ML15239B290

Oconee, Units 1, 2, and 3 - License Amendment Request (LAR) to Add High Flux Trip for 3 Reactor Coolant Pump Operation, License Amendment Request No. 2014-05, Supplement 1

ML15239B290
Person / Time
Site:	Oconee
Issue date:	08/20/2015
From:	Batson S L Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
0NS-2015-096, 2014-05, Supplement 1
Download: ML15239B290 (9)

Text

tDUKE SttL. BasenENERGY Vice PresidentOconee Nuclear StationDuke EnergyONO1VP 7800 Rochester HwySeneca, SC 296720; 864.873.3274864.873.42080NS-201 5-096 10 C FR 50.90 Scott.Batson@duke-energy.comAugust 20, 2015ATTN: Document Control DeskU.S. Nuclear Regulatory Commission11555 Rockville PikeRockville, MD 20852Duke Energy Carolinas, LLC (Duke Energy)Oconee Nuclear Station (ONS), Units 1, 2, and 3Docket Numbers 50-269, 50-270, and 50-287Renewed License Numbers DPR-38, DPR-47, and DPR-55

Subject:

License Amendment Request (LAR) to Add High Flux Trip for 3 Reactor CoolantPump OperationLicense Amendment Request No. 2014-05, Supplement 1On May 19, 2015, Duke Energy submitted a License Amendment Request (LAR) proposing toadd a Reactor Protective System (RPS) Nuclear Overpower -High Setpoint trip for three (3)reactor coolant pump (RCP) operation to Technical Specification Table 3.3.1-1. By letter datedAugust 6, 2015, the Nuclear Regulatory Commission (NRC) requested Duke Energy submitsupplemental information to enable the NRC Staff to complete the acceptance review for theLAR.The enclosure provides the supplemental information. If there are any additional questions,please contact Boyd Shingleton, ONS Regulatory Affairs, at (864) 873-4716.I declare under penalty of perjury that the foregoing is true and correct. Executed onAugust 20, 2015.Sincerely,Scott L. BatsonVice PresidentOconee Nuclear Station

Enclosure:

Duke Energy Response to Acceptance Review Information Request A jwww.duke-energy.com U. S. Nuclear Regulatory CommissionAugust 20, 2015Page 2cc w/enclosure:Mr. Victor McCreeAdministrator Region IIU.S. Nuclear Regulatory CommissionMarquis One Tower245 Peachtree Center Ave., NE, Suite 1200Atlanta, GA 30303-1257Mr. James R. HallSenior Project Manager(by electronic mail only)Office of Nuclear Reactor RegulationU.S. Nuclear Regulatory Commission11555 Rockville Pike-Mail Stop O-8G9ARockville, MD 20852Mr. Jeffrey A. WhitedProject Manager(by electronic mail only)Office of Nuclear Reactor RegulationU.S. Nuclear Regulatory Commission11555 Rockville PikeMail Stop O-8B1ARockville, MD 20852Mr. Eddy CroweNRC Senior Resident InspectorOconee Nuclear StationMs. Susan E. Jenkins, Manager, Infectious and Radioactive Waste ManagementBureau of Land and Waste ManagementDepartment of Health & Environmental Control2600 Bull StreetColumbia, SC 29201 ENCLOSUREDuke Energy Response to Acceptance Review Information Request License Amendment Request No. 20 14-05, Supplement 1 Page 1 of 6August 20, 2015EnclosureDuke Energy Response to Acceptance Review Information RequestNRC Information Request 1Provide a more in-depth discussion on which regulatory criteria are applicable to the LAR. TheLAR cited 10 CFR 50.36 as its regulatory basis. 10 CFR 50.36 states that limiting conditions foroperation (LCO's) must be established for items meeting one of the four criteria cited in theregulation. Specifically, provide which of the 50.36 criteria are applicable to the proposed newsetpoint, and a discussion of whether the existing TS requirements are sufficient to ensureoperation within the bounds of the accident analysis.Duke Energy ResponseThe NRC's regulatory requirements related to the content of the Technical Specification (TS)are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36. Paragraph(c)(2)(i) of 10 CFR 50.36 states that Limiting Conditions for Operation (LCOs) are the lowestfunctional capability or performance levels of equipment required for safe operation of thefacility. Paragraph (c)(2)(ii) of 10 CFR 50.36 lists four criteria for determining whether particularitems are required to be included in the TS LCOs. The third criterion applies to a structure,system, or component that is part of the primary success path and which functions or actuatesto mitigate a design basis accident or transient that either assumes the failure of or presents achallenge to the integrity of a fission product barrier. The Nuclear Overpower -High Setpointtrip function meets Criterion 3. The proposed change adds an additional trip setpoint for threereactor coolant pump (RCP) operation. A new trip function is not added.In MODES 1 and 2, the Nuclear Overpower -High Setpoint trip, along with other ReactorProtective System (RPS) trips, are required to be OPERABLE because the reactor can becritical in these MODES. These trips are designed to take the reactor subcritical to maintain theTS Safety Limits during anticipated transients and to assist the ESPS in providing acceptableconsequences during accidents. While the existing overpower protection for three reactorcoolant pump (RCP) operation (provided by the Nuclear Overpower Flux/Flow/Imbalance tripfunction) is adequate, the proposed Nuclear Overpower flux trip setpoint for three RCPoperation provides improved protection for power excursion events initiated from three RCPoperation, most notably the small steam line break accident. The Nuclear Overpower flux tripprovides an absolute setpoint that can be actuated regardless of transient or Reactor CoolantSystem (RCS) flow conditions. The faster response time provides additional departure fromnucleate boiling (DNB) and RCS protection than provided by the slower acting nuclearoverpower flux/flow/imbalance trip function. The proposed high flux trip setpoint will result insignificant margin improvement to the departure from nucleate boiling ratio (DNBR) acceptancecriterion.NRC Information Request 2Provide the regulatory basis for the new reactor trip. Please describe which regulations the newreactor trip is intended to comply with (e.g., 10 CFR Part 100, GDC 10 or alternative criteria thatestablish the Oconee licensing basis).

License Amendment Request No. 2014-05, Supplement 1 Page 2 of 6Enclosure -Duke Energy Response to Acceptance Review IssuesAugust 20, 2015Duke Energy ResponseThe regulation of General Design Criteria (GDC) 10 of Appendix A to 10 CFR Part 50, "Reactordesign," requires that the reactor core and associated coolant, control, and protection systemsbe designed with appropriate margin to assure that specified acceptable fuel design limits arenot exceeded during any condition of normal operation, including the effects of anticipatedoperational occurrences.The regulation of GDC 20 of Appendix A to 10 CFR Part 50, "Protection System Functions,"requires protection system functions to be designed to 1) to initiate automatically the operationof appropriate systems including the reactivity control systems, to assure that specifiedacceptable fuel design limits are not exceeded as a result of anticipated operationaloccurrences and (2) to sense accident conditions and to initiate the operation of systems andcomponents important to safety.The principal design criteria (POC) for ONS were developed in consideration of the seventyGeneral Design Criteria for Nuclear Power Plant Construction Permits proposed by the AtomicEnergy Commission (AEC) in a proposed rule-making published for 10CFR Part 50 in theFederal Register on July 11, 1967. The ONS, Units 1, 2, and 3, construction permits wereissued on November 6, 1967, preceding the issuance of the GDC specified in 10 CFR 50Appendix A. The proposed trip setpoint is intended to comply with PDC 6 and 14, which arecomparable to GDC 10 and 20, respectively.PDC 6 specifies that the reactor core shall be designed to function throughout its design lifetimewithout exceeding acceptable fuel damage limits which have been stipulated and justified. Thecore design, together with reliable process and decay heat removal systems, shall provide forthis capability under all expected conditions of normal operation with appropriate margins foruncertainties and for transient situations which can be anticipated, including the effects of theloss of power to recirculation pumps, tripping out of a turbine generator set, isolation of thereactor from its primary heat sink, and loss of all off-site power. ONS Updated Final SafetyAnalysis Report (UFSAR) Section 3.1.6 states that the reactor is designed with the necessarymargins to accommodate, without fuel damage, expected transients from steady-state operationincluding the transients given in the criterion. The design margins allow for deviations oftemperature, pressure, flow, reactor power, and reactor turbine power mismatch. Above 15percent power, the reactor is operated at a constant average coolant temperature and has anegative power coefficient to damp the effects of power transients. The Reactor Control Systemwill maintain the reactor operating parameters within preset limits, and the Reactor ProtectiveSystem will shut down the reactor if normal operating limits are exceeded by preset amountsPDC 14 specifies that core protective systems, together with associated equipment, shall bedesigned to act automatically to prevent or to suppress conditions that could result in exceedingacceptable fuel damage limits. ONS UFSAR Section 3.1.14 states that the ONS reactor designmeets this criterion by reactor trip provisions and engineered safety features. The ONS ReactorProtective System is designed to limit reactor power which might result from unexpectedreactivity changes, and provides an automatic reactor trip to prevent exceeding acceptable fueldamage limits.

License Amendment Request No. 20 14-05, Supplement 1 Page 3 of 6Enclosure -Duke Energy Response to Acceptance Review IssuesAugust 20, 2015NRC Information Request 3Provide a description of the accident analysis that demonstrates the 80.5% reactor trip setpointis adequate to meet the applicable AAO*/Accident acceptance criteria. The description shouldbe at a level consistent with the description of accidents in the FSAR and include the analysiscodes and methods, key analysis assumptions as well as the applicable acceptance criteria.*AAO should be AOO (anticipated operational occurrence) per telecon with Randy Hall, ONSNRR Project Manager on August 11, 2015.Duke Energy ResponseThe main accident for which credit will be taken for the proposed three RCP High Flux Tripsetpoint is the UFSAR Chapter 15.17 Small Steam Line Break (SSLB) transient initiated fromthree RCP operation. Other accidents initiated from three RCP operation could credit theproposed trip function, but the motivation for the proposal is the SSLB transient analysis. Theaccident starts at the maximum power level allowed when operating with three RCPs and isanalyzed in such a manner as to maximize the primary system overcooling and subsequentpower increase while avoiding and/or delaying a valid RPS trip signal. The existing three RCPSSLB credits two flux related RPS trip functions. The first is the existing TS High Flux tripfunction and setpoint, which is set at 105.5% Rated Thermal Power (RTP). The second is theFlux/Flow/Imbalance trip function, which is a dynamic setpoint based on the measured powerand measured RCS flow rate. Since SSLB is an overcooling event, as the RCS gets colder thecoolant becomes more dense, the measured RCS flow rate increases. This overcooling causestwo responses to the Oconee RPS. First, the colder reactor vessel downcomer fluid attenuatesneutrons and masks the excore detectors from measuring the true core power. This causes thetrue power to potentially increase much higher than what the excore detectors would indicateand hence, delay and or avoid either a high flux trip or a flux/flow/imbalance trip, both of whichuse indicated (i.e., excore detectors) core power as an input. The SSLB event in UFSARChapter 15.17 conservatively models downcomer attenuation and its impact on the excoredetector signal. The second response is that the colder, more dense coolant, causes measuredRCS flow to increase which causes the flux/flow/imbalance trip to increase. This also delaysand/or avoids reactor trip on flux/flow/imbalance. This is a physical phenomenon and no specialmodeling techniques are required to account for this effect. With four RCPs operating, the highflux trip is more effective at tripping the reactor than the flux/flow/imbalance trip, even before theflux/flow/imbalance trip setpoint increases due to increasing flow. With three RCPs operating,the flux/flow/imbalance trip is the main trip function and, with the dynamic setpoint increasing asthe coolant becomes more dense, a much larger increase in true core power occurs relative tothe power increase calculated with four RCPs in operation. Therefore, the current limiting SSLBaccident documented in UFSAR 15.17 for the DNBR acceptance criterion is the SSLB initiatedfrom three RCP operation.The NRC-approved analysis (documented in DPC-NE-3005-PA) of the SSILB transient uses theNRC approved code RETRAN-3D to determine a limiting combination of steam line break sizeand moderator temperature coefficient (MTC) to produce the largest true core power excursionthereby challenging DNBR and centerline fuel melt (CFM), both of which are acceptance criteria License Amendment Request No. 2014-05, Supplement 1 Page 4 of 6Enclosure -Duke Energy Response to Acceptance Review IssuesAugust 20, 2015for the UFSAR Chapter 15.17 event. A larger break size increases the primary systemovercooling while a more negative moderator temperature coefficient (MTC) maximizes thepower excursion as a result of that overcooling. Too large a break will result either in a low RCSpressure or variable low pressure-temperature reactor trip or a faster power increase resulting ina high flux or flux/flow/imbalance trip. Too negative an MTC will result in a faster powerincrease resulting in a high flux or flux/flow/imbalance trip. The most conservative combinationof break size and MTC either avoids a RPS trip altogether or delays it long enough to maximizethe true core power.Centerline fuel melt is only a concern for 4 RCP operation and will not be addressed further inthis LAR. Departure from nucleate boiling ratio calculations are performed with the NRCapproved VIPRE-01 code using the RETRAN-3D forcing functions as input. Departure fromnucleate boiling ratio is more limiting for three RCP operation (vs. four RCP operation) due tothe combination of lower RCS flow and higher relative power increase. The type of steam linebreaks analyzed in UFSAR 15.17 can only occur if there were an actual pipe break (vs. valvefailures), which is classified as an infrequent fault and therefore, DNB fuel failures are allowed.However, Duke Energy treats the SSLB as a fault of moderate frequency with respect to theDNB acceptance criteria and consequently, no DNB related fuel failures are allowed. Theanalysis of the three RCP SSLB with the proposed high flux trip setpoint for when three RCPsare operating demonstrates that true core power is significantly reduced before reactor tripoccurs. In fact, with the proposed high flux trip setpoint for three RCP operation, the limitingSSLB transient with respect to both CFM and DNB becomes the SSLB initiated from four RCPs.NRC Information Request 4Provide a sample calculation that shows the uncertainty determination in the elements of thesetpoint calculations for the high flux trip.Duke Energy ResponseAs stated in the [AR, the 80.5% RTP setpoint was chosen to maintain the delta betweennominal 100% RTP and the current TS allowable value of 105.5% RTP. The 5.5% RTP delta issimply added to the maximum power level allowed for three RCP operation, which is 75% RTP.Adding 5.5% RTP to 75% RTP results in the proposed high flux trip setpoint of 80.5%RTP. Thisvalue is verified acceptable in the SSLB analysis initiated from three RCP operation.The method described in the NRC-approved DPC-NE-3005-PA (Chapter 4), for performingChapter 15 analyses specifies that the trip setpoint assumed in the analyses is the TS tripsetpoint plus (or minus) an uncertainty to account for the trip setpoint uncertainty itself. Anyuncertainty or adjustments in the signal that is used to compare to the setpoint is accounted forin the specific analysis, if applicable. For the SSLB DNB analyses, the Statistical Core Design(SCD) method is employed (NRC approved DPC-NE-2005-PA) which accounts for the variousuncertainties in core power and RCS flow in the DNB limit itself. What is not accounted for inthe SCD method is transient effects such as downcomer attenuation, which the SSLBRETRAN-3D analysis specifically accounts for as described in DPC-NE-3005-PA. Asmentioned previously, reactor vessel downcomer attenuation affects the excore detector signal License Amendment Request No. 2014-05, Supplement 1 Page 5 of 6Enclosure -DLdke Energy Response to Acceptance Review IssuesAugust 20, 2015response and acts to mask the true power increase. Basically, if this were put in mathematicalterms, it would be:q~r > s + trip setpoint uncertainty allowanceWhere era = flux measured at excore detectors adjusted for transient effects (e.g.,downcomer attenuation) and excore detector calibration tolerances=~p Technical Specification allowable value trip setpointTrip setpoint uncertainty = current analysis assumes 1.0% RTP for convenience sincethat is the old analog RPS trip bistable uncertainty and it bounds the uncertainty on thesetpoint in the digital RPS. There is no uncertainty on the trip setpoint in the digital RPS.NRC Information Request 5Identify and describe the procedure that will be used by control room operators to manuallyinsert the high flux trip setpoint when going from 4 RCP operation to 3 RCP operation. Pleasealso describe how this procedure accomplishes the setpoint changes to avoid overpoweroperation or spurious trips.Duke Energy ResponseIf a condition arises which requires Operations to reduce reactor power on an operating unit sothat a reactor coolant pump can be shutdown, Operations procedural guidance(OP/I1,2,3/A/11102/004 -Operation at Power) triggers a notification to maintenance personnel tochange the RPS high flux trip set point from the four RCP value to the three RCP value. This isdone following power reduction and shutdown of the problematic RCP. A maintenanceprocedure (AM/I1,2,3/A/031 5/017 -TXS RPS Channels A, B, C, and D Parameter Changes ForAbnormal/Normal Operating Conditions) is utilized to perform the following action one RPSchannel at a time. The RPS is a digital system. From the RPS service unit, a graphical servicemonitor screen which has design features specific to changing the high flux trip set point is usedto lower the high flux set point to the required three RCP value.When conditions permit returning to four RCP operation, the fourth RCP is placed in service, thehigh flux trip set point for each RPS channel is changed to the four RCP value via Operationsnotification to Maintenance who use the same Maintenance procedure to change the set point,and then escalation to full power operation is allowed.

License Amendment Request No. 2014-05, Supplement 1 Page 6 of 6Enclosure -Duke Energy Response to Acceptance Review IssuesAugust 20, 2015NRC Information Request 6Provide an explanation of how the 80.5% RTP high flux trip setpoint will be verified to beapplicable to each new reactor core loading.Duke Energy ResponseMaximum allowed peaking limit curves are generated with the VIPRE-01 computer code for theSSLB transient, and will continue to be generated for the three RCP SSLB transient once theproposed high flux trip setpoint is approved and implemented. These peaking limit curves areperformed once for the bounding analysis then verified acceptable for each reload core. Thepeaking limit curves for SSLB restrict peaking to preclude the occurrence of DNB. Duke Energyverifies that the DNB acceptance criteria is met for each reload by comparing potential pinpowers from the SIMULATE neutronics code to the peaking limit curves generated by VIPRE.Duke Energy intends to continue this practice for the SSLB transient initiated from three RCPoperation unless and until it is demonstrated that the four RCP SSLB transient is more limitingand bounding than the three RCP SSLB transient. If such a situation were to occur, DukeEnergy would perform reload checks to the four RCP SSLB peaking limit curves and only verifythe three RCP SSLB peaking limits remain bounded for any future fuel design change, DNBcorrelation change, or any plant modification that would be unbounded by the existing UFSARChapter 15.17 analysis assumptions.