ML062480328

D.C. Cook, License Amendment 296, Regarding Elimination of the Resistance Temperature Detector (RTD) Bypass Loop

ML062480328
Person / Time
Site:	Cook
Issue date:	10/06/2006
From:	Tam P S NRC/NRR/ADRO/DORL/LPLIII-1
To:	Nazar M K Indiana Michigan Power Co
P. TAM
Shared Package
ML062830153	List: ML062480328 ML062830313
References
TAC MD2106
Download: ML062480328 (13)

Text

October 6, 2006Mr. Mano K. NazarSenior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNIT 1 (DCCNP-1) - ISSUANCE OFAMENDMENT REGARDING ELIMINATION OF THE RESISTANCE TEMPERATURE DETECTOR (RTD) BYPASS LOOP (TAC NO. MD2106)

Dear Mr. Nazar:

The Commission has issued the enclosed Amendment No. 296 to Renewed Facility OperatingLicense No. DPR-58 for DCCNP-1. The amendment consists of changes to the TechnicalSpecifications in response to your application dated May 31, 2006.The amendment approved elimination of the RTD bypass piping and installing fast responsethermowell-mounted RTDs in the reactor coolant system loop piping. The amendment alsorevised Surveillance Requirement 3.3.1.15 of the Technical Specifications, deleting therequirement to perform surveillance on the reactor coolant system RTD bypass loop flow rate. A copy of our related safety evaluation is enclosed. A Notice of Issuance will be included in theCommission's next biweekly Federal Register notice.Sincerely,/RA/Peter S. Tam, Senior Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-315

Enclosures:

1. Amendment No. 296 to DPR-58

2. Safety Evaluationcc w/encls: See next page Mr. Mano K. NazarOctober 6, 2006Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNIT 1 (DCCNP-1) - ISSUANCE OFAMENDMENT REGARDING ELIMINATION OF THE RESISTANCE TEMPERATURE DETECTOR (RTD) BYPASS LOOP (TAC NO. MD2106)

Dear Mr. Nazar:

The Commission has issued the enclosed Amendment No. 296 to Renewed Facility OperatingLicense No. DPR-58 for DCCNP-1. The amendment consists of changes to the TechnicalSpecifications in response to your application dated May 31, 2006.The amendment approved elimination of the RTD bypass piping and installing fast responsethermowell-mounted RTDs in the reactor coolant system loop piping. The amendment alsorevised Surveillance Requirement 3.3.1.15 of the Technical Specifications, deleting therequirement to perform surveillance on the reactor coolant system RTD bypass loop flow rate. A copy of our related safety evaluation is enclosed. A Notice of Issuance will be included in theCommission's next biweekly Federal Register notice.Sincerely,/RA/Peter S. Tam, Senior Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-315

Enclosures:

1. Amendment No. 296 to DPR-58

2. Safety Evaluation cc w/encls: See next pageDISTRIBUTIONPUBLICGHill, OISLPL3-1 R/FRidsOGCRpRidsNrrLATHarrisRidsNrrPMPTam RidsDorlDprRidsNrrDirsltsbRidsAcrsAcnwMailCenter SMirandaRidsRgn3MailCenterPackage Accession Number: ML062830153Amendment Accession Number: ML062480328TS Page Accession Number: ML062830313OFFICENRR:LPL3-1/PMNRR:LPL3-1/LANRR:SPWB/BCOGCNRR:LPL3-1/BC(A)NAMEPTamTHarrisJNakoski*JRund#LRaghavan forMMurphyDATE09/13/0609/12/0608/31/0609/21/0610/6/06

• Safety evaluation transmitted by memo of 8/31/06. Concurred on changes to the safety evaluation on 10/5/06.# K. Winsberg reviewed revised version on 10/5/06. OFFICIAL RECORD COPY Donald C. Cook Nuclear Plant, Units 1 and 2 cc:Regional Administrator, Region IIIU.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville RoadLisle, IL 60532-4351Attorney GeneralDepartment of Attorney General 525 West Ottawa Street Lansing, MI 48913Township SupervisorLake Township Hall

P.O. Box 818 Bridgman, MI 49106U.S. Nuclear Regulatory CommissionResident Inspector's Office 7700 Red Arrow Highway Stevensville, MI 49127James M. Petro, Jr., EsquireIndiana Michigan Power Company One Cook Place Bridgman, MI 49106Mayor, City of Bridgman P.O. Box 366 Bridgman, MI 49106Special Assistant to the GovernorRoom 1 - State Capitol Lansing, MI 48909Susan D. SimpsonRegulatory Affairs Manager Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106Michigan Department of Environmental Quality Waste and Hazardous Materials Div.

Hazardous Waste & Radiological Protection Section Nuclear Facilities Unit Constitution Hall, Lower-Level North 525 West Allegan Street P. O. Box 30241 Lansing, MI 48909-7741Lawrence J. Weber, Plant ManagerIndiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106Mark A. Peifer, Site Vice President Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 INDIANA MICHIGAN POWER COMPANYDOCKET NO. 50-315DONALD C. COOK NUCLEAR PLANT, UNIT 1AMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 296 License No. DPR-581.The U.S. Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Indiana Michigan Power Company (thelicensee) dated May 31, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by this amendmentcan be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'sregulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Renewed Facility OperatingLicense and Technical Specifications as indicated in the attachment to this license amendment.3.This license amendment is effective as of its date of issuance and shall be implementedprior to entry into Mode 2 from the fall 2006 refueling outage.FOR THE NUCLEAR REGULATORY COMMISSION/RA by L. Raghavan/Martin C. Murphy, Acting ChiefPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Operating License Date of Issuance: October 6, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 296RENEWED FACILITY OPERATING LICENSE NO. DPR-58DOCKET NO. 50-315Replace the following page of Renewed Facility Operating License No. DPR-58 with theattached revised page. The change area is identified by a marginal line.REMOVEINSERT 33Replace the following page of Appendix A, Technical Specifications, with the attached revisedpage. The change area is identified by a marginal line.REMOVEINSERT3.3.1-93.3.1-9 and radiation monitoring equipment calibration, and as fission detectors inamounts as required.(4)Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess anduse in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5)Pursuant to the Act and 10 CFR 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.C.This renewed operating license shall be deemed to contain and is subject to theconditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of thee Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:(1)Maximum Power LevelThe licensee is authorized to operate the facility at steady state reactor corepower levels not to exceed 3304 megawatts thermal in accordance with the conditions specified therein.(2)Technical SpecificationsThe Technical Specifications contained in Appendix A and Appendix B, asrevised through Amendment No. 296, are hereby incorporated in the renewedl operating license. The licensee shall operate the facility in accordance with theTechnical Specifications.(3)Less Than Four Loop OperationThe licensee shall not operate the reactor at power levels above P-7 (as definedin Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operatinglicense) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than found loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.

(4)Indiana Michigan Power Company shall implement and maintain, in effect, allprovisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SERs datedDecember 12, 1977, July 31, 1979, January 10, 1981, February 7, 1983,November 22, 1983, December 23, 1983, March 16, 1984, August 27, 1985 Renewed License No. DPR-58 Amendment No. 1 through 295 , 296 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58INDIANA MICHIGAN POWER COMPANYDONALD C. COOK NUCLEAR PLANT, UNIT 1DOCKET NO. 50-31

51.0INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC, Commission) dated May 31,2006 (Accession No. ML061600449), Indiana Michigan Power Company (I&M, or the licensee) requested an amendment to the Operating License for Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1). The proposed amendment would allow the licensee to remove the resistancetemperature detector (RTD) bypass piping and install fast response thermowell-mounted RTDs located in the reactor cool ant system (RCS) loop piping. With this approval, the DCCNP-1Technical Specifications (TS) would be modified, deleting Note 1 (regarding verification of reactor coolant system RTD bypass loop flow rate) from Surveillance Requirement (SR) 3.3.1.15. The licensee stated that removal of the RTD bypass piping would occur during the fall 2006refueling outage. The licensee expected that removal of the RTD bypass piping would result ina reduction of approximately 30 person-rem in radiation exposure to personnel working in containment during refueling outages. The licensee also expected that removal of the RTDbypass piping would reduce refueling outage costs and the likelihood of RCS leakage.

2.0 REGULATORY EVALUATION

Removal of the RTD bypass piping involves the replacement of each RTD, installed in the RTDbypass piping, with three RTDs, installed in thermowells, that are situated 120 degrees apart around the RCS pipe walls. The new RTDs will have a slightly longer response time, which willalter the time response characteristics of the overtemperature delta temperature (OTT) andoverpower delta temperature (OPT) reactor protection logic. The NRC staff's review,therefore, focused upon the accident analyses that rely upon reactor trips from the OTT and OPT reactor protection logic. In particular, the rod withdrawal at power and the inadvertent opening of a pressurizer safety orrelief valve events can produce conditions that demand an OTT reactor trip. Analyses ofthese two events are used to determine the constants and lead/lag functions used in the OTT and OPT reactor trip setpoint equations. The OPT reactor protection logic is not credited inthe accident analyses of the DCCNP-1 licensing basis. The NRC staff noted that I&Mconsidered the rod withdrawal at power event, but not the inadvertent opening of a pressurizer safety or relief valve event, to show that the OTT reactor trip function, as modified, willcontinue to prevent departure from nucleate boiling (DNB).Chapter 14 of the DCCNP-1 Update Final Safety Analysis Report (UFSAR) does not contain ananalysis of the inadvertent opening of a pressurizer safety or relief valve event. DCCNP-1 waslicensed to operate in 1974, about a year after the American Nuclear Society's Standard ANS 51.1, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor [PWR]

Plants" was issued. This standard categorizes PWR events into four classes, and sets acceptance criteria for each class. Although this standard is not referenced in the DCCNP-1licensing basis, the accident analyses therein abide by the event classes and theircorresponding acceptance criteria. I&M's application of May 31, 2006, references the predecessor of this standard, ANSI N18.2-1973. The inadvertent opening of a pressurizer safety or relief valve event analysis is potentiallyaffected by the changes that are proposed in I&M's application for amendment. This event causes an erosion of thermal margin, as RCS pressure decreases, until the reactor is tripped, either by a signal from the OTT reactor trip function or from low pressurizer pressure. Therefore, this event is a part of the NRC staff's safety evaluation. Using the review criteria set forth in Standard Review Plan (SRP) 15.6.1, the NRC staff performed a confirmatory evaluationand documented its results in Section 3.0 (on page 4 of this safety evaluation).The licensee stated that the proposed amendment, when approved, will result in a reduction ofapproximately 30 person-rem in radiation exposure to personnel performing work in containment during refueling outages. The NRC staff recognizes that this reduction is asignificant safety benefit.

3.0TECHNICAL EVALUATION

3.1Evaluation of Licensing Basis ChangeThe OTT reactor trip function provides primary protection against DNB in WestinghousePWRs. The measured T, which is indicative of nuclear power, is compared to the OTTsetpoint, which is calculated from continually updated values of measured average temperature (Tavg), pressurizer pressure, and reactor axial flux difference. A reactor trip signal is generatedwhen the measured T exceeds the OTT setpoint in two or more reactor coolant loops. The OPT reactor trip function provides protection against fuel centerline melting inWestinghouse PWRs. The OPT reactor trip function generates a reactor trip signal in amanner that is similar to the OTT reactor trip function. Replacing the current RTD bypass system with thermowell-mounted RTDs will affect the RCStemperature measurement response, and consequently, the OTT and OPT reactor tripfunctions. The total time delay for these trip functions is assumed to be 8 seconds (Table 14.1-2 of ANS 51.1). The following table (reproduced from the licensee's May 31, 2006, application) indicates thatthe total time delay for the OTT and OPT reactor trip functions would continue to be 8seconds or less, after the RTD bypass system is removed. Response Time Parameters for RCS Temperature MeasurementComponentExisting RTD Bypass System(seconds) New Fast-ResponseThermowell RTD System(seconds) RTD bypass piping transportand thermal lag4N/ARTD response time 24 Electronics signalprocessing, reactor trip signal, trip breaker opening, and rod cluster control assembly gripper release22Total Response Time 8Less than or equal to 8The current OTT and OPT reactor trip functions, as assumed in the accident analyses, aredelayed by about 8 seconds, which consist of a 6-second lag for the RTD response, and a 2-second delay for the electronic process time needed to transmit the trip signal and release the rod cluster control assemblies (RCCAs). The table indicates that, after the RTD bypass lines are removed and the RTDs are replaced, the overall OTT and OPT reactor trip responsetimes will continue to be 8 seconds or less.The principal event in DCCNP-1's licensing basis that relies upon the OTT reactor trip functionis the rod withdrawal at power. This event is analyzed over a range of initial power levels and reactivity insertion rates, and at the beginning and end of core life. The reactor trip signal is usually generated by high nuclear flux at higher reactivity insertion rates, and by OTT at lowerreactivity insertion rates. In fact, this is one of two event analyses that are used to determine the constants and lead/lag functions used in the OTT reactor trip setpoint equation. The licensee evaluated the rod withdrawal at power event, in consideration of the modified time response of the OTT trip function, and concluded that the proposed change (i.e., removal ofthe RTD bypass lines) would have no significant effect on the results. Therefore, the analysis of this event, in Chapter 14 of the UFSAR, remains valid. The NRC staff reviewed thelicensee's analysis and accepts that the analysis of this event in Chapter 14 of the UFSARremain valid.Another event that can be affected by a change in the time response characteristics of the OTT trip function is the inadvertent opening of a pressurizer safety or relief valve. Theresulting decrease in reactor coolant system pressure could lead to DNB, if the reactor is notautomatically tripped in a timely manner. The reactor trip signal can be generated by OTT orby low pressurizer pressure. A proper analysis of this event demonstrates defense-in-depth by considering two cases, that show that a reactor trip, derived from either signal, will preventDNB. In practice, however, licensing basis analyses of this event typically consider just one case, in which the reactor is assumed to be tripped from either the OTT trip signal or from thelow pressurizer pressure trip signal, whichever is generated first. The NRC staff performed a confirmatory evaluation of the inadvertent opening of a pressurizersafety or relief valve by situating the trajectory of the transient onto the DCCNP-1 core limits plot (UFSAR Figure 14.1-1). The DCCNP-1 core limits plot depicts the locus of points at whichthe departure from nucleate boiling ratio (DNBR) is at its safety analysis limit (SAL), as afunction of pressure, Tavg and T. Primary coolant flow rate is assumed to be constant. Usingthis plot, it is possible to determine the pressure, at which the DNBR is at its SAL, as a function of Tavg for a constant value of T. Using this function, and assuming a nominal operating pointwherein the T is about 65 F, and Tavg is about 579 F, then the pressure, at those conditions,would have to fall to about 1722 psia (from an initial value of 2100 psia) in order to reduce the DNBR to its SAL value. The low pressurizer pressure reactor trip signal will be generated atabout 1840 psia. Subtracting about 50 psi for pressure measurement uncertainty puts the lowpressurizer pressure reactor trip setpoint at about 1790 psia. Once the trip condition is reached, it would take about 2 seconds to transmit the signal to the RCCAs, and another 3seconds for the RCCAs to drop into (and reach the bottom of) the core. Based upon analysesof the inadvertent opening of a pressurizer safety or relief valve in other Westinghouse PWRs of comparable design and pressurizer safety or relief valve capacities, the RCS depressurization rate would be estimated to not more than 10 psi per second. Therefore, in DCCNP-1, the RCS pressure would be estimated to be about 1740 psia at the time of reactorshutdown, 5 seconds after the OTT setpoint is reached. This is st ill about 18 psi higher thanthe pressure at which the DNBR SAL would be expected to be reached. Based upon this evaluation, the NRC staff concludes that DCCNP-1 would be adequately protected by the lowpressurizer pressure reactor trip during an inadvertent opening of a pressurizer safety or relief valve event.The core limits plot (UFSAR Figure 14.1-1) also depicts the OTT trip setpoint as a function ofpressure, Tavg and T. At nominal Tavg and T conditions, the plot indicates that, during a RCSdepressurization, the OTT setpoint would be reached at about the same time as the lowpressurizer pressure reactor trip setpoint (1840 psia) is reached. There are two factors that could cause the OTT reactor trip to occur before the low pressurizerpressure reactor trip. The first factor is the penalty term, in the OTT trip setpoint equation, foradverse axial flux difference. This could reduce the OTT trip setpoint, and thereby cause the OTT trip signal to be generated sooner. However, the adverse axial flux difference penalty isconservatively not modeled in the accident analyses. The second factor is the allowance for pressure measurement uncertainty in the low pressurizer pressure trip setpoint. This allowance, which is conservatively included in accident analysis assumptions, would cause the low pressurizer pressure setpoint to be reached later. Therefore, during an inadvertent opening of a pressurizer safety or relief valve event in DCCNP-1, the OTT trip signal could beexpected to occur before the low pressurizer pressure trip. Since the latter trip function, from low pressurizer pressure, has been evaluated to be effective in protecting the DCCNP-1 corefrom DNB, then the former trip function, from the OTT trip setpoint equation, would also beeffective. Based upon this evaluation, the NRC staff concludes that DCCNP-1 would be adequatelyprotected by either the low pressurizer pressure reactor trip or the OTT reactor trip during aninadvertent opening of a pressurizer safety or relief valve event.

The OPT trip function protects against high linear power density that could lead to fuelcenterline melting. Fuel centerline melting could occur during the more severe, Condition III or IV events, such as major steam line breaks. The OPT trip, and the high nuclear flux trip,could be demanded, for example, during a full power steamline break. In Westinghouse plants, the full power steamline break has been shown to be less limiting than the zero power steamline break. Protection for the zero power steamline break is provided primarily from the low pressurizer pressure and low steamline pressure trip functions, and by automatic steamline isolation. There are no events in the DCCNP-1 licensing basis that rely upon the OPT trip forprotection.3.2Evaluation of TS ChangesThe licensee proposed to delete Note 1 of SR 3.3.1.15, which requires surveillance of the RTDloop bypass flow rate. This change is acceptable, since the RTDs and the associated bypass loops will be replaced by thermowell-mounted RTDs, a system for which verification of RTDloop bypass flow rate would no longer be applicable. The NRC staff had evaluated in detail thetechnical aspects of elimination of the RTD bypass loop in Section 3.1 above.3.3Summary of NRC staff EvaluationThe NRC staff has concluded that the safety-related effects of replacing the DCCNP-1 RTDsand the associated bypass loops with thermowell-mounted RTDs (i.e., slightly changing the response time characteristics of the OTT and OPT trip functions) are not significant. TheNRC staff concludes that there will continue to be adequate protection after the proposedmodification is implemented against the two postulated accidents that rely on the OTT tripfunction for DNB protection (i.e., rod withdrawal at power, and inadvertent opening of a pressurizer safety or relief valve). There are no events in the DCCNP-1 licensing basis that rely

upon the OPT trip for protection.

4.0STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of theproposed issuance of the amendment. The State official had no comments.

5.0ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of facility com ponentslocated within the restricted area as defined in 10 CFR Part 20 or change the surveillancerequirements. The NRC staff has determined that the amendment involves no significantincrease in the amounts, and no significant change in the types, of any effluents that may bereleased offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The NRC has previously issued a proposed finding that the amendmentinvolves no significant hazards consideration and there has been no public comment on such finding (71 FR 38182). Accordingly, the amendment meets the eligibility criteria for categoricalexclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0CONCLUSION

The NRC staff has concluded, based on the considerations discussed above, that: (1) there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimical to thecommon defense and security or to the health and safety of the public. Principal Contributors: S. Miranda C. Somers

Date: October 6, 2006