02 mode" RETRAN-02Noding:Reacor VsselSingle flow path (special split coreReco esloverlay for MSLB only) Multiple parallel flow pathsSingle node secondary. Five axiallevels (10 nodes) for SG tubesSteam Generator primary side. Local Conditions Heat Multi-node secondary.Transfer model available for loss ofheat sink events.Reactivity ModelDoppler-only power coefficientDopper eedack Doppler temperature coefficient that and a Doppler temperatureis a function of TFUEL. coefficient effect driven bymoderator temperature.Moderator Feedback Moderator temperature coefficient Moderator density coefficientANS 1979 StandardU-235 with 1500 day burn.AN 199SadrDecay Heat Q = 190 MeV/fission.Boundss additional 2o uncertainty
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2Attahmet 2pg. 6 of 61Figure 2-1 MPS3 Base Model Nodalization DiagramMOSlt*Vo la mc
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2pg. 7 of 61Figure 2-2MPS3 Split Vessel Nodalization
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2pg8o6pg. 8 of 613.0 Method of AnalysisValidation of the Dominion MPS3 RETRAN method involves comparison of RETRANanalyses to the MPS3 FSAR analysis of record (AOR) for select events. The Dominionanalyses presented herein are not replacements for the existing AORs. These eventsrepresent a broad variation in behavior (e.g. RCS heatup, RCScooldown/depressurization, reactivity excursion, loss of heat sink, etc.), and demonstratethe ability to appropriately model key phenomena for a range of transient responses. Thetransients selected for comparison with their corresponding MPS3 FSAR section areprovided in Table 3-1. For each transient, an analysis is performed using the DominionMPS3 RETRAN model and compared with the current FSAR analysis. Initial conditionsand inputs are established for each benchmark to provide an adequate comparison ofspecific transient behavior.Table 3-1 Transients Analyzed for FSAR ComparisonTransient MPS3 FSAR SectionMain Steam Line Break 15.1.5Loss of Load/Turbine Trip 15.2.3Loss of Normal Feedwater 15.2.7Locked Rotor 15.3.3Control Rod Withdrawal at Power 15.4.2Main Feedwater Line Break 15.2.8
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2pg9o6pg. 9 of 614.0 Benchmarking Analysis ResultsA summary for each transient comparison is presented in the following sections. Includedin each section is an input summary identifying key inputs and assumptions along withdifferences from FSAR assumptions. A comparison of the results for key parameters isprovided with an explanation of key differences between the Dominion and FSAR cases.4.1 Loss of Load/Turbine TripThe Loss of Load/Turbine Trip (LOL) event is defined as a complete loss-of-steam load andturbine trip from full power without a direct reactor trip, resulting in a primary fluidtemperature rise and a corresponding pressure increase in the primary system. This transientresults in degraded steam generator heat transfer, reactor coolant heatup and pressureincrease following a manual turbine trip.The LOL transient scenario presented here was developed to analyze primary RCSoverpressurization. It is initiated by decreasing both the steam flow and feedwater flow tozero immediately after a manual turbine trip. The input sumnmary is provided in Table 4.1-1.Table 4.1-1ILOI Tnniit SulmmaryParameter Value NotesInitial Conditions_______Core Power (MW) 3723 Includes 2% uncertaintyRCS Flow (gpm) 363,200 Thermal DesignVessel TArG (F) 576.5 Low Tavg plus uncertaintyPressurizer Pressure (psia) 2200 Includes -50 psia uncertaintyPressurizer Level (%) 52.5 Low Tavg Target plus uncertaintySG Level (%) 50.0 NominalSG tube plugging (%) 10 MaximumPump Power (MW/Pump) 5.0 MaximumAssumptions/Configuration_______Reactor trip -only Hi Pzr Pressure is activeAutomatic rod control -Not creditedPressurizer sprays, PORVs -Not creditedMain steam dumps, SG PORV -Not creditedAFW flow -Not creditedReactivity ParametersDoppler Reactivity Feedback Least NegativeModerator Feedback Most Positive
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2 p.lo~pg. 10 of 61Results -LOLPressure in the RCS increases during a LOL due to degraded heat transfer in the steamgenerator and is alleviated only when the pressurizer safety valves (PSV) open as well as themain steam safety valves (MSSV). The pressurizer pressure response is shown on Figure4.1-1, RCP outlet pressure in Figure 4.1-2, and the peak RCS pressure values are listed inTable 4.1-2. The Dominion case predicts a pressunizer pressure and RCP outlet pressureresponse that agrees very well with the FSAR results past the point of peak RCS pressure.Following the initial decrease in primary system pressure, the FSAR pressure levels outwhere the Dominion case results continue to decrease. The difference is due to differingsecondary safety valve modeling in the vendor model, specifically in that the Dominionmodel includes the modeling of blowdown in the main steam safety valves and the vendormodel does not. Hence, more energy is removed through the secondary system in theDominion case once the main steam safety valves actuate than is removed from thesecondary system in the vendor model.Figure 4.1-3 shows the power response is nearly identical both before and after the reactortrip on high pressurizer pressure and control rod insertion. The Dominion case trips slightlyearlier than the FSAR data because of the higher RCS pressurization rate.The Dominion model vessel inlet temperature, Figure 4.1-4, and coolant averagetemperature, Figure 4.1-5, agrees in trend and rate of increase although the response lags theFSAR response before the inlet temperature peaks at a slightly lower value. This indicatesthat the FSAR steam generator heat transfer degrades sooner than what is predicted byDominion model and is attributed to the difference expected between the use of a multi-node steam generator (MNSG) in the FSAR model and the single-node steam generator(SNSG) model employed in the Dominion model. Overall, both the Dominion model andFSAR models exhibit similar trends in the temperature responses and the differences haveno effect on peak RCS pressure.Table 4.1-2 LOL RCS Overpressure ResultsParameter Dominion FSARSequence of Events:High Pressurizer Pressure Setpoint 5.6 6.2Reached (see)Peak RCS Pressure (sec) 9.2 9.9Peak RCS Pressure (psia) 2705 2725I
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2 p.lo6pg. 11 of 61Figure 4.1-1 LOL -Pressurizer Pressure27002600" 2400S2300220021002000010 20 30 40Time (sec)50Figure 4.1-2 LOL -RCP Outlet Pressure3000290028002700* 260022500224002300220021002000010 20 30 40 50Time (sec)
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2 p.1o6pg. 12 of 61Figure 4.1-3LOL -Nuclear Power1.21.0S0.80.0.20.0010 20 30 40Time (sec)50Figure 4.1-4LOL- Vessel Inlet Temperature590580570'07560~3550E)I- 540530520510 40.0010.00 20.00 30.00 40.00Time (sec)50.00
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2 p.1o6pg. 13 of 61Figure 4.1-5 LOL -Vessel Average Temperature600595590o- 585580E5- 575570565560010 20 30 40 50Time (sec)Summary -LOLThe Dominion MPS3 analysis provides results that are similar to the FSAR analysis for theLOL event. The RCS peak pressures are essentially the same although the pressure divergesomewhat later in the event after pressure relief begins due to differences in MSSVmodeling. There are small differences in the RCS temperature response due to differencesin the SG models, however, this has no effect on the RCS peak pressure. The DominionMPS3 analysis is presented for benchmark comparison, and does not replace the existingAOR.
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2 p.1o6pg. 14 of 614.2 Locked RotorThe Locked Rotor / Shaft Break (LR) event is defined as an instantaneous seizure of aReactor Coolant Pump (RCP) rotor, rapidly reducing flow in the affected reactor coolantloop leading to a reactor trip on a low-flow signal from the Reactor Protection System. Theevent creates a rapid expansion of the reactor coolant and reduced heat transfer in the steamgenerators, causing an insurge to the pressurizer and pressure increase throughout thereactor coolant system (RCS).The LR transient scenario presented here was developed to analyze primary RCSoverpressurization. It is initiated by setting one RCP speed to zero as the system isoperating at full power. The reactor coolant low ioop flow reactor trip is credited, with asetpoint of 85% of the initial flow. The input summary is provided in Table 4.2-1. Most ofthe input parameters are the same as those used in the FSAR Chapter 15 analyses.Table 4.2-1 LR Input SummaryParameter Value NotesInitial ConditionsCore Power (MW) 3723 Includes 2% uncertaintyRCS Flow (gpm) 363,200 Thermal Design FlowVessel TAVG (F) 594.5 Nominal + 5*FPressurizer Pressure (psia) 2300 Includes +50 psia uncertaintyPressurizer Level (%) 64 NominalSG Level (%) 50 NominalAssumptions/Configuration _______Reactor trip _______Only Low RCS Loop Flow is creditedAutomatic rod control _______Not creditedPressurizer sprays, PORVs -Not creditedMain steam dumps, SG PORV -Not creditedAFW flow Not creditedSG tube plugging (%) 10' Max valueReactivity ParametersDoppler Reactivity Feedback Most Negative Dominion model adjusted to use FSAR_____________________ __________Doppler Power CoeffcientModerator Feedback Most Positive ________________' Original benchmark case inadvertently assumed 0% SG tube pluggingResults -LR RCS Overpressure CasePressure in the RCS increases during a LR event due to degraded heat transfer in the steamgenerator and is alleviated only when the pressurizer safety valves (PSV) open. Themagnitude of the Dominion model pressure response both in the reactor vessel lowerplenum, Figure 4.2-1, and at the RCP exit, Figure 4.2-2, is greater than the FSAR modelresponse, while following the same trends as the FSAR data. At the limiting point in the
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2 p.1o6pg. 15 of 61transient response, the Dominion model conservatively predicts a pressure approximately 63psi greater than the FSAR model in the reactor vessel lower plenum. The difference betweenthe Dominion model and FSAR model's peak responses is the same at the RCP exit as in thelower plenum.The Dominion faulted loop flow response (Figure 4.2-3) and unfaulted loop flow response(Figure 4.2-4) are in good agreement with the FSAR model response up to or just beyondthe point of rod insertion. Following reactor trip there is some divergence in the unfaultedloop flow trends, which are consistent with the core heat flux predictions and assumedminor differences in the loop friction losses between the Dominion and FSAR models.With respect to the faulted loop flow response, the maximum reverse flow seen in the FSARmodel is slightly greater than seen in the Dominion model, which is also attributed to smalldifferences in the loop friction losses between the Dominion and FSAR models.For the total core inlet flow response (Figure 4.2-5), the Dominion model predicts a lowerflow than the FSAR model for approximately the first 4 seconds of the transient. After 4seconds the FSAR and Dominion model core flow responses cross and the Dominion modelpredicts a slightly higher core flow rate. The limiting point in the transient occurs prior to 4seconds such that RETRAN-3D produces a more limiting response than the FSAR modelfor the Locked Rotor/Shaft Break event.The nuclear power response, Figure 4.2-6, predicted by the Dominion model agrees wellwith the FSAR data, with the Dominion model response slightly over predicting powerduring rod insertion following the reactor trip on low RCS flow. Similarly, the Dominionmodel core heat flux response, Figure 4.2-7, also slightly over predicts the FSAR model'sresponse in the same time frame during control rod insertion. Additionally, the Dominionmodel heat flux response shows a slightly larger decrease at the initiation of the event overthe decrease seen in the FSAR data. Both the initial unader prediction of the heat fluxresponse, followed by an over prediction during the rod insertion is indicative of the fuel rodheat transfer being modeled differently in the FSAR methods than in the Dominion model.However, the over prediction of both nuclear power and heat flux will lead to conservativeresults at the limiting point in the transient for both RCS overpressurization and DNB duringrod insertion. Overall the nuclear power and heat flux predictions are very similar.A summary of the LR transient analysis comparison is provided in Table 4.2-2.
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2 p.1o6pg. 16 of 61Table 4.2-2 LR RCS Overpressure ResultsParameter Dominion FSARSequence of Events:Low RCS Flow Setpoint Reached (sec) 0.1 0.1Rods Begin to Drop (sec) 1.1 1.1Peak RCS Pressure (sec) 3.8 4.1Peak RCS Pressure (psia) 2680 2617Summary -LR RCS Overpressure CaseThe Dominion Millstone analysis provides responses that are similar to the FSAR analysisfor the LR event, with the Dominion model predicting higher peak RCS pressures.Differences are attributed to ioop friction losses and fuel rod modeling differences. TheDominion MPS3 analysis is presented for benchmark comparison, and does not replace theexisting AOR.
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2 p.1o6pg. 17 of 61Figure 4.2-1 LR -Reactor Vessel Lower Plenum Pressure2700265026002550"' 25002450240023502300225022000 5 10 15 20Time (sec)Figure 4.2-2 LR -RCP Outlet Plenum Pressure2700265026002550.' 2500v 2450S2400235023002250220005 10 15 20Time (sec)
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2 p.1o6pg. 18 of 61Figure 4.2-3 LR -Faulted Loop Normalized Flow0It-")N.0Z1.201.000.800.600.400.200.00-0.20-0.40-0.60Time (sec)Figure 4.2-4 LR -Unfaulted Loop Normalized Flow0U-~0ci)N0z1.161.060.960.860.760.660.560.460 5 10 15 20Time (sec)
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2 p.1o6pg. 19 of 61Figure 4.2-5 LR -Core Inlet Normalized Flow1.101.000.900.80 .2 0.70IL___ 0.60zO0.50-Dominion0.40 -FA0.30 , , , -05 10 15 20Time (sec)Figure 4.2-6 LR -Nuclear Power1.111.010.910.810.710.. 0.61NS0.51Oz 0.410.31 kI"--Dmn~I0.1105 10 15 20Time (sec)
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2 p.2o6pg. 20 of 61Figure 4.2-7LR -Core Heat FluxxII"1-Ntu0Z1.121.020.920.820.720.620.520.420.320.220.1205 10 15 20Time (sec)
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2 p.2o6pg. 21 of 61LR Peak Cladding TemperatureThe Locked Rotor event is also analyzed to demonstrate that a coolable core geometry ismaintained. A hot spot evaluation is performed to calculate the peak claddingtemperature and oxidation level. The Dominion Hot Spot model is described in TopicalReport VEP-NFE-2-A, "VEPCO Evaluation of the Control Rod Ejection Transient."(Reference 2). The Dominion Hot Spot model was used to evaluate the MPS3 PCT andoxidation level for the LR event.The Dominion hot spot model is used to predict the thermal-hydraulic response of thefuel for a hypothetical core hot spot during a transient. The hot spot model describes aone-foot segment of a single fuel rod assumed to be at the location of the peak corepower location during a transient. The hot spot model uses boundary conditions from theLR system transient analysis to define inlet flow and core average power conditions. Thehot spot model uses MPS3-specific values for fuel dimensions, fuel material properties,fluid volume, and junction flow areas.The hot spot model is run to 0.1 seconds and a restart file is saved. Upon restart, thefuel/cladding gap conductance (thermal conductivity) is modified to simulate gap closureby setting the gap heat transfer coefficient to 10,000 Btu/ft2-hr-°F for a gap conductanceof 2.708 Btulft-hr-°F. The hot spot model input summary is provided in Table 4.2-3. Mostof the input parameters are the same as those used in the FSAR Chapter 15 analyses. Wheredifferences from the FSAR inputs exist, they are indicated in the Notes column.Table 4.2-3 Hot Spot Model Input SummaryParameter Value NotesComputer Code Used RETRAN-3D FSAR uses VIPREInitial Conditions ________Ratio of Initial to Nominal Power 1.02RCS Flow (gpm) 363,200Hot Spot Peaking Factor 2.60Assumptions/ConfigurationPre-DNB Film Heat Transfer Coefficient ThornTime of DNB (sec) 0.1Post DNB Film Boiling Heat Transfer Bishop-Sandberg-Coefficient Tong___________Fuel Pin ModelPost DNB Gap Heat Transfer Coefficient 10,000(Btu/hr-ft2-°F)_________ _____ _______Gap Thermal Expansion Model activated? YesZircaloy-Water Reaction activated? Yes
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2Attahmet 2pg. 22 of 61LR Peak Cladding Temperature ResultsThe peak cladding temperature obtained from Dominion's MPS3 hot spot model for thelocked rotor event is 1760 0F. The maximum zircaloy-water reaction depth is3.60875E-06 feet, which corresponds to approximately 0.19% by weight based on thenominal cladding thickness of 1 .875E-03 feet. A summary of the LR Peak CladdingTemperature Hot Spot analysis comparison is provided in Table 4.2-4. The cladding innersurface temperature is shown in Figure 4.2-8.Table 4.2-4 LR Hot Spot ResultsParameter Dominion FSARPeak Cladding Temperature 1760 0F 1718 0FMaximum Zr-water reaction (w/o) 0.19 0.22The Dominion peak cladding temperature and maximum oxidation values are comparable tothe FSAR values. The Dominion MPS3 analysis is presented for benchmark comparison,and does not replace the existing AOR.
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2 p.2o6pg. 23 of 61Figure 4.2-8LR Hot Spot -Cladding Inner Surface Temperature17501550'-1350,- 1150E9507500.15.1 10.1 15.1Time (sec)
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2 p.2o6pg. 24 of 614.3 Loss of Normal FeedwaterThe Loss of Normal Feedwater (LONE) event causes a reduction in heat removal from theprimary side to the secondary system. Following a reactor trip, heat transfer to the steamgenerators continues to degrade resulting in an increase in RCS fluid temperature and acorresponding insurge of fluid into the pressurizer. There is the possibility of RCS pressureexceeding allowable values or the pressurizer becoming filled and discharging waterthrough the relief valves. The event is mitigated when Auxiliary Feedwater (AFW) flow isinitiated and adequate primary to secondary side heat removal is restored. This analysisshows that the AFW system is able to remove core decay heat, pump heat and stored energysuch that there is no loss of water from the RCS and pressure limits are not exceeded. TheLONE input summary is provided in Table 4.3-1.Table 4.3-1 LON Iput SummaryParameter Value NotesInitial ConditionsCore Power (MW) 3723 Includes 2% uncertaintyRCS Flow (gpm) 363,200 Thermal Design FlowVessel TAVG (F) 583 FSAR valueRCS Pressure (psia) 2300 Nominal + 50 psiPressurizer Level (%) 71.6 Nominal + 7.6%SG Mass -89000 Dominion model adjusted to beconsistent with FSAR analysisAssumptions/ConfigurationLow-Low Level Reactor Trip Setpoint 0% Percent of narrow range spanPressurizer: sprays, heaters, PORVs -Assumed operableAFW Temperature (F) 120 Max valueAFW Pump configuration -2 motor-driven pumps feed 4 SGsAuxiliary feedwater flow rate (gpm) -Variable as function of SG press.Local Conditions Heat Transfer model active SG secondary side__________________________FSAR= multi-node SGDecay Heat -FSAR decay heat constants are_____________________________applied for this caseReactivity ParametersDoppler Reactivity Feedback Most negative Dominion model adjusted to use__________________________ _________FSAR Doppler Power CoeffcientModerator Feedback Most PositiveI
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2 p.2o6pg. 25 of 61Results -LONFThe results for the LONF comparison analysis are presented in Table 4.3-2 and Figures4.3-1 through 4.3-7. The loss of feedwater flow to the steam generators (SG) results in areduction in SG level until a reactor trip occurs on Low-Low SG level. Normalized poweris shown on Figure 4.3-1 and normalized core heat flux in Figure 4.3-2. The nuclear powerresponse and heat flux response predicted by the Dominion model are in excellentagreement with the FSAR data, indicating that the scram on low-low steam generator leveloccurred at essentially the same time shown for the FSAR data. The results continue todemonstrate good agreement through the end of the event.Figure 4.3-3 shows the steam generator pressure response. The Dominion steam generatorpressure is initialized at a slightly different pressure than the FSAR model because theDominion model initial condition is adjusted to minimize the steam generator areaadjustment. Between 10 and 34 seconds the FSAR pressure increases more rapidly to apressure --43 psi greater than the Dominion model prediction when the steam line is isolated.This difference is attributed to differing heat transfer degradation in the MNSG model usedin the FSAR analysis versus the SNSG model used in the RETRAN-3D model. Steam lineisolation occurs at nearly the same time, causing pressure to increase rapidly. The peakpressure is limited by the main steam safety valves (MS SVs), resulting in an almostidentical peak pressure in both the Dominion and FSAR responses. However, the Dominionmodel pressure decreases following the peak value, where the FSAR model responseremains at a constant value near the peak value, due to differences in MSSV modeling.Figure 3.1-4 shows the steam generator liquid mass. The steam generator liquid massdepletes faster in the Dominion cases than in the FSAR cases. This is consistent with theincreased relief flow as shown in the steam generator pressure response.The response in the pressurizer is shown in Figures 4.3-5 and 4.3-6. Between the FSAR andDominion model, the pressure responses are in good agreement until around 45 -50seconds where the Dominion pressure is lower than the FSAR., reflecting less heat transferdegradation during this period. This is followed by a second pressure peak that is higher forDominion than the FSAR. Based on the sharpness of the Dominion peak compared withthe FSAR data, this difference is most likely driven by differences in the pressurizer spraymodels and primary to secondary heat transfer.For the pressurizer water volume, shown in Figure 4.3-6, the Dominion model results followthe same trends as the FSAR data, but drops lower in the period from 63 to 900 seconds,then demonstrates a strong insurge during the second heat-up period in the transient whilepeaking at a somewhat lower value than the FSAR. The difference seen in the pressurizer
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2pg. 26 of 61volume results is primarily due to the previously discussed MSSV modeling differencesand the resultant increased steam release from the Dominion model compared to the FSARmodel as well as possible differences in the pressurizer spray models.Table 4.3-2 LONF ResultsParameter Dominion FSARPeak PZR Liquid Volume (ft3) 1610 1730
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2Figure 4.3-1 LONF -Nuclear Powerpg. 27 of 6lo1~0Z)1.201.000.800.600.400.200.00110 100 1000Time (sec)Figure 4.3-2 LONF -Normalized Core Heat Flux100001.20-1.00-0.80S0. 60LI)0.200.00-1.0010.00 100.00 1000.0010000.00Time (sec)
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2Attahmet 2pg. 28 of 61Figure 4.3-3 LONF -Steam Generator PressureCt)Co091330128012301180113010801030980930880110 100 1000Time (sec)10000Figure 4.3-4LONF -Steam Generator Liquid Mass10000090000800007000060000co50000-~40000*J3000020000100000110 100 1000Time (sec)10000
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2Attahmet 2pg. 29 of 6lFigure 4.3-5 LONF -Pressurizer Pressure0-.E--0..-25002450240023502300225022002150180017001600150014001300120011001000110 100 1000Time (sec)Figure 4.3-6 LONF -Pressurizer Water Volume10000110 100 1000100Time (sec)10000
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2 p.3o6pg. 30 of 61Figure 4.3-7 LONF -Loop Average Temperature595590EI-585580575570110 10O0 1000Time (sec)10000Summary -LONFThe Dominion analysis provides results that are similar to the FSAR analysis for the LONFevent. The major differences result from the main steam safety relief valve modeling, whichresults in higher steam releases and a subsequent increase in heat transfer following thereactor trip. In addition, the steam generator nodalization and related heat transfer along withother modeling differences such as pressurizer spray also affect the transient response.These effects are cumulative resulting in a somewhat smaller long-term pressurizer insurgeand higher pressurizer pressure peak compared to the FSAR results. The Dominion MPS3analysis is presented for benchmark comparison, and does not replace the existing AOR.
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2pg31flpg. 31 of 614.4 Main Steam Line BreakThe Main Steam Line Break (MSLB) event is a rupture in the main steam piping resulting ina rapid depressurization of the SG secondary and corresponding cooldown of the primary.The temperature reduction results in an insertion of positive reactivity with the potential forcore power increase and DNBR violation.The MSLB transient scenario presented here is modeled as an instantaneous, double-endedbreak at the nozzle of one steam generator from hot shutdown conditions with offsite poweravailable. The input summary is provided in Table 4.4-1.Table 4.4-1 MSLB Input SummaryParameter Value NotesInitial ConditionsCore power (MW) ~-1% H-ZPPump power (MW) 0.0RCS Flow (gpm) 363,200 Thermal Design FlowVessel TAVG (F) 557 H-ZP nominalRCS Pressure (psia) 2250 NominalPressurizer Level (%) 28 HZP nominalSG Level (%) 50 NominalAssumptions/ConfigurationHeat transfer option Forced HT Map FSAR uses a proprietary heat(note 1) transfer formulationMain feedwater flow (% HFP value) 100 initiated at time 0 secAuxiliary feedwater flow rate (gpm) Max initiated at time 0 secSG tube plugging (%) 0 Minimum valueReactivity ParametersRWST Boron Credited FSAR does not credit boron from_____________________________ ____________the SI systemAccumulator Boron Not CreditedDoppler Reactivity Feedback Doppler Only FSAR -Doppler power defectPower defect, plus DTC included in moderatorDTC model density feedback_____________________________ disabledModerator Feedback Moderator Moderator density feedback____________________________density feedback_______________1 -Dominion method maximizes heat transfer coefficients for the faulted SG secondary side.
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2 p.3o6pg. 32 of 61Results -MSLB with Offsite Power AvailableThe faulted loop steam flow and steam generator pressure responses shown in Figure4.4-1 and Figure 4.4-3 match the FSAR data reasonably well with the steam flow andpressure in the Dominion model remaining somewhat higher than the FSAR data. This ispartly caused by the slightly larger break junction area and the higher initial steampressure for the Dominion model. In addition, the Dominion model uses conservativelyhigh heat transfer coefficients in the faulted steam generator, which allow the faultedsteam generator to pull heat faster from the primary side.The Intact loop steam flow (Figure 4.4-2) shows a different response due to differences inthe MSIV closure. In the Dominion model, the MSIVs close linearly over 10 seconds,while the FSAR model uses a delay of 10 seconds to conservatively increase RCSovercooling. The initial steam flow is higher for the Dominion case, decreasing belowthe FSAR value as the MSIVs close. The steam generator mass and pressure responses,shown in Figure 4.4-8 and Figure 4.4-4, reveals the differences in MSIV modeling withthe Dominion model releasing somewhat less liquid inventory prior to valve closure.For both the faulted and intact loops the main feedwater and auxiliary feedwaterresponses (Figure 4.4-5) give an excellent match to the FSAR data. The steam generatorinventory (Figure 4.4-7) for the faulted loop depletes faster in the Dominion model thanin the FSAR case due to the higher steaming rate from the faulted steam generator andthe quicker and more conservative return to power.The nuclear power and core heat flux responses (Figure 4.4-9 and Figure 4.4-10)calculated by the Dominion model peak higher and more quickly than the FSAR data.This response is contributed to by the greater cooling effects of the faulted steamgenerator on the RCS due to its higher steam production. The quicker return to power isalso a result of differences in the nodalization and mixing at the core inlet and outletbetween the Dominion model and the FSAR model. The return to power also drops offapproximately 50 seconds sooner in the Dominion model. This is also caused by thehigher steam rate in the Dominion model which causes the faulted steam generator to dryout sooner. The power response for both models is not affected by the delivery of boronto the RCS. This is because the FSAR model does not credit boron and in the Dominionmodel boron does not reach the RCS from the SI system until after the termination of thetransient. Overall, the Dominion model results in a more conservative response for coreheat flux and power.
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2Attahmet 2pg. 33 of61The pressurizer pressure response (Figure 4.4-12) agrees very well with the pressurepredicted by the FSAR model for the first 50 seconds of the transient, after -which theFSAR data falls approximately 100 psi lower than the pressure calculated by theDominion model. This difference is a result of using only a single upper head leakagepath in the Dominion model. The upper head leakage is taken from the three intact loopsand does not credit any flow from the lower temperature, faulted loop. This causes theupper head temperature to remain slightly higher than would actually be the case, whichallows a vapor bubble in the upper head to form sooner and become larger. This in turnprevents the RCS pressure from falling lower.The pressurizer drains at approximately the same rate for the Dominion model and FSARmodels (Figure 4.4-13). However, for the Dominion model the pressurizer begins torefill approximately 100 seconds sooner. The quicker refilling is a result of the higherand quicker return to power which causes the RCS temperature to rise sooner in theDominion model. This causes the RCS fluid inventory to expand which results in thepressurizer refilling sooner in the Dominion model than is seen from the FSAR model.Table 4.4-2 MSLB with Offsite Power ResultsTime (sec) From Start ofTransientEvent Dominion FSARSteam Line Ruptures 0 0Manual Reactor Trip 0 0Increase MFW to 100% of Nominal HFP00Value 0____0 _Initiate Maximum AFW to Faulted SteamGenerator 0 0Main Feedwater Isolation 7.5 8.2MSIVs Closed 12.5 13.5Pressurizer Empty 15.5 20.5Criticality Attained 33.5 28.0Safety Injection Flow Initiation 47.9 72.8Faulted Steam Generator Dries Out 298 -350
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2 p.3o6pg. 34 of 61Figure 4.4-1 MSLB -Faulted Loop Steam Flow30002500i-' 2000E-15005000100 200 300 400 500600Time (sec)Figure 4.4-2MSLB -Intact Loop Steam Flow24001900cic)1400ES900400-1000 10 20 30 40 50Time (sec)60 70 80 90 100
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2Figure 4.4-3 MSLB -Faulted Loop Steam Generator Pressurepg. 35 of 61120010008004002000 100200 300 400 500Time (sec)MSLB -Intact Loop Steam Generator Pressure600Figure 4.4-4110010501000950a900S850~-800750700650I ------- DominionI! 4 qmmm m Im m N4 4Lira m m ,Blf$ w m W W Immm m0100200300Time (sec)400500600
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2 p.3o6pg. 36 of 61Figure 4.4-5MSLB -Faulted Loop Total Feedwater Flow14001200-1000-C,)800-o2 600-C,)C,)S400-200OIIII -----DominionI-" -FSARI01020304050Time (sec)Figure 4.4-6MSLB -Intact Loop Total Feedwater Flow12001000-U)EUt)800-600-400-200-* --- Dominion-- -- FSAR001020304050Time (sec)
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2 p.3o6pg. 37 of 61Figure 4.4-7 MSLB -Faulted Loop SG Liquid Mass180000160000140000-120000100000v80000S6000040000200000163000162000161000160000~1590001580001550001540000 100 200 300 400 500Time (sec)600Figure 4.4-8MSLB -Intact Loop SG Liquid Mass0100 200 300 400 500Time (sec)600
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2 p.3o6pg. 38 of 61Figure 4.4-9 MSLB -Normalized Core Power0n~Z0.200.180.160.140.120.100.080.060.040.020.000 100 200 300 400 500Time (sec)600Figure 4.4-10 MSLB -Normalized Core Heat FluxU.).N('0Z0.200.180.160.140.120.100.080.060.040.020.00010O0 200 300 400 500Time (sec)600
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2Attahmet 2pg. 39 of61Figure 4.4-11 MSLB -Reactivity Feedback100 200 300 4000500600100-100-300o-500*_: -700n, -900-1100-1300-1500Time (sec)Figure 4.4-12 MSLB -Pressurizer Pressure2500230021001900-,1700.-1500S1300o.11009007005000100 200 300 400 500Time (sec)600
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2 p.4o6pg. 40 of 61Figure 4.4-13 MSLB -Pressurizer Liquid Volume800700600i;; 500E400,,200 i--Dominion100 I ~--, ,-FSAR0100 200 300 400 500 600Time (sec)Figure 4.4-14 MSLB -Faulted Loop Vessel Inlet Temperature550530£L 490470 __ _a)I-I430410 !--.-Dmno.. ./- -- FSAR390...0 100 200 300 400 500 600Time (sec)
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2pg41olpg. 41 of 61Figure 4.4-15 MSLB -Intact Loop Vessel Inlet Temperature560550540530_520.~510E 500I--490480470460010O0 200 300 400 500 600Time (sec)Summary -MSLBThis section presents a comparison of a RETRAN-3D Main Steam Line Break transientcalculation with the Millstone model using the Dominion RETRAN transient analysismethods (Reference 1) compared to the FSAR results. The Dominion MPS3 analysis ispresented for benchmark comparison, and does not replace the existing AOR. The keyobservations from these comparisons are that:1) The peak power and heat flux reached with the Dominion methods is higher thanthe FSAR result.2) Core and steam generator nodalization effects asymmetric transients such as aMSLB.
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2 p.4o6pg. 42 of 614.5 Control Rod Bank Withdrawal at PowerThe Control Rod Bank Withdrawal at Power (RWAP) event is defined as the inadvertentaddition of core reactivity caused by the withdrawal of rod control cluster assembly (RCCA)banks when the core is above no load conditions. The RCCA bank withdrawal results inpositive reactivity insertion, a subsequent increase in core nuclear power, and acorresponding rise in the core heat flux. The RWAP event described here is terminated bythe Reactor Protection System on a high neutron flux trip or the overtemperature AT trip(OTAT), consistent with the FSAR analyses.The RWAP event is simulated by modeling a constant rate of reactivity insertion startingat time zero and continuing until a reactor trip occurs. The Dominion analysis involvestwo different reactivity insertion rates, 1 pcm/sec and 100 pcm/sec that match thereactivity insertion rates presented plots in the FSAR. Most of the input parameters are thesame as those used in the FSAR Chapter 15 analyses. Where differences from the FSARinputs exist, they are indicated in the Notes column.Table 4.5-1 RWAP In ut SummaryParameter Value NotesInitial Conditions________________Core Power (MW) 3650 NominalRCS Flow (gpm) 379,200 Minimum Measured FlowVessel TAVG (F) 589.5 NominalRCS Pressure (psia) 2250 NominalPressurizer Level (%) 64 NominalSG Level (%) 50 NominalInitial Fuel Temperature Minimum Uses current FSAR analysisconductivity adjustmentsAssumptions/ConfigurationReactor trip -High neutron flux or OTATAutomatic rod control -Not creditedPressurizer level control -Not creditedPressurizer heaters -Not creditedPressurizer sprays, PORVs -ActiveSG tube plugging (%) 10 Max valueReactivity Parameters ______Doppler Reactivity Feedback Least Negative _________________Moderator Feedback Most Positive Zero MTC for cases from full powerResults -RWAP 1 pcm/sec CaseFigure 4.5-1 shows the core power response. The core power rate of increase for theDominion model is greater than the FSAR data. This leads to the Dominion modeling
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2 p.4o6pg. 43 of 61tripping on high neutron flux at about 74 seconds. The FSAR case rises in power at aslower rate, which trips on an OTAT signal at about 93 seconds. The difference in reactortrip mechanisms between the Dominion and FSAR cases is reasonable considering thebreakpoint for switching between OTAT and high flux as shown in FSAR Figure 15.4-10.The pressure response also affects the OTAT setpoint such that the lower FSAR pressure(see below) will act to reduce the setpoint.The pressurizer pressure response is shown in Figure 4.5-2. For the Dominion model, thepressure rises faster than the FSAR result. At about 42 seconds, the Dominion modelreaches the pressurizer relief valve setpoint and begins to cycle. The FSAR more slowlyincreases in pressure and reaches the relief valve set point around 10 seconds prior to thereactor trip. The difference in pressure response can be attributed to the difference in corepower response as each cases pressure response initially mimics the energy generated bythe core as seen in Figure 4.5-1 and the higher spray flow assumed in the FSAR analysis,which acts to suppress pressure. The same can be seen in the vessel average temperatureresponse where the FSAR case lags the Dominion response, yet reaches a temperatureapproximately 5 degrees higher than the Dominion case due to the FSAR case trippinglater in the transient.Table 4.5-2 RWAP 1 pcm/sec Time Sequence of EventsEvent Tm scnsReactivity Insertion at 1 pcm/sec 0.00.IRatrTrip Signal Initiated 7."9."* Trip on high neutron flux** Trip on OTATResults -RWAP 100 pcm/sec CaseFigure 4.5-4 shows the core power response for the current FSAR analysis and theDominion model. The Dominion model trips on a high neutron flux at about 1.17 seconds,compared to about 1.29 seconds for the current FSAR analysis. The 100 pcm/sec transient isa fast transient and the time period before the reactor trip is so brief that any differences infuel pin heat transfer modeling assumptions have little impact on Doppler reactivityfeedback. Overall, the Dominion model peaks at a higher, thus more conservative powerlevel.
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2 p.4o6pg. 44 of 61The pressurizer pressure response is shown in Figure 4.2-5. The Dominion model matchesvery well with the FSAR analysis. The main difference being that the Dominion modelpeaks at a higher pressure than the FSAR analysis. This correlates with the power responseshown in Figure 4.2-4 where the Dominion model peaks at a higher overall nuclear power.Figure 4.2-6 shows the vessel average temperature. For the 100 pcm/sec case the Dominionmodel matchs very closely with the FSAR analysisTable 4.5-3 RWAP 100 peru/see Time Sequence of EventsEvent Time (seconds)__________________ Dominion FSARReactivity Insertion at 100 pcrn/sec 0.0 0.0Reactor Trip Signal Initiated 1.17* 1.29**Trip on high neutron flux
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2 p.4o6pg. 45 of 61Figure 4.5-1 RWAP -1 pcrn/sec Nuclear Power1.401.201.00~50.80.' 0.60oz 0.400.200.002400235023002250220021502100205020001950--~IIIIIIIIII-" -- FSAR---- DominiontI iIi ii0 20 40 60 80 100 120 140Time (sec)Figure 4.5-2 RWAP -1 pcrn/sec Pressurizer PressureVIII 'I--- Dominion020 4060Time (sec)80100120
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2 p.4o6pg. 46 of 61Figure 4.5-3 RWAP -1 pcm/sec Vessel Average TemperatureU-0Q)I-U)0.E0)I-610605600595590585580575570565020 40 60 80 100Time (sec)Figure 4.5-4 RWAP -100 pcm/sec Nuclear Power1201.401.201.00S0.80o4)N~ 0.60E0z 0.400.200.0002 4 6 8 10Time (sec)
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2Attahmet 2pg. 47 of 61Figure 4.5-5 RWAP -100 pem/sec Pressurizer Pressure240023502300&2200~2150210020502000195002 4 6 8 10Time (sec)Figure 4.5-6 RWAP -100 pcm/sec Vessel Average Temperature600595590585.~580S5755705655600 2 4 6 8 10Time (sec)
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2 pg. 48 of 61Summary -RWAPThe Dominion Millstone model provides results that are similar to the FSAR analysis for theRWAP event. At higher insertion rates, the results match very well. At lower insertionrates, the power increases at a greater rate in the Dominion model than the FSAR model.However, the temperature increases to a higher peak in the FSAR analysis. The DominionMPS3 analysis is presented for benchmark comparison, and does not replace the existingAOR.
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2 p.4o6pg. 49 of 614.6 Main Feedwater Line BreakThe Main Feedwater Line Break (MFLB) event is defined as a break in a feedwater linelarge enough to prevent the addition of sufficient feedwater to the steam generators tomaintain shell side fluid inventory in the steam generators. If the break is postulated in afeedline between the check valve and the steam generator, fluid from the steam generatormay also be discharged through the break. Depending upon the size of the break and theplant operating conditions at the time of the break, the break could cause either a RCScooldown (by excessive energy discharge through the break) or a RCS heatup. The FSARanalysis presents the RCS heatup scenario.A major feedwater line rupture is classified as an ANS Condition IV event as discussedin FSAR Section 15.0.1. A main feedwater line rupture is the most limiting event in thedecrease in secondary heat removal category. Based on a number of prior analyses, it isconcluded in FSAR Section 15.2.8 that the most limiting feedwater line rupture is adouble ended rupture of the largest feedwater line, occurring at full power with andwithout offsite power available. Cases both with and without offsite power available aresimulated for the benchmark analysis herein.The MFLB transient is initiated in the Dominion model by opening the break on steamgenerator 1 and stopping, main feedwater to all four steam generators (SG) as the reactoris operating at full power. Upon transient initation, the break path opens and allowsblowdown from the faulted SG secondary side inventory to the atmosphere. The inputparameters are the same as those used in the FSAR Chapter 15 analyses as shown in Table4.6-1 below.The results for the MFLB transient need to demonstrate that the reactor core remainscovered, the RCS does not overpressurize, and the AFW system is able to adequatelyremove decay heat.
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2 p.5o6pg. 50 of 61Table 4.6-1 M4FLB Input SummaryParameter Value NotesInitial ConditionsCore Power (MW) 3723, Includes 2% uncertaintyRCS Flow (gpm) 363,200 Thermal Design FlowVessel TAvG (F) 594.5 Nominal + 5 0FRCS Pressure (psia) 2300 Nominal + 50 psiPressurizer Level (%) 71.6 Nominal + 7.6%SG evl %)62 Nominal + 12% (Faulted Loop)-SG evl %)38 Nominal -12% (Intact Loops)SG tube plugging (%) 10 MaximumPump Power (MW/pump) 5.0 MaximumAssumptions/ConfigurationLow-Low Level Reactor Trip Setpoint 0% % narrow range span in faulted SGPressurizer: sprays, heaters, PORVs -Not creditedAFW Temperature (F) 120 Max valueAuxiliary feedwater flow rate (gpm) .Variable as function of SQ press.All MFW assumed lost at time ofMain Feedwater 0 braReactivity ParametersDoppler Reactivity Feedback Most CnevtvasmtoModerator Feedback Negative CosraiesumtnResults -MFLB Case With Offsite Power AvailableThe results for the MFLB case with offsite power available are presented on Figure 4.6-1through Figure 4.6-8. The nuclear power response (Figure 4.6-1) predicted by theDominion model is in good agreement with the FSAR data, with the reactor tripoccurring on low-low steam generator level. There is a return to power betweenapproximately 100-200 seconds due primarily to moderator reactivity feedback effectsduring the primary side cooldown prior to steam line isolation (SLI). After that time, thecore remains subcritical for the duration of the transient.The response for pressurizer pressure and pressurizer water volume are shown on Figure4.6-2 and Figure 4.6-3. The Dominion results trend well with the FSAR results forpressurizer pressure and water volume. One difference is a brief increase, in pressurizerpressure and associated insurge into the pressurizer around the point of reactor trip for theDominion case. ,This increase occurs due to differences in the primary-to-secondary heattransfer following the reactor and turbine trips between the MNSG FSAR model and theDominion SNSG. The SNSG responds more quickly to the decrease in secondary Sidelevel following the loss of main feedwater compared to the MNSG, which initially
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2 p.5o6pg. 51 of 61experiences less reduction in SG level and associated heat transfer. This effect onlyoccurs for a relatively brief duration. Eventually, steam line isolation (SLI) occurs on lowsteam line pressure resulting in a primary side heatup as the intact SGs repressurize.Pressurizer pressure increases until the pressurizer safety valve (PSV) setpoint is reachedand remains essentially constant at the PSV relief pressure until a downturn in pressureoccurs near the end of the transient. This indicates the termination of the event assufficient cooling is being provided by auxiliary feedwater (AFW) for the removal ofprimary side energy.The hot leg and cold leg temperature response is shown on Figure 4.6-4 for the faultedloop and on Figure 4.6-5 for the intact loops. There is good agreement between theDominion and FSAR cases with temperatures exhibiting the same trends throughout the.event and deviating only slightly prior to SLI, which has a negligible effect on the overallresults for this comparison due to the long term nature of this event. As noted for thepressure response discussion above, the temperatures are decreasing at the end of thetransient indicating adequate long term heat removal.The Dominion RCS flow fraction results are shown on Figure 4.6-6. Since power to thereactor pumps is not lost for this case, flow is maintained throughout the transient andvaries only with coolant conditions. The Dominion case is in good agreement with theFSAR data throughout the transient.The secondary system pressure response is presented on Figure 4.6-7 where SG pressureincreases briefly following the reactor trip then decreases due to the loss of fluid massthrough the feed line break. After SLI occurs, the intact SG pressure increases to theMSSV setpoint while the faulted SG pressure continues to decrease to atmosphericpressure as the remaining fluid mass is depleted. The Dominion and FSAR cases showgood agreement as both the magnitude and trends of faulted and intact loops areconsistent following the point of reactor trip and subsequent SLI.Figure 4.6-8 shows excellent agreement between the main feedwater break flow rateresponse in both the Dominion and FSAR case. One difference is seen around the pointof reactor trip over a period of approximately 12 seconds that is related to the steamgenerator modeling differences. As discussed relative to the pressurizer pressureresponse, the Dominion SNSG model results in a faster reduction in liquid level and morerapid increase in break flow quality such that flow falls off more quickly as the break isuncovering. After this brief transition period the break flow rates continue to agree welland this difference has a negligible effect on the overall transient response.
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2pg. 52 of 61Figure 4.6-1 MFLB -Nuclear Power (case with power)1.2wo0.6n- 0.40.2110 100 1000Time (sec)10000Figure 4.6-2 MFLB -Pressurizer Pressure (case with power)260024002200200018001600140012001000100001 ~~~10 10 00100Ta me (sec)1000
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2 p.5o6pg. 53 of 61Figure 4.6-3 MFLB -Pressurizer Liquid Volume (case with power)200016001400.o 12001001000600400200180 i ~ ~ iiIi i i" i 00NW Iomr°' li...... L1010 100 1000lime (sec)10oo10000Figure 4.6-4 MFLB -RCS Temperatures -Faulted Loop (case with power)650E 5505004501010 100 1000-time (sec)100010000
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2pg. 54 of 61Figure 4.6-5 MFLB -RCS Temperatures -Intact Loops (case with power)700650a,E5505004501010O0 1000 10000Tlime (sec)Figure 4.6-6 MFLB -Normalized RCS Flow (case with power)1.15 ____l ii i i I i1.1 i i iCo 1CS0.90.80.80.750.710100"time (sec)10 000 10000
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2pg. 55 of61Figure 4.6-7 MFLB -Steam Generator Pressure (case with power)140012001000* 80060040020010 100 1000lime (sec)10000Figure 4.6-8 MFLB -Feed Line Break Flow (case with power)8000 " )7000 .... "= T6000 /.G5000 .... .....20001000 ---lime (sec)100001000 qI ~~~Attachment 2 p.5o6pg. 56 of 61Results -MFLB Case Without Offsite Power AvailableThe results for the MFLB case without offsite power are similar to the case with poweravailable but are generally less limiting for long-term primary side heat removal since theRCPs are not running and adding heat to the primary side fluid.The nuclear power response (Figure 4.6-9) predicted by the Dominion case is in goodagreement with the FSAR data. As shown for this case, there is no return to powerduring the early portion of the cooldown due to less reactivity feedback and the reactorcore remains subcritical for the duration of the transient.The responses for pressurizer pressure and primary side temperatures are shown onFigures 4.6-10 through 4.6-12. As discussed above for the case with offsite power, theDominion case exhibits a brief increase in pressure around the time of reactor trip butotherwise the response is similar to the FSAR case with long-term pressure maintained atthe PSV setpoint. The hot leg and cold leg temperature response shown on Figure 4.6-11and Figure 4.6-12 also demonstrate similar trends. One difference is that the cooldownthat occurs prior to SLI is more pronounced for the Dominion case, which is primarilyattributed to higher primary to secondary heat transfer. This is the result of a somewhatslower rate of flow decrease following the RCP trip for the Dominion case, resulting inmaintaining better primary side heat removal during that phase. In addition, SLI occursslightly later in the Dominion case, which also enhances heat removal prior to the time ofisolation. Similarly, the delay in break isolation delays the point of steam generator dry-out, such that additional heat is extracted through the break. As shown, these differenceshave little effect on the long-term temperature response as the Dominion and FSARtemperatures agree very well through the end of the transient. This case results in lowerlong-term temperatures, as the RCPs trip due to the loss of offsite power and do notcontribute any pump heat to the system.The secondary system pressure response, presented in Figure 4.6-13, is similar to theresponse for the case with power. Since there is less primary side heat generation andheat removal for this case, the SG depressurizes more quickly and SLI occurs earlier inthe transient, compared to the case with offsite power available. Long term trends aresimilar with heat removal via the MSSVs on the intact SGs. There is good agreementbetween the Dominion and FSAR cases with the FSAR case depressurizing slightly fasterprior to SLI.The Dominion RCS flow fraction results are in good agreement with the FSAR result asshown on Figure 4.6-14, where the loss of flow associated with the loss of power andassociated RCP trip are seen. As noted above, the flow decreases somewhat more
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2 p.5o6pg. 57 of 61quickly for the FSAR case, which appears to affect the intermediate temperatures butdoes not impact the long term temperature results.Figure 4.6-15 shows good agreement between the main feedwater break flow rateresponse in both the Dominion and FSAR data. The small differences seen around thepoint of reactor trip are due to differences in the Dominion SNSG and the FSAR MNSGas discussed above for the case with power available. That is, the Dominion SNSG modelresults in a faster reduction in liquid level and more rapid increase in break flow qualitysuch that flow falls off more quickly as the break is uncovering. After this brieftransition period the break flow rates continue to agree well and this difference has anegligible effect on the overall transient responseFigure 4.6-9 MFLB -Nuclear Power (case without power)I i i 1! I I t l iE o nihIgO. ii i i iiU04 ~0.2 _ _ i110 100 1000 10000"Time (sec)
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2pg. 58 of 61Figure 4.6-10 MFLB -Pressurizer Pressure (case without power)250024002300220000200019001800170016001500110 100 1000 10000Trme (sec)Figure 4.6-11 MFLB -RCS Temperatures -Faulted Loop (case without power)7006506005O045O1 10 100 1000-time (sec)10000 SIAttachment 2 pg. 59 of 61Figure 4.6-12 MFLB -RCS Temperatures -Intact Loops (case without power)700650u_600S50,45O110 100 1000100Tlime (see)10000Figure 4.6-13 MFLB -Steam Generator Pressure (case without power)1200 t }-,00 __ i I*200 ~ Dii ! M ip j ll " ! -- edoPi I 11i i 10 1 Do Ill tei ap 10 -*~~~10101000 10000"lime (sec)
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2pg. 60 of 61Figure 4.6-14 MFLB -Normalized RCS Flow (case without power)1.2 i i " -' -1 !i FSAR0u.8 --006 I -°0.: i 1 } !a ___ --I10 10O0 1000 10000Figure 4.6-15 MFLB -Feed Line Break Flow (case without power)8000 -_______7000 li ii'i° i i 0o n o o5000O ,-.-t- ii -fi i I i ii -Time (sec)
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2pg. 61 of 61Summary -MFLBThe Dominion Millstone model provides results that are similar to the FSAR analysis forthe MFLB event. Two cases are analyzed, one with offsite power available and anotherwithout offsite power. Some small differences are observed early in the transient forRCS temperatures, which are attributable to differences in the Dominion SNSG modeland the FSAR MINSG model; however, these differences have a negligible effect on thelong-term primary side heat removal and associated temperature response. All acceptancecriteria are satisfied for both cases.5.0 ConclusionsThis attachment presents benchmarking transient analyses performed with the MPS3RETRAN model developed in accordance with VEP-FRD-4 1-P-A. These analysis resultsare compared with current Millstone FSAR results. The following conclusions are drawnbased on these analyses.1) It is demonstrated that the Dominion RETRAIN-3D model and analysis methods canpredict the response of transient events with results that compare well to FSARresults.2) Where there are differences between the Dominion results and the FSAR results, theyare understood based on differences in noding, inputs, or other modeling assumptions.3) The Dominion Millstone RETRAN-3D model is consistent with current Dominionmethods (Reference 1). These methods have been applied extensively for Surry andNorth Anna licensing, engineering and plant support analyses.4) The RETRAN comparison analyses satisfy the applicability assessment criteria andprovide further validation of the conclusion that Dominion's RETRAN analysismethods are applicable to Millstone and can be applied to Millstone licensing analysisfor reload core design and safety analysis.6.0 References1) Topical Report, VEP-FRD-41-P-A, Rev. 0.2, "VEPCO Reactor System TransientAnalyses Using the RETRAN Computer Code," March 2015.2) Topical Report, VEP-NFE-2-A, "VEPCO Evaluation of the Control Rod EjectionTransient," December 1984.