ML081230178

Issuance of License Amendment No. 256 Regarding Uprate of Shutdown Cooling System Entry Conditions

ML081230178
Person / Time
Site:	Fort Calhoun
Issue date:	05/09/2008
From:	Markley M NRC/NRR/ADRO/DORL/LPLIV
To:	Bannister D Omaha Public Power District
Markley, M T, NRR/DORL/LP4, 301-415-5723
Shared Package
ML081230175	List: ML081230178 ML081230188
References
TAC MD6993
Download: ML081230178 (25)

Text

May 9, 2008 Mr. David J. Bannister Vice President and CNO Omaha Public Power District Fort Calhoun Station FC-2-4 Post Office Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:

UPRATE OF SHUTDOWN COOLING SYSTEM ENTRY CONDITIONS (TAC NO.

MD6993)

Dear Mr. Bannister:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 256 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station (FCS), Unit No. 1. The amendment consists of changes to the Technical Specifications (TS) in response to your application dated October 12, 2007, as supplemented by letters dated March 22 and April 4, 2008.

The amendment modifies the FCS design and licensing basis to increase the shutdown cooling (SDC) system entry temperature from 300 degrees Fahrenheit (°F) to 350 °F (cold leg), and the SDC entry pressure from 250 pounds per square inch absolute (psia) to 300 psia (indicated at the pressurizer). The revised TS pages and associated basis documents, Updated Safety Analysis Report described design methodology, and American Society of Mechanical Engineers Boiler and Pressure Vessel Code as applied to the SDC heat exchangers are approved.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Michael T. Markley, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 256 to DPR-40

2. Safety Evaluation cc w/encls: See next page

Pkg ML081230175, Amdt ML081230178, License/TS Pgs ML081230188

(*) Concurrence via SE

(**) See previous concurrence OFFICE NRR/LPL4/PM NRR/LPL4/LA DE/EMCB/BC DSS/SRXB/BC NAME MMarkley (**)

JBurkhardt (**)

KManoly*

GCranston*

DATE 5/5/08 5/5/08 3/14/08 4/22/08 OFFICE DSS/SBPB/BC NRR/ITSB/BC OGC - No legal objection NRR/LPL4/BC NAME DHarrison RElliott (**)

APHodgdon (**)

THiltz DATE 4/25/08 5/6/08 5/7/08 5/9/08

Ft. Calhoun Station, Unit 1 (May 2, 2008) cc:

Winston & Strawn ATTN: James R. Curtiss, Esq.

1700 K Street, N.W.

Washington, DC 20006-3817 Chairman Washington County Board of Supervisors P.O. Box 466 Blair, NE 68008 Mr. John Hanna, Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 310 Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. Thomas C. Matthews Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550 Ms. Melanie Rasmussen Radiation Control Program Officer Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 256 Renewed License No. DPR-40

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Omaha Public Power District (the licensee), dated October 12, 2007, as supplemented by letters dated March 22 and April 4, 2008, 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 amendment can be 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 license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 256, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance and shall be implemented prior to startup from the 2008 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications Date of Issuance: May 9, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 256 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

License Page REMOVE INSERT Page 3 Page 3 Technical Specifications REMOVE INSERT 2.1 - Page 1 2.1 - Page 1 2.1 - Page 2 2.1 - Page 2 2.1 - Page 3 2.1 - Page 3 3.16 - Page 1 3.16 - Page 1

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 256 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By application dated October 12, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072890192) (Reference 1), as supplemented by letters dated March 22 and April 4, 2008 (ADAMS Accession Nos. ML080850254 and ML081070336, respectively) (References 3 and 4), Omaha Public Power District (OPPD, the licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Renewed Facility Operating License No. DPR-40) and modifications to the plant design and licensing basis for the Fort Calhoun Station (FCS), Unit No. 1.

The supplemental letters dated March 22 and April 4, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on November 20, 2007 (72 FR 65370).

The proposed amendment would revise the plant design and licensing basis to increase the shutdown cooling (SDC) system entry temperature from 300 degrees Fahrenheit (°F) to 350 °F (cold leg), and the SDC entry pressure from 250 pounds per square inch absolute (psia) to 300 psia (indicated at the pressurizer). The amendment would approve the OPPD proposed changes to the TS and associated basis documents, the Updated Safety Analysis Report (USAR) that describes the design methodology, and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) as applied to the SDC heat exchangers.

The NRC staff issued a request for additional information (RAI) on February 27, 2008 (ADAMS Accession No. ML080560007) (Reference 2). The NRC staff also held discussions with Mr. Thomas Matthews and others of the FCS staff and its contractors on January 31 and February 28, 2008, to clarify mutual understanding of issues to be included in the RAI.

The proposed amendment would revise the current TS 2.1, Reactor Coolant System [RCS],

Limiting Conditions for Operation (LCOs) Sections 2.1.1(2), 2.1.1(3), and 2.1.1(11), to permit

operation of the SDC system at RCS cold-leg temperatures above 300 °F, up to and including 350 °F. The proposed amendment would also revise TS Surveillance Requirement (SR) 3.16, Residual Heat Removal [RHR] System Integrity Testing, Section (1)a. to increase the system leakage test pressure from 250 pounds per square inch gauge (psig) to 300 psig and SR 3.16 Section (1)d. to make a typographical correction to change the spelling of the word "frequence" to "frequency." The TS 2.1 Basis and the FCS Technical Data Book (TDB) will also be revised via the TS Basis Change Control program.

2.0 REGULATORY EVALUATION

In Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), Technical Specifications, the NRC established its regulatory requirements related to the content of TS.

Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings, (2) LCOs, (3) SRs, (4) design features, and (5) administrative controls.

The rule does not specify the particular requirements to be included in a plants TS. As stated in 10 CFR 50.36(d)(2)(i), the [l]imiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulations in 10 CFR 50.36(d)(3) state that [s]urveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components will be maintained within safety limits, and that the limiting conditions for operation will be met.

In a memorandum dated September 18, 1992, the Commission approved the staff proposal in SECY-92-223, Resolution of Deviations Identified During the Systematic Evaluation Program (ADAMS Accession No. ML003763736) (Reference 5), not to apply 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, to plants with construction permits prior to May 21, 1971. FCS was licensed for construction prior to May 21, 1971, and at that time committed to the draft General Design Criteria (GDC). The draft GDC, which are similar to Appendix A, General Design Criteria for Nuclear Power Plants in 10 CFR Part 50, are contained in Appendix G of the FCS USAR.

In its license amendment request (LAR) dated October 12, 2007, the licensee appropriately identified the applicable regulations in 10 CFR 50.55a, Codes and standards. The licensee also identified the following preliminary design criteria applicable to this amendment in the Appendix G, Responses to 70 Criteria, of the FCS USAR:

FCS Design Criterion 1, Quality Standards.

FCS Design Criterion 2, Performance Standards.

FCS Design Criterion 9, Reactor Coolant Pressure Boundary.

FCS Design Criterion 33, Reactor Coolant Pressure Boundary Capability.

FCS Design Criterion 34, Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention.

FCS Design Criterion 35, Reactor Coolant Pressure Boundary Brittle Fracture Prevention.

FCS Design Criterion 36, Reactor Coolant Pressure Boundary Surveillance.

FCS Design Criterion 37, Engineered Safety Features Basis for Design.

FCS Design Criterion 38, Reliability and Testability of Engineered Safety Features.

FCS Design Criterion 41, Engineered Safety Features Performance Capability.

FCS Design Criterion 42, Engineered Safety Features Components Capability.

FCS Design Criterion 43, Accident Aggravation Prevention.

FCS Design Criterion 44, Emergency Core Cooling System Capability.

FCS Design Criterion 45, Inspection of Emergency Core Cooling Systems.

FCS Design Criterion 46, Testing of Emergency Core Cooling System Components.

FCS Design Criterion 47, Testing of Emergency Core Cooling Systems.

FCS Design Criterion 51, Reactor Coolant System Boundary Outside Containment.

FCS Design Criterion 53, Containment Isolation Valves.

FCS Design Criterion 57, Provision for Testing Isolation Valves.

FCS Design Criterion 67, Fuel and Waste Storage Decay Heat.

2.1 Low Temperature Overpressure Protection Analysis The NRC has established requirements in 10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Specifically, Appendix G to 10 CFR Part 50 requires the facility pressure/temperature (P/T) limit curves for the reactor pressure vessel (RPV) be least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the ASME Code.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR [Light-water Reactor] Edition," March 2007 (ADAMS Accession No. ML070660036) (Reference 6), Section 5.2.2, Overpressure Protection, provides the review guidance for the low temperature overpressure protection (LTOP) transient analyses.

2.2 SDC and Component Cooling Water Systems and Components The licensee proposes to change the entry temperature of the SDC system (placed in service) from 300 °F to 350 °F. The licensee evaluated the ability of the SDC system to cool the RCS temperature to 140 °F from 350 °F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as described in USAR Section 14.15 for the plants design-basis loss-of-coolant accident (LOCA). The current TS restricts the licensee, during a plant cooldown, from transitioning to the SDC system until the RCS temperature decreases below 300 °F. The proposed amendment increases this temperature limit to 350 °F, which is consistent with conditions at other comparable Combustion Engineering plants. The SDC system provides a more effective method of heat removal than using the steam generators (SGs) at this low temperature and pressure. The staff evaluation focuses on FCS Design Criteria 1, 33, and 37 for this portion of the review and whether the SDC system, in conjunction with the component cooling water (CCW) system, has a sufficient heat-removal capability to perform this intended function at the proposed pressure and temperature, considering design limitations. The licensee did not identify any regulatory precedents for uprating the SDC system entry conditions.

2.3 Structural and Pressure Boundary Integrity The NRC staffs review covers the structural and pressure boundary integrity of the SDC system piping and its associated supports and components. Technical areas covered by this review include stresses, cumulative usage factors (CUFs), and high-energy line break (HELB) locations.

The affected piping systems were originally designed in accordance with the rules of the United States of America Standards (USAS) Code for Power Piping code B31.7. The NRC staff review focused on FCS Design Criteria 1, 2, and 9 in verifying that the licensee has provided reasonable assurance of the structural and functional integrity of piping systems and their supports under normal and vibratory loadings, including those due to fluid flow, postulated accidents, and natural phenomena such as earthquakes.

The acceptance criteria are based on continued conformance with the requirements of the following regulations: (1) 10 CFR 50.55a and FCS Design Criterion 1 as they relate to structures and components being designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed; (2) FCS Design Criterion 2 as it relates to structures and components important to safety being designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; and (3) FCS Design Criterion 9 as it relates to the reactor coolant pressure boundary being designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. The specific review areas are contained in the NRC SRP Sections 3.2 and 3.9.

3.0 TECHNICAL EVALUATION

The SDC system removes decay heat from the RCS at a controlled rate from 300 °F to normal refueling temperature. The SDC system also serves as the low-pressure safety injection (LPSI) system, which connects directly to the RCS. The SDC heat exchangers branch off the discharge of the LPSI pumps and are isolated from the LPSI system when LPSI is in the injection mode. The CCW system removes RCS decay heat from the SDC heat exchangers and transfers the heat to the raw service water system via the CCW heat exchangers. The SDC

also functions to maintain proper RCS temperature during refueling. The SDC can also provide flow to the reactor during the long-term core cooling mode following a large-break LOCA, or provide cooled containment sump water to the high-pressure safety injection (HPSI) pumps for long-term cooling.

3.1 Low Temperature Overpressure Protection Analysis 3.1.1 LTOP Evaluation Section 5.2.2 of the NRC SRP specifies that the LTOP system should be designed in accordance with the guidance of Reactor System Branch Technical Position (RSB) 5-2. The RSB 5-2 specifies that the LTOP system should be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to prevent the RCS pressure from exceeding 10 CFR Part 50 Appendix G limits while operating at low temperatures.

The existing LTOP system for the FCS is provided by one of two means depending on the transient. For mass-addition transients, the LTOP system relies on opening of the pressurizer power-operated relief valves (PORVs) to mitigate the consequences of the events. For this mode of the LTOP protection, the PORV control system compares the pressurizer pressure to LTOP setpoints (a function of pressure versus cold-leg temperature). If the pressurizer pressure exceeds the LTOP setpoints, the PORVs open and discharge to the quench tank in the containment.

For heat-addition transients, the consequence mitigation is controlled by a TS requirement for a minimum steam void in the pressurizer prior to initiation of a heat-addition transient. The steam void maintains pressure as it collapses, since the steam is at saturated conditions and experiences a moderate temperature rise as it is condensed. The time required for the void collapse allows the RCS to heat up and reach equilibrium with the temperature of the fluid in the RCS secondary side. Once the temperatures of the fluid in the RCS primary and secondary sides are equal, the consequences of the transient are effectively mitigated.

In the LAR (Reference 1), the licensee proposed a new LTOP enabling temperature of 350 °F, which is increased from 300 °F in the current TS 2.1.1(11)(b). The supporting LTOP analyses documented in References 1, 3, and 4 were performed to show adequacy of (1) the minimum LTOP pressure setpoints of the pressurizer PORVs, and (2) the minimum steam void required in the TS to prevent the RCS pressure from exceeding the reactor P/T limits calculated in accordance with the requirements of Appendix G of 10 CFR Part 50. In the LTOP analyses, two types of events were analyzed:

1.

mass-addition transients such as an event caused by a spurious safety-injection (SI) signal; and

2.

heat-addition transients such as the startup of one reactor coolant pressure (RCP) event while the SGs contain hot fluid in the secondary side.

These events were previously identified by the licensee as the limiting mass-and heat-addition events for the design of the LTOP system at the FCS.

3.1.1.1 The Computer Code Used for the LTOP Analyses The LTOP analyses were performed using the RELAP5 Mod 3.2 code. The code simulated a multi-loop system using a model containing a reactor vessel, hot-and cold-leg piping, SGs, and pressurizer. The NRC staff found that the use of the RELAP5 Mod 3.2 code for the LTOP analyses was previously approved (Reference 6) by NRC in support of the LTOP design at the FCS. In its March 22, 2008, response to the NRC RAI, the licensee stated that no model changes were made to the RELAP5 code after it was approved by NRC. In the proposed LTOP analyses, the values of input parameters to the code were changed to reflect the SG and pressurizer replacement implemented at FCS. The specific changes involved head losses through the SG, SG heat transfer area, SG volumes, and pressurizer volume. The net impact of the changes resulted in a small increase of the margin to the P/T limit because of a larger pressurizer volume that decreased the pressure increase rate during overpressurization events.

Also, the decay heat was increased to include a power increase for a potential power uprate request for FCS that has not yet been submitted by the licensee. In addition, the safety evaluation (SE) approving the RELAP5 code identified that the maximum SG temperature was 314 °F, while in the proposed LTOP analyses, the maximum SG temperature was changed to 364 °F. Since the changes of input values to the RELAP5 for the SGs and pressurizer reflected the characteristics of the SG and pressurizer replacement at FCS, and the changes of the decay-heat level and SG maximum temperature reflected the changes of the operating conditions, the NRC staff determined that the changes did not affect the applicable ranges approved for the RELAP5 code and, therefore, the use of the code remained valid.

3.1.1.2 Assumptions Used in the LTOP Analyses In the OPPD letter dated October 12, 2007, the LTOP analyses used the following assumptions for the applicable mass-and heat-addition transient analyses:

1.

Initial conditions were bounded by TSs or controlled by operating procedures.

The initial conditions relating to RCPs, HPSIs, SDC, pressurizer steam void, and RCS pressure were included in Table 2 in Attachment 3 to the OPPD letter dated October 12, 2007. In the OPPD letter dated March 22, 2008, the licensee identified the following applicable TSs and operating procedures that bound the initial conditions:

The initial conditions regarding RCPs were standard plant operation requirements that existed in procedure OP-2A, Plant Startup. The requirement that only two RCPs were allowed to be operating while the RCS temperature was below 224 °F was specified in Attachment 1, Step 37. The requirement that no more than three RCPs could be running while the temperature was below 500 °F was specified in Precaution Step 19. During plant shutdown, only one RCP was operating in order to minimize pump heat input to the RCS. However, all mass-addition events in the LTOP analyses assumed that three RCPs were in operation. This assumption was conservative, maximizing the pump heat input and resulting in a maximum reactor vessel beltline pressure, and acceptable.

The initial conditions regarding HPSIs were: only the equivalent of two HPSI pumps and three coolant charging pumps (CCPs) could be

operational once the LTOP was enabled; only the equivalent of one HPSI pump and three CCPs were enabled when the RCS temperatures were below 320 °F (indicated RCS temperature); and only the equivalent of three CCPs were enabled when the RCS temperatures were below 270 °F (indicated RCS temperature). The restrictions regarding HPSIs were specified in TS 2.3(3).

The initial conditions regarding the SDC conditions were: the unit could not be put on SDC until the RCS had cooled to 350 °F indicated RCS temperature and 300 psia (pressure at the pressurizer). The initial maximum temperature was specified in TS 2.1.1(11)b. The protection of the SDC system overpressure due to RCS pressure was provided by redundant isolation valves (HCV-347 and HCV-348). Each of the valves was equipped with two redundant interlocks to the pressurizer. The overpressure protection by these interlocks per two scenarios (Reference 4) was to prevent opening of the isolation valves (thereby preventing initiation of SDC) until the RCS pressure was below the SDC entry pressure of 300 psia, and to close the valves if the RCS pressure increased above 300 psia after the SDC system was in operation.

The initial condition regarding the pressurizer steam void was that, when starting the first RCP, there had to be an indicated steam void of 50 percent in the pressurizer. The restriction was specified in TS 2.1.1(11)c.

The initial condition regarding the RCS pressure was that, when starting the first RCP, the RCS pressure should be at least 61 pounds per square inch (psi) below the LTOP setpoint pressure at the given RCS temperature, in order to prevent a PORV lift. Adherence to the maximum pressure to start an RCP avoided PORV opening and reduced the number of activations of a safety system. This guidance was provided as Curve 3 in the licensees TDB III.7a, which was identified in Step 21 of the startup procedure, OP-2A. The calculation of the 61 psi restriction was discussed in Section IX of Reference 3. Specifically, Case 11a showed that the RCS pressure rise following an RCP startup would be 61 psi. The initial conditions used in Case 11a based on a RCS pressure of 350 psia (the maximum initial pressure of 300 psia plus a pressure measurement uncertainty of 50 psi), a maximum SG temperature of 364 °F, a minimum RCS temperature of 50 °F and the decay heat associated with approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (based on a cooldown rate greater than the allowable rate of 100 °F per hour to cool down the RCS from 532 °F to 364 °F). Since the initial conditions assumed in Case 11a represented the limiting operating conditions and were conservative, resulting in a highest pressure rise during the RCP startup event, the NRC staff determined that the calculated value of 61 psi was adequate and acceptable for a margin to the LTOP setpoint pressure.

2.

The failure of one pressurizer PORV to open on demand was identified as the limiting single-failure event. As a result, only one PORV was credited for pressure relief.

3.

No credit was taken for letdown, RCS volume expansion or heat absorption by the RCS metal.

4.

Water-solid conditions in the pressurizer were assumed for mass-addition events and the steam void in the pressurizer was assumed to be bounded by the TS 2.1.1(11)c requirements for heat-addition events.

5.

Full pressurizer (backup and proportional) heater capacity was assumed.

6.

In the OPPD letter dated March 22, 2008, a value of 30.3 megaWatts (MW) was assumed for the decay heat used in the LTOP analyses. This value was determined from the decay heat at the minimum time to reach 314 °F (7848 seconds based on an initial temperature of 532 °F and a maximum cooldown rate of 100°F per hour), which was 21.4 MW based on the 1971 American Nuclear Society (ANS) 5.1 standard with 1997 modifications. The 21.4 MW decay heat was increased by 20 percent to 25.7 megaWatts thermal (MWt) for conservatism, and then increased by the ratio of 1765/1500 to 30.3 MW in anticipation of a potential uprate from 1500 MWt to 1765 MWt. The 30.3 MW was conservative in support of the LTOP analyses based on the full power of 1500 MWt, since the decay-heat-to-full-power ratio of 30/1500 was 2.02 percent at 7848 seconds. By comparison, sample cores in American National Standards Institute (ANSI)/ANS 5.1-1979 showed a power ratio of around 1 percent at this time. The NRC staff also noted that one transient, Case 9, was performed with an RCS temperature of 364 °F to reflect the proposed LTOP enabling temperature of 350 °F with inclusion of the measurement uncertainty of 14 °F. For Case 9, the minimum time to reach 364 °F from 532 °F was 6048 seconds based a maximum cooldown rate of 100 °F per hour. Based on the decay-heat model of the 1971 ANS 5.1 standard with 1997 modifications, the decay power level was 0.16 percent of the full power and the associated decay heat was about 24.5 MWt, which was 5 percent below the value of 25.7 MWt. Therefore, the decay heat of 30.3 MWt remained conservative as it was applied to Case 9.

7.

Mass-addition flow rates were bounded by the combined flow from all available CCPs and HPSI pumps. For conservatism, the HPSI pump-flow rates were increased by 10-percent over the design values in the corresponding pump curve.

The number of operable CCPs and HPSI pumps was specified in the TS 2.3(3) as a function of temperature.

8.

PORV setpoint uncertainty was included. The derivation of the actual LTOP setpoints to open the PORV was discussed in the response to RAI D.4.a of Reference 3. The derivation included consideration of the LTOP analysis curve and the measurement uncertainties for the temperature of 16.3 °F and pressure of 66.9 psi. The results of a Monte Carlo analysis showed that for cases with 1,000 and 100,000 trials, 95 percent and 96.8 percent of the data points were below the LTOP analysis curve, respectively. The derivation of actual LTOP setpoints was documented in a previously NRC-approved LAR (ADAMS Accession No. ML032300305) (Reference 8). Therefore, the approach of

considering the temperature and pressure uncertainties remained valid for the determination of the actual LTOP setpoints.

9.

The PORV flow rate was based on a reduced area of 0.77 square inches (instead of the nominal area of 0.94 square inches). The flow area was determined by using the flow equations in RELAP5 to generate the flow that was equal to the rated flow of 110,220 pounds mass (lbm) per hour at the rated pressure of 2385 psia. The calculated flow rate was a best estimate flow, rather than a conservative flow. The LTOP analysis showed that in all analyzed cases, at the moment that PORV opened, the flow rate was well above the rate required to reduce the peak pressure and the pressure trace showed an immediate decrease in pressure. Based on the results of the LTOP analysis, the NRC staff agreed that even large uncertainties that could exist in the calculated PORV flow rate would not change the peak pressure and, therefore, determined that the PORV flow model was acceptable.

10.

The RCS pressure drop-loss coefficients were developed using design-basis documents. Then a larger pressure drop across the reactor vessel outlet nozzles was added for conservatism. The calculations of the increased pressure drop were consistent with a parallel RELAP5 model that was approved by NRC (ADAMS Accession No. ML032300305). Therefore, the calculations of the RCS pressure drop remained acceptable.

11.

The volumes used in the RCS model were conservatively applied. Specifically, a minimum volume in the SG primary side was assumed when conservative for mass-addition events, and a maximum volume in the SG primary side was assumed when conservative for heat addition events. The minimum volume was based on the design limit of 10 percent SG tube plugging, and the maximum volume was based on the unplugged SG volume. Both assumptions resulted in a higher peak pressure during transients and, therefore, were conservative and acceptable.

12.

The surface area of SGs was modeled as the area associated with zero tube plugging for the heat-addition cases. For mass-addition cases, heat loss to the secondary side was neglected. Both assumptions resulted in a higher peak pressure during transients and, therefore, were conservative and acceptable.

13.

Mass-addition cases included no credit for steam void that existed in the pressurizer. The results of the sensitivity analysis showed that the transient was delayed as the steam volume decreased to zero, and then continued as if the steam void had not existed. The sensitivity study demonstrated that steam void did not affect the margin to the P/T limit for mass-addition cases.

14.

Heat-addition cases were performed with extreme RCS primary-to-secondary temperature differences to maximize the heat-addition effect from SGs.

15.

Heat-addition cases were performed with conservative SG water inventory. The model assumed the entire secondary side was filled with hot water. Since a sufficiently large SG water inventory was assumed, the heat energy stored in the

SG metal mass other than the tube metal, that is, the exterior surface of the SG and non-modeled SG internals, was neglected.

The NRC staff reviewed the assumptions discussed above and determined that they were acceptable based on the findings that: (1) the assumptions were consistent with those used in the LTOP analytical methods previously approved by NRC (ADAMS Accession No. ML010780017) (Reference 7) for Combustion Engineering-manufactured pressurized-water reactors, of which the FCS is one, (2) they were made to assure that the LTOP analyses were conducted in a manner that bounded actual plant operations, and (3) they were conservative, resulting in higher peak pressures during applicable transients.

3.1.1.3 Results of LTOP Analyses The NRC staffs review on the LTOP analyses for mass-and heat-addition transients were based on the information provided in Sections VII and VIII of Attachment 3 to the OPPD letter dated October 12, 2007.

3.1.1.3.1 Mass-addition Transients A mass-addition transient may occur by an operator error or a spurious SI signal that activates the SI pumps to inject water into the RCS. Below the SI shutoff head pressure when the SDC system is isolated, there will be an RCS inventory and pressure increase. Without sufficient pressure relief capacity, the pressure increase caused by this transient may exceed the P/T limits that were developed in accordance with Appendix G requirements of 10 CFR Part 50 at low RCS temperatures.

The licensee performed the analyses for LTOP mass-addition transients using the previously NRC-approved code, RELAP5, with the values of the input parameters representing the SG and pressurizer replacement. The analyses used the conservative assumptions discussed in Section 3.1.2 of this SE. Mass-addition flow rates were the combined flow rates from all available CCPs and HPSI pumps required by the TSs. For conservatism, the HPSI pump flow rates were increased by 10 percent over the design values in the corresponding pump curve.

The density assumed for the SI water was based on water at just above 32 °F and the enthalpy was based on 250 °F water at 90 psia. The assumptions maximized the heat and mass added to the RCS and bounded all possible HPSI water conditions. Decay heat was maximized by assuming the fastest cooldown rate, assuring the shortest cooldown time and the highest decay power. Only one PORV was credited for pressure relief during the transients. No credit was taken for RCS volume expansion, letdown, heat absorption in metal, or steam void in the pressurizer. Three RCPs were assumed in operation to maximize the RCP heat-addition that resulted in a higher pressure increase during the transients. With the reactor at zero power, the primary side SG inventory and the RCS coolant temperatures were assumed to be the same.

The minimum SG volume was assumed to maximize the mass-addition rate as a percentage of total RCS volume. The SG tubes were modeled as being fully isolated so that heat was not lost, which added to the conservatism.

As described in Table 10 of Attachment 3 to the OPPD letter dated October 12, 2007, the licensee analyzed nine cases with assumed initial RCS temperatures varying from 50 °F to 364 °F and the pressure varying from 200 psia to 400 psia, which bounded the SDC entry temperature of 350 °F (indicated at the RCS cold leg), and the pressure of 300 psia (indicated at

the pressurizer). For the analyses of the LTOP mass-addition cases, the licensee relied on opening of the pressurizer PORVs to mitigate the consequences of the events. In the LTOP analyses, the calculated pressurizer pressures were compared to LTOP analytical setpoints (a function of pressure versus cold-leg temperature as shown in Figure 3 of Attachment 3). If the pressurizer pressure exceeded the LTOP analytical setpoints, the PORVs opened and discharged to the quench tank in the containment.

The results of the mass-addition cases were compared to the analytical P/T limits that were previously approved by NRC (ADAMS Accession No. ML032300305) and showed (Table 11 of of the OPPD letter dated October 12, 2007) that there was a large margin to the P/T limits, demonstrating that the proposed LTOP function and the enabling temperature of 350 °F were adequate to protect the P/T limits. Therefore, the NRC staff concluded that the mass-addition transient analyses met the acceptance criteria in SRP Section 5.2.2, and were acceptable.

3.1.1.3.2 Heat-addition Transients A heat-addition transient may occur by an operator error or a spurious signal that starts an RCP.

When the RCS is cooled and the temperature decreases to less than the SG temperature, reverse heat transfer takes place, the pressurizer goes solid, and the RCS pressure increases.

The licensee performed the analyses for the startup of one RCP event, the limiting LTOP heat-addition transient, using the previously NRC-approved code, RELAP5. The analyses used the conservative assumptions discussed in Section 3.1.2 of this SE: a maximum decay heat was assumed; full pressurizer heater capacity was used; no credit was taken for RCS volume expansion, letdown, or heat absorption in metal; and the steam volume in the pressurizer was 40 percent, which was bounded by the TS value of 50 percent.

In the analysis, an extreme RCS primary-to-secondary temperature difference was assumed to maximize the heat-addition effect from SGs: the analyses assumed that the SG secondary side temperature was 364 °F, while the reactor vessel, hot leg, and cold leg were cooled; the initial RCS temperature and pressure were 50 °F and 170 psia, which were the lowest credible temperature and pressure for an RCP startup. The SG temperature of 364 °F represented the limiting operating conditions corresponding to the LTOP enabling temperature of 350 °F plus the temperature measurement uncertainty of 14 °F. A sensitivity study (Table 10 of Reference 3) was performed to assess the effect of the initial temperature and pressure on the pressure response during transients for a temperature range from 50 °F to 364 °F and a pressure range from 170 psia to 390 psia, which bounded the SDC entry temperature of 350 °F (indicated at the RCS cold leg), and the pressure of 300 psia (indicated at the pressurizer). The limiting heat-addition case was mitigated by the existence of a minimum size steam volume (bounded by TS 2.1.1(11)c) in the pressurizer. The steam bubble maintained pressure, and allowed the RCS temperature to reach equilibrium with the SG temperature, thus, terminating the transient.

The results of the analysis for heat-addition cases were compared to the P/T limits that were previously approved by NRC (Reference 8) and showed (Table 12 of Attachment 3 to the OPPD letter dated October 12, 2007) that there was a large margin to the P/T limits, demonstrating that the proposed LTOP function and the enabling temperature of 350 °F were adequate to protect the P/T limits. Therefore, the NRC staff concluded that the mass-addition transient analyses met the acceptance criteria in SRP Section 5.2.2 and were, therefore, acceptable.

3.1.2 Summary The licensee performed the LTOP analysis using the NRC-approved methods. The staff concluded that the values of the input parameters were adequately within the bounding operating conditions and margins subject to the SG and pressurizer replacement. The results of the LTOP analysis were compared to the previously NRC-approved P/T limits and showed that there was a large margin to the P/T limits, demonstrating that the proposed LTOP function and the enabling temperature of 350 °F were adequate to protect the P/T limits. Therefore, the NRC staff concluded that the LTOP analyses met the acceptance criteria in SRP Section 5.2.2, and were acceptable. The NRC concludes that the results of the applicable LTOP or stress analyses, therefore, are acceptable.

3.2 SDC and CCW Systems and Components 3.2.1 Analysis of Proposed Change The staff reviewed the effects of the proposed change on the CCW and SDC systems to determine whether the systems would remain capable of providing adequate cooling water during all operating conditions. As proposed, the SDC system is required to remove a higher amount of RCS decay heat at 350 °F. The licensee evaluated the impacts of the proposed change on the SDC and CCW heat exchangers and determined the SDC and CCW system heat exchangers remain bounded by the current design-basis post-accident heat-load capacities.

There are two independent SDC heat exchangers. The current design heat load for each SDC heat exchanger during SDC operations is 37.1 mega British thermal units (MBtu) per hour. The proposed 50 °F increase in SDC initiation temperature resulted in a maximum of 12 percent increase in SDC heat load. The post-accident design heat-load capability for each SDC heat exchanger is 58.9 MBtu per hour. Hence, the post-accident heat-load capability for SDC heat exchangers bound the expected RCS heat load at 350 °F. The total CCW system heat load during SDC operation is 61.1 MBtu per hour. There are four CCW heat exchangers. The current design heat load for each CCW heat exchanger during normal operations is 23.75 MBtu per hour. Likewise, the proposed 50 °F increase in SDC initiation temperature resulted in a maximum of 12 percent increase in heat load on the CCW system. The post-accident design heat-load capability for each CCW heat exchanger is 117.8 MBtu per hour. Therefore, the post-accident heat-load capability for the CCW heat exchangers bounds the expected increased heat load from the SDC system when the RCS temperature is at 350 °F. The staff concludes that the CCW and SDC heat exchangers continue to operate within their current analyzed limitations under the proposed change; therefore, the CCW and SDC system will continue to satisfy the requirements of FCS Design Criterion 37.

In the October 12, 2007, LAR, the licensee indicated that 10 CFR 50.59 would be used to perform the re-rating of the SDC system piping and components to ensure compliance with Section XI of the ASME Code. On April 23, 2008, the licensee informed the NRC via electronic mail to the NRC Licensing Project Manager that the 10 CFR 50.59 process was completed and reviewed by the Plant Review Committee on September 19, 2007. The documents supporting the re-rating of the SDC system piping and components will be reviewed under the 10 CFR 50.59 regulatory process, and therefore, were not included in the docketed LAR.

As part of the re-rating of the SDC system to increase the temperature and pressure rating to 350 °F and 350 psig (SDC pump suction piping), the licensee states that the revised LPSI pump discharge piping design temperature rating will be 350 °F. However, USAR Section 6.2.3.2 indicated that the current design temperature limit for the LPSI pump seals is 300 °F. No evaluation of the pump seals at the revised temperature was noted and the NRC requested additional information regarding the difference in temperature ratings and whether the pump seals can operate safely at the proposed operating limits. In the OPPD letter dated March 22, 2008, the licensee stated that the pump manufacturer reported that the LPSI pump seals were rated for a temperature of 400 °F and pressure of 600 psig. However, in order to ensure proper seal and bearing cooling, the pump seals cooling water from the CCW system must be kept below 100 °F when the RCS temperature is above 300 °F. The licensee further evaluated the 100 °F limitation on the CCW supply temperature to support cooling of the LPSI pump seals.

The normal operating temperature range for the CCW system is 55 °F to 110 °F. However, the initial LAR stated that the CCW inlet temperature at the seal cooler must not exceed a design limit of 100 °F, when the RCS temperature is above 300 °F. The licensee stated that the 100 °F limitation applies only to the normal operating conditions for the LPSI/SDC seal and bearing cooling. The pump vendor indicated that, when the reactor temperature is below 300 °F, no cooling from the CCW is necessary for the seals and bearings. The NRC staff noted that the CCW system supply temperature was limited to 100 °F and asked the licensee to verify that the CCW system would retain sufficient capability to remove the design decay-heat load with the RCS temperature at 350 °F. In the OPPD letter dated April 4, 2008, the licensee indicated that an evaluation was performed because it was discovered that the assumed seal water flow configuration was misinterpreted by the pump vendor. This resulted in overly conservative cooling water limitations for the CCW inlet temperature to the LPSI pump seal coolers. The licensee indicated that the pump vendor subsequently provided a written verification that a 120 °F CCW temperature limit is acceptable. The licensees response also indicated that the maximum allowed CCW design temperature post-LOCA is 160 °F. Therefore, the post-accident design temperature bounds the analyzed operational 120 °F CCW temperature limit. The staff concludes that CCW and SDC components will continue to operate within their current analyzed limitations and will continue to satisfy the requirements of FCS Design Criterion 1.

The CCW system must be able to provide a sufficient supply of cooling water to the SDC heat exchangers to remove the decay heat in order to cool down the RCS from 350 °F and remain within system design parameters. The licensee performed an evaluation of the heat removal capability of the SDC system in Calculation FC05694, Calculation of Minimum Reactor Coolant Time Using Shutdown Cooling System. The results of the calculation show that with CCW temperature limited to 120 °F and the river water at its projected maximum temperature of 90 °F, the SDC system can cool down the RCS from 350 °F to 130 °F in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one heat exchanger operating at expected service condition fouling factors and remain within required system parameters. A second SDC heat exchanger is available to bring the RCS to 130 °F in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based upon the calculation, the licensee concluded the SDC system, in conjunction with the CCW system, retains the ability to cool down the RCS from 350 °F to 130 °F in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as assumed in the event of a LOCA. The staff concludes that CCW system will continue to satisfy the requirements of FCS Design Criterion 37 in providing sufficient cooling water to engineered safety feature systems in the event of a design-basis accident (DBA).

3.2.2 Summary The licensee proposes to change its TS to increase the allowable maximum temperature for operation of the SDC system from 300 °F to 350 °F. The staff reviewed the licensees assessment of the impacts that the proposed TS change will have on the SDC and CCW systems and components important to safety. Based upon its review, the staff finds that the proposed change will not compromise the design limits or functional capabilities of systems and components important to safety and there is reasonable assurance that the regulatory criteria and the plant licensing basis in the USAR will continue to be satisfied following implementation of the proposed change. Therefore, the staff concludes that the proposed change to TS LCO 2.1.1 and TS SR 3.16 are acceptable.

3.3 Structural and Pressure Boundary Integrity This section addresses the effects of the FCS SDC Entry Condition Uprate LAR on the structural and pressure boundary integrity of the SDC piping system and components including their associated supports.

The proposed SDC Entry Condition Uprate will increase the SDC system entry temperature from 300 °F to 350 °F as measured at the cold leg of the RCS. The uprate will also increase the SDC system entry pressure from 250 psia to 300 psia, indicated at the pressurizer. In the OPPD letter dated October 12, 2007, Table 1 of Section 3.2.1 shows the pertinent design pressures and temperatures for the SDC system piping and associated components for the current and proposed (rerated) conditions.

3.3.1 SDC Piping and Supports The SDC system piping was designed in accordance with the USAS B31.7, 1968 Edition, Code for Pressure Piping. The licensee reviewed the revised design conditions for impact on the existing design basis analyses for the SDC system piping and supports. Specifically, the licensee indicated that its evaluation included analysis of support loads, valve accelerations, pump and heat exchanger nozzle loads, and pipe movements at wall penetration bellows. In the OPPD letter dated March 22, 2008, the licensee provided a quantitative summary of the piping analysis that was performed to demonstrate compliance with the ASME Code requirements.

This summary provided the allowable and maximum stresses along with the corresponding stress ratios produced from the analysis of the SDC piping and supports. The results of this summary indicate that the rerated SDC entry conditions proposed by OPPD for FCS are within ASME Code-allowable values for the piping and associated supports. It is noted that piping support SIH-287 will be removed in order to meet the ASME Code-allowable values for the SDC piping system. Based on the above, the NRC staff concludes that the licensees assessment that the SDC piping and piping supports provides sufficient assurance that operation at the rerated SDC entry conditions is acceptable.

3.3.2 SDC Pump Hold-Down Bolting Stresses Due to the effects of increased nozzle loads as determined in the piping analysis noted in the previous section, the licensee indicated that stronger hold-down bolts for the LPSI pumps would be necessary in order to meet ASME Code-allowable stress limits for the existing pump hold-down bolts at the rerated conditions. In response to NRC staff RAIs, the licensee provided

1) a quantitative summary of the analysis supporting the need for a stronger bolting material, and

2) justification for the reconciliation of the code of reference used to evaluate the LPSI pump hold-down bolts.

3.3.2.1 Code Reconciliation The licensee indicated that the following reconciliation of the code of reference used for the LPSI pump hold-down bolting at the rerated conditions was accepted by the Authorized Nuclear Inspector/Authorized Nuclear Inservice Inspector (ANI/ANII) at FCS. The original codes of analysis for the components under evaluation were ASME Section III, 1965 Edition with Addenda through Winter 1966; ASME Code,Section VIII, 1965 Edition with Addenda through Winter 1966; USAS B31.1 Power Piping Code, 1967 Edition; and American Standards Association (ASA) B16.5, 1961 Edition. For the pressure and temperature rerate commensurate with this LAR, the code of analysis used for the component evaluation is ASME Code,Section VIII, Division 1, 1992 Edition and ANSI B16.5, 1996 Edition. The original code of analysis for the materials under evaluation was USAS B31.1, 1967 Edition. For the purposes of the rerated conditions, ASME Code,Section III, 1989 Edition Class I requirements were used for the materials evaluation.

3.3.2.2 Analysis The results of the LPSI pump analysis summary (Appendix B, Reference 4) indicate that the hold-down bolts for the LPSI pumps at the rerated conditions are the critical components when analyzed in accordance with the ASME Code-allowable stress limits. The analysis concludes that due to the inability of the current bolting of the LPSI pump to meet the applicable stress limits, a newer, stronger material is recommended for the FCS LPSI pump hold-down bolts. In the OPPD letter dated March 22, 2008, the licensee indicates that the hold-down bolting modification will take place during the 2008 refueling outage as recommended by the aforementioned analysis. Based on the above, the NRC staff concludes that the licensees assessment that the LPSI pump hold-down bolting modifications provides sufficient assurance of the acceptability of operation at the rerated SDC entry conditions is acceptable.

In the OPPD letter dated March 22, 2008, the licensee responded to the staffs RAI regarding the maximum calculated stresses, fatigue Usage Factors, and Code-allowable values. Based on the licensees response stating that the Cumulative Usage Factors was less than 1.0 and the fact that the SDC system is not subject to FCS temperatures and pressures, the staff finds the proposed changes to the USAR-described methodology acceptable, as applied to the SDC heat exchangers. The staff approves the proposed change from ASME Code,Section III, Class A to Class C.

3.3.3 SDC System HELB Analysis In the OPPD letter dated March 22, 2008, the licensee stated that a pipe-break review was not performed (RAI B.2. of Reference 3). The licensee indicated that the SDC system is a fluid system in which the fluid operates at high-energy conditions less than 2 percent of the time and that it is, therefore, considered a moderate energy system, and no HELBs are postulated in the current design basis (USAR, Appendix M and PLDBD-ME-11). The NRC staff review concluded that this was consistent with Branch Technical Position (BTP) 3-4 (ADAMS Accession No. ML070800008). Consequently, no HELBs are postulated in the current design basis for the

facility. Based on the above, the NRC staff concludes that no HELB analysis is necessary for the SDC system entry condition rerate.

3.3.4 Summary The NRC staff has reviewed OPPDs assessment of the impact of the proposed SDC Entry Condition Uprate on SDC piping, supports, and components with regard to stresses, CUFs, and HELB requirements. On the basis of this review described above, the NRC staff concludes that the proposed SDC Entry Condition Uprate will not have an adverse impact on the structural integrity of the SDC piping systems, supports, and associated components.

3.4 Proposed TS Changes

3.4.1 Revision to TS LCO 2.1.1(2)

Current TS LCO 2.1.1(2) states:

Hot Shutdown or 300 °F Tcold [cold-leg temperature] 515 °F The proposed TS LCO 2.1.1(2) would state:

Hot Shutdown or 350 °F Tcold 515 °F 3.4.2 Revision to TS LCO 2.1.1(3)

Current TS LCO 2.1.1(3) states:

210 °F Tcold < 300 °F or Tcold < 210 °F with fuel in the reactor and all reactor vessel head closure bolts fully tensioned.

The proposed TS LCO 2.1.1(3) would state:

210 °F Tcold 350 °F or Tcold < 210 °F with fuel in the reactor and all reactor vessel head closure bolts fully tensioned.

3.4.3 Revision to TS LCO 2.1.1(11)(b)

The current TS 2.1.1(11)(b) states:

The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 300 °F.

The proposed TS LCO 2.1.1(11)(b) would state:

The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 350 °F.

The acceptability of the proposed changes to the temperature limits for the reactor coolant loops and decay heat-removal loops, and the LTOP enabling temperature was demonstrated by the

LTOP analysis and the stress analysis discussed in Sections 3.1.1 and 3.2.1 of this SE.

Therefore, the NRC staff determined that the proposed TSs 2.1.1(2), 2.1.1(3), and 2.1.1(11)b discussed above were acceptable.

3.4.4 Revision to TS SR 3.16(1)a.

Current TS SR 3.16(1)a. states:

The portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 250 psig. This shall be performed on a refueling frequency.

The proposed TS SR 3.16(1)a. would state:

The portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 300 psig. This shall be performed on a refueling frequency.

The basis for TS 3.16(1)a indicated (Reference 4) that the limiting leakage to the atmosphere from the SDC system was based on a plant-specific leak rate analysis for the SDC system components operating after a DBA. Since the SDC piping uprate had no effect on the DBA conditions (supported by the stress analysis discussed in Section 3.2 of this SE) and the required leakage limit of the 3800 cubic centimeters per hour was not changed, the test pressure of 300 psig increased from 250 psig was a more restrictive TS requirement. In addition, the proposed test pressure of 300 psig was within the uprated design pressure (350 psig) of the SDC suction piping that was supported by the stress analysis discussed in Section 3.2 of this SE. This approach was consistent with that used in the existing TS, which required that the leakage test be performed at 250 psig, which was within the existing design pressure of 300 psig for the SDC suction piping. Therefore, the NRC staff determined that the proposed TS 3.16(1)a was acceptable.

3.4.5 Revision to TS SR 3.16(1)d.

Current TS SR 3.16(1)d. states:

An internal leakage test shall be performed on a refueling frequence.

The proposed TS SR 3.16(1)d. would state:

An internal leakage test shall be performed on a refueling frequency.

The new TS SR 3.16(1)d. corrects a typographical error by replacing the misspelled word frequence with frequency. The NRC staff determined that the proposed TS 3.16(1) was acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component 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. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on November 20, 2007 (72 FR 65370). 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.

7.0 REFERENCES

1.

Letter dated October 12, 2007, from D. J. Bannister, Acting Site Director, Omaha Public Power District (OPPD), to NRC,

Subject:

Fort Calhoun Station Unit No. 1, License Amendment Request (LAR), Uprate of Shutdown Cooling (SDC) System Entry Conditions (ADAMS Accession No. ML072890192).

2.

Letter dated February 27, 2008, from Michael T. Markley, NRC, to David J.

Bannister, Vice President and CEO [Chief Executive Officer], OPPD,

Subject:

Fort Calhoun Station, Unit No. 1, - Request for Additional Information Re: License Amendment Request, Uprate of Shutdown Cooling System Entry Conditions (MD6993) (ADAMS Accession No. ML080560007).

3.

Letter dated March 22, 2008, from R. P. Clemens, Division Manager, OPPD, to NRC,

Subject:

Response to Request for Additional Information Regarding License Amendment Request, Uprate of Shutdown Cooling System Entry Conditions (TAC NO. MD6993) (ADAMS Accession No. ML080850254).

4.

Letter dated April 4, 2008, from Richard P. Clemens, Division Manager - Nuclear Engineering, OPPD, to NRC,

Subject:

Transmittal of Revised Response to Request for Additional Information Re: License Amendment Request, Uprate of

Shutdown Cooling [SDC] System Entry Conditions (TAC No. MD6993) and Calculation FC05694 (ADAMS Accession No. ML081070336).

5.

Memorandum dated September 18, 1992, from Samuel J. Chilk, Secretary, NRC, for James M. Taylor, Executive Director for Operations,

Subject:

Resolution of Deviations Identified during the Systematic Evaluation Program (ADAMS Accession No. ML003763736).

6.

U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, March 2007 (ADAMS Package Accession No. ML070660036).

7.

Letter dated March 16, 2001, from S. A. Richards, to Richard Bernier, Chairman, CE Owners Group,

Subject:

Safety Evaluation of Topical Report CE NPSD-683, Revision 6, Development of a RCS Pressure and Temperature Limits Report

[PTLR] for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications (TAC NO. MA9561) (ADAMS Accession No. ML010780017).

8.

Letter dated August 15, 2003, from A. B. Wang, NRC, to R.T. Ridenoure, Division Manager - Nuclear Operations, OPPD,

Subject:

Fort Calhoun Station, Unit 1 -

Issuance of Amendment [No. 221] (TAC NO. MB6468) (ADAMS Accession No. ML032300305).

Principal Contributors: S. Sun W. Jessup A. Tsirigotis S. Gardocki Date: May 9, 2008