ML072680013

Issuance of Amendment 262 Regarding Reactor Coolant System Pressure-Temperature Limit

ML072680013
Person / Time
Site:	Three Mile Island
Issue date:	10/15/2007
From:	Bamford P J NRC/NRR/ADRO/DORL/LPLI-2
To:	Crane C M AmerGen Energy Co
Bamford, Peter J., NRR/DORL 415-2833
Shared Package
ML072680011	List: ML072680013 ML072880418
References
TAC MD4910
Download: ML072680013 (15)

Text

October 15, 2007Mr. Christopher M. CranePresident and Chief Executive Officer AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OFAMENDMENT REGARDING REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMIT (TAC NO. MD4910)

Dear Mr. Crane:

The Nuclear Regulatory Commission (Commission) has issued the enclosed AmendmentNo. 262 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated March 22, 2007, as supplemented by letter dated July 25, 2007.The amendment consists of changes to various technical specifications related to the variablelow reactor coolant system pressure-temperature core protection safety limit, which is being changed to accommodate the introduction of AREVA NP's Mark-B-HTP fuel design in the TMI-1 cycle 17 reload (fall 2007).A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included inthe Commission's biweekly Federal Register notice. Sincerely,/ra/Peter J. Bamford, Project ManagerPlant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-289

Enclosures:

1. Amendment No. 262 to DPR-50 2. Safety Evaluationcc w/encls: See next page October 15, 2007Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OFAMENDMENT REGARDING REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMIT (TAC NO. MD4910)

Dear Mr. Crane:

The Nuclear Regulatory Commission (Commission) has issued the enclosed AmendmentNo. 262 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated March 22, 2007, as supplemented by letter dated July 25, 2007.The amendment consists of changes to various technical specifications related to the variablelow reactor coolant system pressure-temperature core protection safety limit, which is being changed to accommodate the introduction of AREVA NP's Mark-B-HTP fuel design in the TMI-1 cycle 17 reload (fall 2007).A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included inthe Commission's biweekly Federal Register notice. Sincerely,/ra/Peter J. Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-289

Enclosures:

1. Amendment No. 262 to DPR-50 2. Safety Evaluationcc w/encls: See next pageDISTRIBUTIONPUBLICLPL1-2 R/FRidsNrrDorlLPL1-2RidsNrrPMPBamfordRidsRgn1MailCenterGHill(2)RidsOgcRpRidsNrrDirsItsbRidsNrrLACSolaRidsNrrLAABaxterRidsDorlDprSSun, NRR YOrechwa, NRRRidsNrrDssSrxbRidsAcrsAcnwMailCenter RidsNrrDssSnpbBMarcus, NRRRidsNrrDeEicbAccession Nos.: Package/ML072680011; Amendment/ML072680013; TecSpecs/ ML072880418; *SE provided, no substantive changesOFFICELPL1-2/PMLPL2-2/LALPL1-2/LASRXB/BC*SNPB/BC*EICB/BC*ITSB/BCOGCLPL1-2/BCNAMEPBamfordCSolaABaxterGCranstonAMendiolaWKemperTKobetzEWilliamsonHChernoff(w/co mments)DATE9/25/079/27/079/27/0709/06/200707/02/200709/12/200710/04/0710/11/0710/12/07Official Record Copy Three Mile Island Nuclear Station, Unit 1 cc:

Site Vice President - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057Vice President - Operations, Mid-Atlantic AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348Vice President - Licensing and Regulatory AffairsAmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555Regional Administrator Region I U.S. Nuclear Regulatory Commission

475 Allendale Road King of Prussia, PA 19406ChairmanBoard of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120ChairmanBoard of Supervisors of Londonderry Township R.D. #1, Geyers Church Road Middletown, PA 17057Senior Resident Inspector (TMI-1)U.S. Nuclear Regulatory Commission P.O. Box 219 Middletown, PA 17057Director - Licensing and Regulatory AffairsAmerGen Energy Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348DirectorBureau of Radiation Protection Pennsylvania Department of Environmental Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469Plant Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057Regulatory Assurance Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P.O. Box 480 Middletown, PA 17057Ronald Bellamy, Region IU.S. Nuclear Regulatory Commission

475 Allendale Road King of Prussia, PA 19406Michael A. SchoppmanFramatome ANP Suite 705 1911 North Ft. Myer Drive Rosslyn, VA 22209Dr. Judith JohnsrudNational Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803Eric EpsteinTMI Alert 4100 Hillsdale Road Harrisburg, PA 17112Correspondence Control DeskAmerGen Energy Company, LLC P.O. Box 160 Kennett Square, PA 19348 Three Mile Island Nuclear Station, Unit 1 cc:

Manager Licensing - Three Mile Island NuclearStation, Unit 1 Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348Mr. Christopher M. CranePresident and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555Assistant General CounselAmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348Associate General CounselExelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 AMERGEN ENERGY COMPANY, LLCDOCKET NO. 50-289THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 262License No. DPR-501.The Nuclear Regulatory Commission (the Commission or NRC) has found that:A.The application for amendment by AmerGen Energy Company, LLC (the licensee),dated March 22, 2007, as supplemented by letter dated July 25, 2007, 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 Title 10 of the Code of FederalRegulations (10 CFR) Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act, andthe rules and regulations of the Commission;C.There is reasonable assurance: (i) that the activities authorized by this amendment canbe conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;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 Technical Specifications as indicated inthe attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No. DPR-50 is hereby amended to read as follows:(2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised through AmendmentNo. , are hereby incorporated in the license. The AmerGen Energy Company, LLC, shall operate the facility in accordance with the Technical Specifications. 3.This license amendment is effective as of its date of issuance and shall be implemented within30 days of issuance.FOR THE NUCLEAR REGULATORY COMMISSION/ra/Harold K. Chernoff, ChiefPlant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-50 and the Technical SpecificationsDate of Issuance: October 15, 2007 ATTACHMENT TO LICENSE AMENDMENT NO. 262FACILITY OPERATING LICENSE NO. DPR-50DOCKET NO. 50-289Replace page 3 of Facility Operating License No. DPR-50 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicatingthe area of change.Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.RemoveInsertvivi2-4a2-4a 2-102-10 2-112-11 4-44-4 4-7a4-7a SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 262 TO FACILITY OPERATING LICENSE NO. DPR-50AMERGEN ENERGY COMPANY, LLCTHREE MILE ISLAND NUCLEAR STATION, UNIT 1DOCKET NO. 50-28

91.0 INTRODUCTION

By application dated March 22, 2007, as supplemented by letter dated July 25, 2007, AmerGenEnergy Company, LLC (the licensee), requested changes to the Technical Specifications (TSs)for Three Mile Island Nuclear Station, Unit 1 (TMI-1). The supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 24, 2007(72 FR 20377).The amendment consists of changes to various TSs related to the variable low reactor coolantsystem (RCS) pressure-temperature core protection safety limit (CPSL), which is being changed to accommodate the introduction of AREVA NP's Mark-B high thermal performance (HTP) fuel design in the TMI-1 cycle 17 reload (fall 2007). The Mark-B-HTP fuel assemblies have a modified design that reduces the likelihood of fuel rod defects caused by spacer grid to fuel rod fretting. This modified design has a higher pressure drop across the fuel assemblies relative to residual fuel. Therefore, the mixed-core thermal-hydraulic conditions require more restrictive safety limits and more restrictive limiting safety system settings (LSSSs) for the reactor protection system. The licensee has developed these more restrictive limits in accordance with the methods described in the Nuclear Regulatory Commission (NRC, Commission)-approved Topical Report BAW-10179P-A, ASafety Criteria andMethodology for Acceptable Cycle Reload Analyses,@ using the BHTP critical heat fluxcorrelation described in NRC-approved Topical Report BAW-10241P-A, ABHTP DNBCorrelation Applied with LYNXT.

@Specifically, the proposed changes would:(1)revise the CPSL specified in Figure 2.1-1 of TS 2.1, "Safety Limits - ReactorCore;" and(2)revise the trip setting limits of the variable low RCS pressure trip setpoint (TSP)in TS Table 2.3-1, "Reactor Protection System Trip Setting Limits," and Figure 2.3-1 "Protection System Maximum Allowable Setpoints," and (3)add notes (a) and (b) to TS Table 4.1-1, "Instrument Surveillance Requirements,"for the channel calibration and channel function test of Function 11, "Reactor Coolant Pressure-Temperature Comparator," to reflect the as-found/as-left acceptance criteria. (4)revise the Table of Contents, List of Tables, to correctly identify the page numberfor Table 2.3-1 Reactor Protection System Trip Setting Limits. This change is administrative in nature, provides a correct page reference to the associated table and is therefore acceptable to the NRC staff. It will not be discussed further in this evaluation.

2.0 REGULATORY EVALUATION

The construction permit for TMI-1 was issued by the Atomic Energy Commission (AEC) onMay 18, 1968, and an operating license was issued on April 19, 1974. The plant design approval for the construction phase was based on the proposed General Design Criteria (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafterreferred to as "draft GDC"). The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria for NuclearPower Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referredto as "final GDC").Differences between the draft GDC and final GDC included a consolidation from 70 to64 criteria. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes TMI-1. Plants licensed before the final GDC were promulgated in 1971 are presumed to comply with the intent of the final GDC because those licenses were granted using comparable evaluation criteria. The TMI-1 Updated Final Safety Analysis Report (UFSAR),Section 1.4 provides an evaluation of the design bases of TMI-1 against the draft GDC.Draft GDC Criterion 6, "Reactor Core Design," and Criterion 14, "Core Protection Systems," areapplicable to this amendment. Criterion 6 requires that the reactor core be designed to function throughout its lifetime without exceeding acceptable fuel damage limits, which have been stipulated and justified. This capability must be provided under all expected conditions of normal operation with appropriate margins for uncertainties and for transient conditions, which can be anticipated. Criterion 14 states that core protection systems be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits. Final GDC 10 specifies that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOO). For the purposes of this amendment, an evaluation of conformance with the final GDC 10 will also ensure continued conformance with the UFSAR evaluation of conformance with the draft GDC sections listed above. The NRC regulatory requirements related to the content of the TSs are set forth in10 CFR 50.36, "Technical specifications," which requires that the TSs include LSSSs. This regulation requires, in part, that "Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded."

Accordingly, the CPSL and the variable low RCS pressure trip function must be included in the TSs. The NRC staff evaluation of the proposed changes will assure continued compliance with the requirements of 10 CFR 50.36.NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from TechnicalSpecifications," allows the removal of cycle-dependent variables from TSs, provided the values of these variables are determined with NRC-approved methodologies and are included in a core operating limits report. This approach has been extended to the cycle-dependent protective and maximum allowable setpoint limits. (Memorandum to Jose A. Calvo dated July 19, 1989, from David C. Fisher, "Minutes of Meeting with the B&W [Babcock and Wilcox] Owners Group (B&WOG) on the Technical Basis and Scope of the B&WOG Core Operating Limits Report (COLR)," ADAMS Accession No. 8908080374).Regulatory Guide 1.105 - "Setpoints for Safety-Related Instrumentation,@ describes a methodacceptable to the NRC staff for complying with the NRC regulations for ensuring that setpointsfor safety-related instrumentation are initially within and remain within the TS limits.Regulatory Issue Summary 2006-17, ANRC Staff Position on the Requirements of10 CFR 50.36, "Technical specifications," Regarding Limiting Safety System Settings duringPeriodic Testing and Calibration of Instrument Channels,@ dated August 24, 2006, providesadditional clarification on the requirements of 10 CFR 50.36.

3.0 TECHNICAL EVALUATION

3.1CPSLThe CPSL curve in TS Figure 2.1-1, "Core Protection Safety Limit," is the analytical limit for thevariable low RCS pressure trip function. This line represents the most severe pressure and temperature conditions that the core can operate in a steady-state condition without violating the departure from nucleate boiling (DNB) criterion, namely, the thermal design limit (TDL) of the plant. The line corresponds to the reactor core outlet pressure and reactor core outlet temperature conditions at which the DNB ratio (DNBR) is equal to or greater than the minimum DNBR limit predicted for the maximum possible overpower of 112 percent rated thermal power and the required minimum RCS flow rate of 102 percent of the design flow. The acceptable operating region in TS Figure 2.1-1 represents the operating window bounded by the RCS high pressure, RCS low pressure, and variable low RCS pressure trip functions. 3.1.1Proposed TS Figure 2.1-1 The current CPSL curve in TS Figure 2.1-1 was determined based on the design of the existingMark-B12 fuel placed in the reactor core for current operation. Because of the licensee-planned reload with the Mark-B-HTP fuel for Cycle 17 and subsequent cores that would result in a more restrictive "Acceptable Operation" region, the license amendment request (LAR) would revise the current CPSL curve, creating a more restrictive acceptable operation region. The allowable values (AVs) of the variable low pressure trip (VLPT) setpoints are also revised accordingly. No change is made to the RCS high pressure and RCS low pressure trip functions.3.1.2Methods for Calculating the CPSL The proposed CPSL curve in TS Figure 2.1-1 was based on an updated analysis in support ofthe proposed Mark-B-HTP fuel design and the use of the BHTP correlation. The thermal hydraulic analysis was performed using the NRC-approved statistical core design (SCD) methodology described in the NRC-approved topical report BAW-10187P-A, "Statistical Core Design for B&W-Designed 177FA Plants," dated March 1994. In the SCD methodology, the uncertainty distributions of statistically-treated parameters were determined. To account for the effects of uncertainties of these parameters, the SCD DNBR limit, referred to as the statistical design limit (SDL), is determined using a Monte Carlo propagation of these uncertainties in convolution with the DNBR limit of the DNB correlation. The SDL (proprietary value) for TMI-1 Cycle 17 was calculated with the NRC-approved LYNXT code (BAW-10156P-A, "LYNXT Thermal-Hydraulics Code," revision 1, dated February 1996) and with the nominal values and ranges for the state parameters and uncertainty parameters that are consistent with that described in the NRC-approved report (BAW-10187P-A). The TDL used for the determination of the CPSL curve and for safety analyses is generallysufficiently greater than the SDL to provide the necessary DNB margin to offset cycle-specific needs, such as the transition mixed-core DNB penalty. In the analysis, the TDL was calculated based on the total value of the SDL value, the mixed-core DNB penalty and reserved DNB margin. In support of the proposed CPSL curve in TS Figure 2.1-1, the licensee provided in of the LAR letter dated March 22, 2007, the calculated values for the SDL, TDL, and the mixed-core DNB penalty for Cycle 17 and subsequent cycles when the higher-resistance Mark-B-HTP fuel assemblies co-exist with the existing fuel assemblies in the core. 3.1.3Determination of the Mixed-Core DNB Penalty The Mark-B-HTP fuel design has slightly different hydraulic characteristics as compared to theresident Mark-B12 fuel design at the lower end fitting and at all the spacer grids. The net effect of the differences causes flow diversion out of fuel flow channels and results in a lower flow rate for the Mark-B-HTP fuel. The mixed-core DNB penalty is used to account for a decrease in the calculated DNBRs resulting form the lower flow rate. In determining the mixed-core DNB penalty, the licensee analyzed the DNB performance for a full core of Mark-B-HTP for statepoint conditions from the CPSL curve in TS Figure 2.1-1, for the limiting DNB AOO, and forthe range of axial power shapes that are used for calculating the CPSL and core operating limits. In order to calculate a maximum value of the mixed-core DNB penalty, the licensee used a core model in which a certain number of the Mark-B-HTP fuel assemblies were placed into the core of the resident fuel in a conservative manner to maximize flow diversion out of the limiting Mark-B-HTP fuel assembly. In response to a Request for Additional Information dated July 25, 2007, the licensee stipulated that placement of the limiting Mark-B-HTP fuel assembly at the center of the core and surrounding it with the resident fuel design, with the remaining Mark-B-HTP fuel assemblies placed on the core periphery, results in a lowest DNBR for the limiting Mark-B-HTP fuel assembly in the mixed core. The mixed-core DNB penalty for a specific mixed core model was calculated based on the largest DNBR difference between thelimiting Mark-B-HPT fuel rod in a full core model of the Mark-B-HTP fuel and a specific mixed core model (with the resident Mark-B12 fuel) for all of the statepoints, AOOs, and axial power shapes. Since the results of the licensee's sensitivity study indicated that the mixed-core DNB penalty decreases as the number of Mark-B-HTP fuel assemblies increases in the core, the licensee used the calculated DNB penalty for TMI-1 Cycle 17 with a minimum of 72 Mark-B-HTP fuel assemblies to provide protection for TMI-1 cores transitioning from Mark-B12 to Mark-B-HTP fuel designs. Since cores subsequent to Cycle 17 will have more Mark-B-HTP fuel assemblies in place, resulting in a decreased DNB penalty, the NRC staff determined that the DNB penalty based on the Cycle 17 with 72 Mark-B-HTP assemblies will provide additional DNB margin for cores subsequent to Cycle 17, and therefore, concluded that the use of the DNB penalty for Cycle 17 and subsequent cores is conservative and acceptable. The NRC staff also found that in the analysis supporting the proposed CPSL curve, the TDL(proprietary value) provides a sufficiently large margin from the SDL (proprietary value) to cover the mixed-core DNB penalty and other cycle-specific needs. Therefore, the NRC staff determined that the TDL value is acceptable. Since the SDL is calculated based on the NRC-approved LYNXT code, BHTP correlation andSCD methodology, and since the revised CPSL curve in TS Figure 2.1-1 is based on the TDL having sufficient margin to the SDL, the operation within the revised "Acceptable Operation" region with the revised CPSL curve continues to assure that the SAFDLs will not be exceeded during normal operating conditions and AOOs. Therefore, the NRC staff concluded that the revised TS Figure 2.1-1 is consistent with the final GDC 10 and meets the requirements of 10 CFR 50.36, and TS Figure 2.1-1 is acceptable. 3.2 Reactor Protection System Setpoints3.2.1Discussion As part of the proposed LAR, TS Table 2.3-1, and Figure 2.3-1, has been modified to revise theVLPT setpoint to reflect the implementation of the Mark-B-HTP fuel design in Cycle 17.The VLPT is a safety limit-related trip since it initiates an automatic reactor trip and providesreactor protection for the core safety limit contained in TS Figure 2.1-1. This protection is accomplished by tripping the reactor before the system parameters exceed the pressure-temperature combination specified in TS Figure 2.1-1, and, thereby, prevent reaching DNB limits in the core. The revised limits specified in TS Table 2.3-1, for the variable low RCS pressure, are computed based on the AREVA NP

=s NRC-approved reload methodology ofTopical Report BAW-10179P-A, Section 7.6, AVariable Low RC Pressure Trip.

@ This is thesame methodology used to develop the current TMI-1 variable low RCS pressure limit.The calculations of the VLPT limiting setpoint are consistent with Method 1 inISA-RP67.04-2000, Section 7.3, in accordance with American National Standards Institute (ANSI)/Instrument Society of America (ISA) Standard 67.04.01-2000, ASetpoints for NuclearSafety-Related Instrumentation,@ and Recommended Practice ISA-RP67.04.02-2000, AMethodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.

@ The applicable portions of ANSI/ISA-67.04.01-2000 and ISA-RP67.04.02-2000 are equivalentto the corresponding NRC-endorsed sections of ANS I/ISA-S67.04-Pa rt 1-1994. These calculations consist in establishing the relation between three quantities: The analyticallimit (AL), an AV, and the TSP. The AV provides a means to identify an unacceptable instrument performance that may require corrective actions. In Method 1 of ISA-RP67.04-2000, the AV is determined by calculating the instrument channel uncertainty without including drift, calibration uncertainties, and uncertainties observed during normal operations. This computed uncertainty is subtracted from the AL, whose value is derived from the TDL, to establish the AV.

The TSP is then determined by subtracting from the AV the combination of the uncertainties:

drift, calibration uncertainties, and uncertainties observed during normal operation. 3.2.2AV for TS Table 2.3-1, Function 8, "Variable low reactor coolant system pressure, psig[pounds per square inch gauge] min. [minimum]"The licensee derived the AV as follows:a.A pressure-temperature curve consisting of the limiting acceptable pressure andtemperature values from the safety analysis was developed. This pressure-temperature curve is provided in TS Figure 2.1-1.b.The pressure-temperature curve bows outward toward lower pressures andhigher temperatures. The licensee used a straight line connecting the endpoints in lieu of the pressure-temperature curve developed above. This line yields a higher pressure at each temperature, and is, therefore, conservative in terms of limiting the DNBR.c.The AV is based on the application of the pressure differential between the coreoutlet and hot leg tap plus a total loop uncertainty (TLU) of +/- 28.148 psig. The TLU is determined by applying the square root of the sum of the squares technique to those uncertainties that are independent, random, and approximately normally distributed. Other uncertainty components are combined algebraically. The TLU includes allowances for maintenance and test equipment accuracy and instrument drift during the surveillance test interval. The pressure differential and TLU are conservatively added to the AL to obtain the AV. The AV for TS Table 2.3-1, Function 8 is "16.21 Tout - 7973." The AV is also shown in TS Figure 2.3-1 as "VLPT = 16.21 Tout - 7973 psig." 3.2.3Nominal Trip Setpoint (NSP)a.The licensee reduced the slope of the pressure-temperature line from16.21 pounds per square inch (psi) per degree Fahrenheit (°F) to 14.29 psi per °F. The slope of the line has been reduced by rotating the AV equation clockwise around the point where the RCS temperature is equal to 618.8 °F, the RCS temperature LSSS. The modified pressure-temperature line will indicate higher pressure at each value of temperature, and is, therefore, conservative relative to the AV, the originally constructed line, and the original pressure-temperature curve in terms of limiting DNBR. b.A total margin of 40 psig has been added to the adjusted pressure-temperatureline to obtain the NSP. The total margin includes instrument hardware and process errors and additional discretionary margin. The total margin is composed of a TLU of +/- 28.148 psig, a surveillance test procedures NSP as-left tolerance of +/-1.6 psig, and additional discretionary margin of 10 psig. The NRC staff has reviewed the licensee's proposed TS changes concerning AVs and limitingsetpoints. The implementation of the limiting setpoints incorporated into TS Table 2.3-1 and TS Figure 2.3-1, as described above, is in accordance with RG 1.105, revision 3. This method is an acceptable approach, previously endorsed by the NRC staff, for ensuring that setpoints for safety related instrumentation provide protection to ensure that NRC regulations regarding saftey related instrumentation are met. Based on its review of the licensee's submittals, including the licensee's implementation of RG 1.105, revision 3 for the VLPT setpoint calculation, the staff finds that the AV and NSP determinations for TS Table 2.3-1, Function 8 and in TS Figure 2.3-1, meet the requirements of 10 CFR 50.36(c)(1)(ii)(A) and are therefore acceptable.3.3 As-Found Setpoint Evaluation The licensee proposed the addition of notes to TS Table 4.1-1, for Function 11. The notesprovide requirements for operability determinations for as-found setpoints for this LSSS. The TS along with the proposed notes provide for the following:a.As-Found Setpoint Exceeds AVIf the surveillance test as-found TSP exceeds the AV, the TS will require, (1) therequired TS actions are taken, and (2) the instrument channel will be declared inoperable pending further evaluation and calibration.b.As-Found Setpoint Conservative with Respect to AVIf the surveillance test as-found TSP is conservative with respect to the AV, butexceeds the pre-defined limits for as-found tolerance, the TS will require, (1) a determination if the instrument is functioning as required prior to returning the channel to service; if it cannot be determined that the instrument is functioning as required, the channel will be declared inoperable, (2) the trip set point is to be reset within the surveillance test procedures as-left setting tolerance band, and (3) the condition will be entered into the Corrective Action Program.c.As-Found Setpoint Conservative with Respect to Pre-Defined LimitsIf the surveillance as-found TSP is conservative with respect to the pre-definedlimits, but exceeds the as-left setting tolerance band, the TS will require, (1) the trip set point to be reset within the surveillance test procedures as-left setting tolerance band.The notes require that the channel be declared inoperable if it cannot be adjusted within the as-left tolerance band around the nominal trip setpoint at the end of the surveillance. The NRC staff has reviewed the licensee's proposed TS changes concerning the surveillancetesting notes to be added to the TS. The addition of notes is consistent with the methodology provided in NRC RIS 2006-17 and RG 1.105, revision 3, which are methods previously accepted by the NRC staff for showing compliance with NRC regulations. The notes provide actions to be taken during channel surveillance that ensure continued operability of the VLPT setpoint instrumentation. The staff therefore finds the proposed notes being added by this LAR continue to meet the requirements of 10 CFR 50.36(c)(3) and are acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. TheCommission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (72 FR 20377). Accordingly, the amendment meets the eligibility criteria for categorical exclusion 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.Principal Contributors:Yuri OrechewaBarry Marcus Summer SunDate: October 15, 2007