ML023100365

Issuance of Amendment 196 to License DPR-23 1.7-Percent Power Uprate

ML023100365
Person / Time
Site:	Robinson
Issue date:	11/05/2002
From:	Subbaratnam R NRC/NRR/DLPM/LPD2
To:	Moyer J Carolina Power & Light Co
Subbaratnam R, NRR/DLPM, 415-1478
References
-nr, TAC MB5206
Download: ML023100365 (52)

Text

Mr. J. W. Moyer, Vice President November 5, 2002 Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 (HBRSEP2) -

ISSUANCE OF AMENDMENT REGARDING A 1.7-PERCENT POWER UPRATE (TAC NO. MB5106)

Dear Mr. Moyer:

The Commission has issued the enclosed Amendment No. 196 to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2. This amendment consists of changes to the Facility Operating License and Technical Specifications (TS) in response to your application dated May 16, 2002, as supplemented by letters dated July 25, August 12, September 6, October 15, and October 31, 2002.

This amendment approves revision to the Facility Operating License and TS to reflect an increase in the HBRSEP2 maximum steady-state core power level from 2300 megawatts thermal (MWt) to 2339 MWt, an increase of approximately 1.7 percent. This increase is facilitated by the utilization of the Caldon Leading Edge Flowmeter for feedwater flow measurements.

The NRC staff authorizes the amendment subject to an additional condition in Appendix B that limits operation of HBRSEP2 to 504 effective full-power days.

A copy of the Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions bi-weekly Federal Register notice.

Sincerely,

/RA/

Ram Subbaratnam, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 196 to DPR-23

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

Mr. J. W. Moyer, Vice President November 5, 2002 Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 (HBRSEP2) -

ISSUANCE OF AMENDMENT REGARDING A 1.7-PERCENT POWER UPRATE (TAC NO. MB5106)

Dear Mr. Moyer:

The Commission has issued the enclosed Amendment No. 196 to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2. This amendment consists of changes to the Facility Operating License and Technical Specifications (TS) in response to your application dated May 16, 2002, as supplemented by letters dated July 25, August 12, September 6, October 15, and October 31, 2002.

This amendment approves revision to the Facility Operating License and TS to reflect an increase in the HBRSEP2 maximum steady-state core power level from 2300 megawatts thermal (MWt) to 2339 MWt, an increase of approximately 1.7 percent. This increase is facilitated by the utilization of the Caldon Leading Edge Flowmeter for feedwater flow measurements.

The NRC staff authorizes the amendment subject to an additional condition in Appendix B that limits operation of HBRSEP2 to 504 effective full-power days.

A copy of the Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions bi-weekly Federal Register notice.

Sincerely,

/RA/

Ram Subbaratnam, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 196 to DPR-23

2. Safety Evaluation cc w/encls: See next page ADAMS Letter Accession Number ML023100365 Distribution: See attached list

• See previous concurrence OFFICE PM:PDII-S2 LA:PDII-S2 DSSA:SPSB DE:EEIB* DSSA:SRXB* DE:EMCB* DE:EMCB*

NAME RSubbaratnam EDunnington MReinhart ASGill FAkstulewicz LLund SCoffin DATE 11/04/2002 11/04/2002 9/12/2002 8/23/2002 10 / 22/02 9/ 16/2002 10/7/2002 OFFICE DE:EMEB* SPLB* IEHB* OGC SC:PDII-2 D:PDII-S2 DD/D:DLPM NAME KManoly SWeerakkodi DTrimble RWeisman AHowe HBerkow TMarsh/JZwolinski DATE 9/28/02 10/24/2002 10/24/2002 11/04/2002 11/04/2002 11/04/2002 11/05/2002 C:\ORPCheckout\FileNET\ML023100365.wpd OFFICIAL RECORD COPY

AMENDMENT NO. 196 TO FACILITY OPERATING LICENSE NO. DPR H. B. Robinson, UNIT 2 DISTRIBUTION:

PUBLIC PDII-2 Reading File OGC H. Berkow, DLPM/NRR A. Howe, DLPM J. Hayes, SPSB, DSSA L. Lund, EMCB, NRR C. Holden, DE, NRR F. Akstulewicz, NRR M. Mitchell, NRR C. Luaren, NRR B. Fu, NRR B. Marcus, NRR S. Sun, NRR M. Mitchell, EMCB, DE K. Manoly, EMEB, DE D. Trimble, IEHB, DIPM G. Hill (2)

R. Subbaratnam M. McConnell E. Dunnington ACRS S. Cahill, RII

CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 196 License No. DPR-23

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

A. The application for amendment by Carolina Power & Light Company (the licensee) dated May 16, 2002, as supplemented by letters dated July 25, August 12, September 6, October 15, and October 31, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions regulations; D. The issuance of this 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 Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to paragraph 3.A. of Facility Operating License No. DPR-23, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(1) Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level of 2339 megawatts thermal.

3. The license is also amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-23 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, are hereby incorporated in the license. Carolina Power &

Light Company shall operate the facility in accordance with the Technical Specifications.

4. The license is also amended by the inclusion of an additional condition in Appendix B that limits operation of H. B. Robinson Steam Electric Plant, Unit No. 2, to 504 effective full-power days.

5. This license amendment is effective as of the date of its issuance and shall be implemented within 45 days of startup from Refueling Outage 21.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

John A. Zwolinski, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the License, Technical Specifications, and Appendix B Date of Issuance: November 5, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 196 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Page 3 Page 3 Page 4d Page 4d Replace the following pages of Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-4 1.1-4 2.0-2 2.0-2 3.3-18 3.3-18 3.3-19 3.3-19 3.3-25 3.3-25 3.4-7 3.4-7 3.4-8 3.4-8 3.7-1 3.7-1 3.7-9 3.7-9 Replace the current Appendix B, Additional Conditions with the attached revised Appendix B (one page). The revised page is identified by amendment number and contains a marginal line indicating the area of change.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT TO INCREASE THE AUTHORIZED POWER LEVEL OF H. B. ROBINSON, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR),

Part 50, Section 90, Carolina Power & Light Company (CP&L), by letter dated May 16, 2002, as supplemented by letters dated July 25, August 12, September 6, October 15, and October 31, 2002, submitted a request for an amendment to the Facility Operating License (FOL), including the Appendix A Technical Specifications (TS) and Appendix B Additional Conditions for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). The proposed amendment would increase the authorized reactor core power level from 2300 megawatts thermal (MWt) to 2339 MWt (approximately 1.7 percent). This is considered as a measurement uncertainty recapture power uprate.

This power uprate is facilitated by using the Caldon, Inc. (Caldon), Leading Edge FlowmeterUTM (LEFMUTM) and LEFM CheckPlusTM systems to measure feedwater flow at HBRSEP2.

The July 25, August 12, September 6, October 15, and October 31, 2002, supplements contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.

2.0 BACKGROUND

Nuclear power plants are licensed to operate at a specified core thermal power. Prior to June 2000, Appendix K to 10 CFR Part 50 required loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) analyses to assume that the reactor has been operating continuously at a power level of at least 102 percent of the licensed thermal power to allow for uncertainties. As documented in the Federal Register on June 1, 2000 (65 FR 34916),

at the time of the original ECCS rulemaking, the 2-percent power margin requirement was solely based on the considerations associated with reactor power measurement uncertainty.

In order to reduce unnecessary regulatory burden and to avoid unnecessary exemption requests, the Commission revised Appendix K on June 1, 2000, to allow licensees the option of justifying a smaller margin of power measurement uncertainty by using more accurate instrumentation to calculate the reactor thermal power. The final rule, by itself, did not allow licensees to increase the licensed power level without U.S. Nuclear Regulatory Commission (NRC) staff approval. The proposals to increase the licensed power level must be reviewed and approved through the license amendment process.

3.0 EVALUATION 3.1 Reactor Systems 3.1.1 Regulatory Evaluation The uncertainty of the calculated values of this thermal power determines the probability of exceeding the power levels assumed in the design-basis transient and accident analyses.

Appendix K to 10 CFR 50 requires the licensees to base their LOCA analysis on an assumed power level of at least 102 percent of the licensed thermal power level (see section 2.0 above).

This required power ratio is to allow for uncertainties in determining thermal power. The NRC concluded that, at the time of original ECCS rulemaking, the 2-percent power margin requirement was based solely on consideration associated with power measurement uncertainty as is reflected in Appendix K. Appendix K did not require a demonstration of the power measurement uncertainty and mandated a 2-percent margin, notwithstanding that the instruments may be more accurate than originally assumed in the ECCS rulemaking. On June 1, 2000, the NRC published a final rule (65 FR 34913) that allows licensees to justify a smaller margin for power measurement uncertainty (see section 2.0 above). Licensees may apply the reduced margin to operate the plant at a level higher than the previously licensed power.

The NRC staff review is to verify that the licensees analytical results meet the required acceptance criteria, and to ensure that the proposed TS appropriately reflect the results of acceptable safety analyses. The following evaluation is based on the staff review of the licensees safety analyses, proposed TS changes (References (Refs.) 1 and 2) and the responses to the staffs requests for additional information (RAIs) (Refs. 3, 4, and 5). This review includes the following areas: (1) power measurement uncertainty, (2) non-LOCA and LOCA transients analyses, (3) anticipated transient without scram (ATWS) analysis, (4) mid-deck plate-induced steam generator (SG) water level uncertainties, and (5) proposed TS changes.

3.1.2 Power Measurement Uncertainty The licensees submittals discuss the instrument uncertainties to support the proposed power calorimetric measurement uncertainty of 0.3 percent of the rated power. The daily power is measured based on the SG thermal output. Assuming that the reactor coolant system (RCS) primary and secondary sides are in equilibrium, the core power is determined by summing the thermal output of the SGs and the RCS primary side heat loss, and subtracting the reactor coolant pump (RCP) heat addition. The SG thermal output is determined by secondary side calorimetric measurement, which is determined by the feedwater flow multiplied by the difference in the steam and feedwater enthalpy, with the correction of SG blowdown. The feedwater flow is measured using the Caldon LEFM Check Plus system placed in the feedwater header for HBRSEP2.

The uncertainty calculations for the secondary side power calorimetric measurement, provided in Table 3.2-1 of Ref. 1, consider uncertainties for power measurement parameters including the feedwater flow, feedwater temperature, main steam pressure, SG blowdown flow, and pump heat addition. The methodology used to combine the uncertainties for power

measurement parameters is the square-root-of-sum-of-the-squares (SRSS) of those groups of parameters that are statistically independent. The uncertainties used are considered to be random, two-sided distributions.

Based on its review, the NRC staff finds that the uncertainties associated with feedwater temperature and flow, main steam pressure, SG blowdown flow, and various heat losses and gains are adequately considered in the calculation of the total uncertainty of the power measurement, and that the proposed value of 0.3 percent of the rated thermal power (RTP) bounds the calculated total uncertainty of the power measurement. Therefore, the NRC staff concludes that the proposed power measurement uncertainty of 0.3 percent of RTP is acceptable.

3.1.3 Non-LOCA and LOCA Transients Analyses The licensee discussed the Updated Final Safety Analysis Report (UFSAR) Chapter 15 transients and LOCA analyses in Ref. 1 for the power uprate conditions. The licensee identified the limiting cases for each event category discussed in UFSAR Chapter 15 and evaluated the effects of power uprate on plant transients and accidents. For those cases that were bounded by the corresponding cases in UFSAR Chapter 15, the licensee provided supporting rationales.

The majority of them were analyzed at the thermal power level of 2346 MWt, which bounds the current request at 2339 MWt. For those cases with values of plant parameters outside the applicable range of the corresponding UFSAR cases, the licensee provided results of reanalyses to show compliance with applicable acceptance criteria used in the corresponding UFSAR analysis. The licensee considered the following plant conditions (Ref. 1):

. A maximum core power of 2339 MWt (increased from the current core power of 2300 MWt;)

. A full-power normal average temperature (Tavg) of 575.9 oF (increased from 575.4 oF);

and

. A steam pressure decrease of 2.8 psig and a steam mass flow rate increase of 1.7 percent.

Based on sensitivities for the high thermal performance (HTP) departure from nucleate boiling (DNB) correlation, the licensee estimated that the effect of a slight increase in the reactor vessel average temperature due to the power uprate will be a decrease in DNB ratio (DNBR) of about 1 to 2 percent. The licensee indicated that the additional margin afforded by application of the approved statistical DNBR methodology is much larger than the estimated DNBR decrease for the power uprate. The NRC staff reviewed the licensees safety assessment (Ref. 1) and the RAI responses (Refs. 3 and 4) and found that the current UFSAR analysis for the design-basis transients and accidents remains conservative. Therefore, the NRC staff concludes that it remains valid and acceptable for the uprated power. A summary of the NRC staff evaluation is shown in Table 1. The table includes a description of the events analyzed, the impacted UFSAR sections, and a disposition for each of those events. The NRC staff found all of these transients acceptable for the uprated conditions.

3.1.4 ATWS Analysis An ATWS event is defined as an anticipated operational occurrence (such as loss of normal feedwater, loss of load, or loss of offsite power) combined with an assumed failure of the reactor trip to shut down the reactor. For the pressurized-water reactors (PWRs) manufactured by Westinghouse, the basic requirements of the ATWS rule are specified in 10 CFR 50.62(c)(1). The licensee satisfies the ATWS rule (Ref. 1) by installing NRC-approved ATWS Mitigating System Actuation Circuitry (AMSAC). The NRC staff requested the licensee to provide a discussion of an ATWS analysis demonstrating that the operation of the plant at the proposed uprated power level is within the bounds considered by the NRC staff during the licensees documentation of compliance with the ATWS rule.

In its response dated July 25, 2002 (Ref. 3), the licensee indicated that HBRSEP2 currently relies upon the generic ATWS analyses to demonstrate the acceptability of the analytical results. The generic analyses were performed by Westinghouse and documented in WCAP-8404, Anticipated Transient Without Trip Analysis for Westinghouse PWRs with 44 Series Steam Generators. These analyses include 2-, 3-, and 4-loop PWRs with various SG models. The base case of the generic ATWS analyses for a Westinghouse 3-loop PWR and a power level of 2300 MWt adequately represents the current plant configurations and licensed power level; therefore, it is applicable to HBRSEP2. The sensitivity study (WCAP-8404 and Ref. 3) of the generic ATWS analyses was performed to determine the changes of the calculated peak RCS pressure resulting from the changes in various initial plant conditions such as initial power level, power-operated relief valves (PORVs) relief capacity, and auxiliary feedwater (AFW) capacity. The results of the sensitivity study showed that a 39 MWt increase (2 percent from the base case) causes a 26 psi increase in the calculated peak pressure, resulting in a peak RCS pressure of approximately 2913 psia.

The sensitivity study also showed that an increase in the pressurizer PORV capacity from the value of 358,000 lbm/hr used in the ATWS base case for a Westinghouse 3-loop PWR to a value of 420,000 lbm/hr produces a 63 psi decrease in the peak RCS pressure. Since HBRSEP2 has a total pressurizer PORV capacity of 420,000 lbm/hr, the 26 psi increase in peak RCS pressure caused by the 2-percent power increase is more than offset by the 63 psi reduction due to the larger HBRSEP2 PORVs. In addition, the licensee confirmed (Ref. 3) that the AFW flowrate and pressurizer safety valve (PSV) capacity at HBRSEP2 are greater than the total AFW flowrate and PSV relief flow assumed in the ATWS analysis base case for Westinghouse 3-loop PWRs. The licensee also verified (Ref. 3) that the moderator reactivity feedback for HBRSEP2 remains sufficiently negative to be comparable to the generic Westinghouse analyses presented in WCAP-8404. Since the pressurizer PORV capacity is greater than that assumed in the generic analyses (offsets the increase in the core power level), and the moderator temperature coefficient (MTC), PSV capacity, and AFW flowrate are within the applicable ranges of the ATWS base case, the NRC staff concludes that the generic analysis for a Westinghouse 3-loop PWR remains bounding and acceptable for the HBRSEP2 uprated power.

3.1.5 Mid-Deck Plate-Induced SG Water Level Uncertainties Westinghouse issued three Nuclear Service Advisory Letters (NSALs), NSAL-02-3 and revision 1, NSAL-02-4, and NSAL-02-5, to document potential problems with the Westinghouse-designed SG water level setpoint uncertainties. NSAL-02-3 and its revision, issued on February 15 and April 8, 2002, respectively, deal with uncertainties caused by the

mid-deck plate located between the upper and lower taps used for SG measurements. These uncertainties affect the SG low-low level trip setpoint (used in the analyses for events such as the feedwater line break, ATWS, and steamline break). NSAL-02-4, issued on February 19, 2002, deals with uncertainties created because the void contents of the two-phase mixture above the mid-deck plate were not reflected in the calculation and these affect the SG high-high level trip setpoint. NSAL-02-5, issued on February 19, 2002, deals with the initial conditions assumed in the SG water level-related safety analyses. The analyses may not be bounding because of velocity head effects or mid-deck plate pressure differential pressure, which have resulted in significant increases in the control system uncertainties. The NRC staff requested the licensee to discuss how HBRSEP2 accounts for the uncertainties documented in these advisory letters in determining the SG water level setpoints.

The licensee indicated (Ref. 3) that the effects identified in NSAL-02-3 were incorporated into the Total Loop Uncertainty (TLU) calculation for SG low-low water level trip setpoint under normal containment conditions (i.e., containment pressure and temperature within the limits specified in the HBRSEP2 TS limiting condition for operation (LCO) 3.6.4 and LCO 3.6.5).

Based on the results of the calculation, the licensee indicated that the SG low-low water level reactor trip allowable value and the normal trip setpoint, as listed in the HBRSEP2 TS LCO 3.3.1, Function 13 remain valid due to sufficient margin in the TLU calculation to accommodate the condition of concern in NSAL-02-3.

In Refs. 3 and 4, the licensee further indicated that effects identified in NSAL-02-4 were incorporated into the TLU calculation for the SG high-high water level control function. The calculation indicated that the SG high-high water level feedwater isolation setpoint should be reduced from 75 percent to 74 percent. The SG high-high water feedwater isolation function is to prevent SG overfill in the event of a failure in the SG level control system. This function is not used for mitigation of any design-basis accidents (DBAs) in Chapter 15 of the HBRSEP2 UFSAR. Therefore, the UFSAR analysis remains valid and acceptable for the power uprate operation.

In addition, the licensee indicated (Refs. 3 and 4) that the SG water level issues identified in NSAL-02-3 and Revision 1, NSAL-02-4, and NSAL-02-5 were evaluated in accordance with the Progress Energy and CP&L Nuclear Generation Group procedures that establish the methods for evaluation of operating experience of this type (i.e., vender technical information). The licensees evaluation as discussed above indicated that (1) no plant changes are required for the power uprate, and (2) the UFSAR Chapter 15 analysis is unaffected by the SG water level issues identified in the three NSAL letters and thus remains valid. Based on the results of the licensees evaluation, the NRC staff concludes that the SG water level issues are adequately addressed for the uprated power.

3.1.6 Reactor System Summary The NRC staff has reviewed the licensees safety analyses and the associated TS changes in support of operation of HBRSEP2 at a maximum core power level of 2339 MWt. For the reasons set forth above, the NRC staff finds that the supporting safety analyses show that the uprated power conditions are bounded by the current UFSAR analysis. Therefore, the NRC staff concludes that the current UFSAR analyses for the design-basis transients and accidents remain valid and acceptable for the uprated power. The NRC staff also finds that the proposed TS discussed in Section 3.11 of this evaluation adequately reflect the results of the acceptable safety analyses and therefore concludes that the proposed TS are acceptable.

3.2 Electrical Systems 3.2.1 Regulatory Evaluation Robinsons original licensing basis for emergency power at HBRSEP2 is found in UFSAR Chapter 3 and follows proposed General Design Criterion (GDC) 39, Emergency Power for Engineered Safety Features (ESF). This is one of the General Design Criteria proposed by the Atomic Energy Commission in a proposed rulemaking published in the Federal Register on July 11, 1967, and was used to evaluate the adequacy of the electric power systems.

Proposed GDC 39 provided that sufficient offsite and redundant, independent, and testable standby auxiliary sources of electrical power are available to attain a prompt shutdown and continued maintenance of the plant in a safe condition under all credible circumstances.

Section 50.63 of 10 CFR, Station Blackout, requires that all nuclear power plants must have the capability to withstand a loss of all ac power for an established period of time, and to recover therefrom.

Section 50.49 of 10 CFR, Environmental Qualification of Electric Equipment important to Safety for Nuclear Power Plants, requires licensees to establish programs to qualify electric equipment important to safety. Under the rules, each licensee must (1) prepare and maintain a record of qualification to document that each item of equipment subject to the rule is qualified for its application, and (2) meets its specified performance requirements when subjected to the environmental conditions predicted to be present when it must perform its safety function up to the end of qualified life.

3.2.2 Electrical and Instrument & Controls (I&C) Systems Technical Evaluation The NRC staff reviewed the licensees regulatory and technical analysis in support of its proposed license amendment described in Sections 3.9, 3.10.7, and 4.3 of the licensees submittal (Ref. 1).

3.2.2.1 Grid Stability The licensee performed a grid stability analysis to support the proposed power uprate.

The grid analysis was performed using a bounded generator output value of 810 MWe gross, which is in excess of the maximum expected post-uprate generator output. The analysis included simulated disturbances (a 3-phase fault with delayed clearing) that exceeded those normally specified by the North American Electric Reliability Council Planning Standards. The results of the grid stability analysis indicate that there is no adverse effect on grid stability, and that the power uprate will not adversely impact the availability of offsite power for the unit auxiliary loads in the event of a unit trip.

The NRC staff reviewed the licensees information and concluded that there is reasonable assurance that proposed GDC 39 for grid stability will be met at this power uprate condition.

3.2.2.2 Main Generator The main generator is currently rated to produce 854.1 MVA (769 MWe) at 22 kV, when operating at 60 Hz, with 75 psig hydrogen pressure, and a 0.90 power factor. The increase in

gross generating capacity due to the power uprate is approximately 12.4 MWe (summer) and 13.0 MWe (winter). Main generator gross power capability was evaluated for power uprated conditions at 780 MWe (summer) and 800 MWe (winter), with a maximum gross generator output of 800 MWe and 250 MVAR. This output corresponds to approximately 838 MVA at a 0.954 pf. This value is within the main generator nameplate rating, and is within the generator reactive capability curve at 68 psig generator hydrogen pressure. An evaluation of main generator protective relay settings did not identify any required changes to equipment protective relay settings.

The generator exciter has a nameplate rating of 4700 kW, 550 V, 8545 amps. An evaluation of excitation requirements for operation at power uprate indicates that the required excitation needed is 4240 kW, 525 V, 7355 amps, which is well within the generator exciter capability.

Based on its review, the NRC staff concludes that the net increase in the power uprate condition is within the generator reactive capability curve. Therefore, the NRC staff has reasonable assurance that the main generator can operate safely at the power uprate condition.

3.2.2.3 Main Power Transformers and Switchyard The main power transformer bank consists of three single-phase units, each rated at 260 MVA for a total capacity of 780 MVA. The licensee performed two studies for the main components of the power conversion system, including the main transformer and the electrical equipment in the switchyard. The licensee evaluated the transmission parameters for the power uprate and assessed its effect on protective relaying. The studies showed that the main power transformer is adequately sized and no changes to the protective relaying are required. Therefore, the NRC staff has reasonable assurance that the main power transformer can operate safely at the power uprate condition.

3.2.2.4 Auxiliary Power System The auxiliary power system includes the 4160 V, 480 V, vital 120 Vac, and 125 Vdc electrical systems and the startup and unit auxiliary transformers. The main feedwater pumps and condensate pumps will operate with slightly increased flow under the power uprate conditions, which will increase their motor brake horsepower requirements. The motors for these pumps were evaluated and it was determined that the main feedwater pump motors will remain within their nameplate rating. The motor for the condensate pumps will operate above the nameplate motor rating under the power uprate condition, but will remain within the 1.15 service factor rating; thus, there will be no adverse impact as a result of the power uprate. The heater drain pumps and service water pumps will have slightly increased flow due to the proposed power uprate. The motors for these pumps have been evaluated, and it has been determined that the pump motors will remain within their nameplate ratings. Loads on the 120 Vac and 125 Vdc systems are not changed with the power uprate.

The NRC staff reviewed the licensees submittal and concluded that all the pump motors are adequately sized, and there is reasonable assurance that the auxiliary power system can operate safely at the power uprate condition.

3.2.2.5 Isolated Phase Duct The isolated phase bus is rated at 25,000 amps. The loadings on the isolated phase bus are not expected to exceed 22,000 amps due to the proposed power uprate. Therefore, the NRC staff has reasonable assurance that the isolated phase bus is adequate for the power uprate condition.

3.2.2.6 Emergency Diesel Generators Power required to perform safety-related functions (pump and valve loads) is not increased with the power uprate and the current emergency power system remains adequate.

3.2.2.7 Station Blackout HBRSEP2 is classified as an alternate ac plant and is analyzed to the 8-hour station blackout (SBO) coping duration. The licensee confirmed for the proposed power uprate that the available condensate inventory in the condensate storage tank (CST) is sufficient to supply AFW for the 8-hour coping duration. The licensee re-evaluated SBO using the guidelines of NUMARC 87-00. During the first hour of the SBO event, it is assumed that AC power is not available and decay heat removal is provided by using AFW pumps. After the first hour, the alternate ac source can power a service water pump to provide an additional 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of cooling water to the SGs through the steam-driven AFW pump from the service water system (SWS).

As a result of the power uprate-related increase in decay heat rate, there will be a coincident increase in condensate inventory requirements from the CST during the SBO event. The current condensate volume (required by TS) to remove decay heat during the first hour of an SBO event is 23,000 gallons with no cooldown. Under power uprate conditions, the required condensate volume required to remove decay heat during the first hour of an SBO event is 23,270 gallons with no cooldown. Since this is bounded by the TS 3.7.5 requirement of 35,000 gallons, the NRC staff finds the licensees analysis conservative and adequate. Also, the proposed power uprate will have no impact on the methodology or implementation of the SBO coping analysis, and consequently will not adversely impact the ability of the plant to achieve and maintain safe shutdown conditions.

Based on its review, the NRC staff concludes that HBRSEP2 continues to meet the requirements of 10 CFR 50.63 and will not be affected at the power uprate conditions.

3.2.2.8 Equipment Qualification (EQ) of Electrical Equipment For the main steamline break, the current containment pressure and temperature profiles will remain bounding for the proposed power uprate. There are no changes in plant parameters following the proposed power uprate that would change the results of the LOCA analysis. The impact of the power uprate on radiological doses on EQ equipment was also evaluated by the licensee and the existing analysis was found to be bounding. In addition to the inventory of radioisotopes released from the core, the evaluation considered the contribution of the normal RCS radioactivity to the DBA LOCA source term. Compliance with TS 3.4.16 will ensure that the RCS equilibrium activities listed in UFSAR Table 11.1.1-2 will remain bounding under uprate

conditions. The impact of the uprate on normal operational doses to EQ equipment has been evaluated. The existing analyses bounds the radiation dose increase associated with the 1.7 percent increase in power for all access inside and outside of the containment.

3.2.3 Electrical Systems Summary Based on its evaluation, as set forth above, the NRC staff concluded that the proposed power uprate will have no adverse impact on the pressure, temperature, or radiation environments used in the environmental qualification of equipment, and that the plant would meet the requirements of 10 CFR 50.49 at the power uprate conditions.

3.3 I&C Technical Evaluation 3.3.1 I&C Regulatory Analysis The licensees submittals referenced Caldon Engineering Reports ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Using the LEFM Check System," (Ref. 16) and ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM TTM or LEFM CheckPlusTM System," (Ref. 17) to provide a generic basis for the proposed 1.7-percent power uprate. Engineering Reports ER-80P and ER-157P were approved by the NRC staff in safety evaluation reports (SERs) dated March 8, 1999, and December 20, 2001, respectively. This Safety Evaluation (SE) addresses the licensees plant-specific justification for a 1.7-percent power uprate. The two SERs were used by the NRC staff as the bases for verifying and approving the methodology used for analysis of the proposed modifications. The NRC staff also verified the setpoint calculations using the licensees engineering procedure EGR-NGGC-0153, based on ISA-S67.04, Part 1, 1994, "Setpoints for Nuclear Safety-Related Instrumentation," which is endorsed by NRC Regulatory Guide (RG) 1.105, Rev. 3, "Setpoints for Safety-Related Instrumentation."

3.3.2 I&C Technical Evaluation 3.3.2.1 Power Calorimetric Instrumentation Neutron flux instrumentation is calibrated to the core thermal power, which is determined by an automatic or manual calculation of the energy balance around the plant nuclear steam supply system. This calculation is called the "secondary calorimetric" for a PWR. The accuracy of this calculation depends primarily upon the accuracy of feedwater flow and feedwater net enthalpy measurements. Thus, an accurate measurement of feedwater flow and temperature will result in an accurate calorimetric calculation and an accurate calibration of the nuclear instrumentation. The HBRSEP2 design uses a venturi for flow measurement and resistance temperature detectors for temperature measurement in each of the three feedwater lines. The veuturi, however, is subject to fouling, which causes the meter to indicate a higher differential pressure, and hence a higher than actual flow rate. Calibrating the nuclear instrumentation to indicate higher than actual core power is conservative with respect to reactor safety, but causes the generation of electrical power to be lower when the plant is operated at its indicated thermal power rating.

The use of an ultrasonic flow meter implementing transit time technology was found to be a viable alternative. The Caldon LEFM is an ultrasonic flow meter, using acoustic energy pulses

to determine the feedwater mass flow rate and temperature. The LEFM TTM system, as described in ER-80P, uses eight transducers in a configuration of two transducers on each of the four acoustic measurement paths in a single path plane of the spool piece. The LEFM CheckplusTM system as described in ER-157P uses 16 transducers in a similar configuration in two orthogonal planes of the spool piece. As such, the LEFM CheckplusTM system is a combination of two LEFM TTM systems taking the average of two numerical integrations of four measurements each in two orthogonal planes. This measurement is inherently more accurate than the integration of four measurements in a single plane and, therefore, provides a better measurement accuracy.

Caldon Engineering Report ER-157P describes the LEFM CheckPlusTM system (Ref. 17) and provides calculated uncertainties in percent power for a typical PWR using measurements by a single meter LEFM TTM and LEFM CheckPlusTM systems. This report also provides a generic basis for an uprate up to 1.7 percent of the licensed reactor power with the use of the LEFM CheckPlusTM system.

Staff verification of ER-157P methodolgy:

In approving the methodology described in the licensees submittal while evaluating the enhanced system requirement, the NRC staff sought verification of the four items below.

These are depicted in Caldon Engineering Report ER-157P. The licensees responses to the four staff questions are as follows:

1) The licensee should discuss the maintenance and calibration procedures that will be implemented with the incorporation of the LEFM, and those procedures should include processes and contingencies for an inoperable LEFM. The licensee should also discuss the effect on thermal power measurement and plant operation.

(i) In response, the licensee stated in section 3.2.1.1 (Ref. 1) that implementation of the power uprate license amendment will include development of the necessary procedures and documents required for operation, maintenance, calibration, testing, and training at the power uprate level with the new LEFM CheckPlusTM system. Plant maintenance and calibration procedures will be revised to incorporate Caldon maintenance and calibration requirements prior to declaring the LEFM CheckPlusTM system operable and raising power above the current licensed level of 2300 MWt to a maximum of 2339 MWt.

Operability criteria for the LEFM CheckPlusTM system will be contained in the HBRSEP2 Technical Requirements Manual (TRM). The revision to the TRM is done under Corporate Procedure REG-NGGC-0002, 10 CFR 50.59 and Other Regulatory Evaluation. This will be evaluated for acceptance by performing a 10 CFR 50.59 SE because the TRM is incorporated into the UFSAR by reference. A TRM Specification (TRMS) had been drafted for inclusion in the TRM stating that the LEFM CheckPlusTM system must be available to perform calorimetric measurements in order to support plant operation with RTP greater than the current licensed power level of 2300 MWt. The TRMS will be incorporated into the TRM prior to operation at the power uprate level.

(ii) The HBRSEP2 TRM and other appropriate plant procedures will specify that if the LEFM CheckPlusTM system becomes unavailable during the interval between daily performances of the secondary calorimetric as per TS Surveillance Requirement (SR) 3.3.1.2, plant operations may remain at a thermal power of 2339 MWt while continuing to use the power indications from the nuclear instrumentation system channels. However, in order to remain in compliance with the bases of operation at an RTP of 2339 MWt, the LEFM CheckPlusTM system must be returned to service prior to the next performance of SR 3.3.1.2.

If the LEFM CheckPlusTM system has not been returned to service prior to the next performance of SR 3.3.1.2, the procedural guidance/TRM would specify that the reactor power be reduced to, or maintained at, a power level of no greater than 2300 MWt.

(iii) This power level is consistent with the uncertainty previously assumed for the venturi-based indication of feedwater flow. This power reduction is intended to occur prior to performing SR 3.3.1.2 using the venturi-based feedwater flow indications. Once SR 3.3.1.2 is performed using the venturi-based feedwater flow indications, the assumed power uncertainty is 2-percent RTP even though the actual uncertainty may be better than 2-percent RTP. In order to maintain compliance with the safety analyses, it would be necessary to operate the plant at a maximum core thermal power of 2300 MWt until the LEFM CheckPlusTM system is restored. Once the LEFM CheckPlusTM system is restored, performance of SR 3.3.1.2 is required using the LEFM CheckPlusTM indication of feedwater flow. Upon completion of SR 3.3.1.2, the plant could again be operated at 2339 MWt.

(2) For plants that currently have LEFMs installed, the licensee should provide an evaluation of the operational and maintenance history of the installation and confirm that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Engineering Report ER-80P.

In response, the licensee stated that this question was not applicable to HBRSEP2. HBRSEP2 currently uses venturis to obtain the daily secondary calorimetric measurements. HBRSEP2 is installing a new LEFM CheckPlusTM system as the basis for the requested uprate. It is planned for installation during the current refueling outage, which began on October 11, 2002.

(3) The licensee should confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation be based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty).

If an alternative methodology is used, the application shall be justified and applied to both venturi and the LEFM for comparison.

The licensee states that the Caldon-provided LEFM CheckPlusTM uncertainty calculations followed American National Standard Institute/American Society of Mechanical Engineers standard ANSI/ASME PTC 19.1-1985, "Test Uncertainty, Instruments and Apparatus" methodology. This methodology utilizes the SRSS to combine the power measurement uncertainty components. The SRSS methodology is an acceptable methodology for combining instrumentation in

accordance with uncertainties in accordance with Instrument Society of America standard ISA-S67.04, Part 1-1994, "Setpoints for Nuclear Safety-Related Instrumentation." The licensees submittals included plant-specific power measurement uncertainty calculations for HBRSEP2. These calculations statistically combined the Caldon-provided LEFM CheckPlusTM system feedwater mass flow and temperature measurement uncertainties with other instrumentation uncertainties affecting the plant power calorimetric uncertainty.

The licensee stated that the plant-specific uncertainty calculations are in accordance with HBRSEP2 engineering design procedure EGR-NGGC-0153, "Engineering Instrument Setpoints." The licensee also stated that EGR-NGGC-0153 is based on ISA-S67.04, Part 1-1994, which has been endorsed by NRC Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation."

(4) Licensees of plants where the ultrasonic meter (including the LEFM) was not installed with flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant-specific installation) should provide additional justification for use. The justification should show either that the meter installation is independent of the plant-specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and the plant configuration for the specific installation, including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed and calibrated LEFM, the licensee should confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

The licensee states that this question did not apply to HBRSEP2. The calibration factors for the three HBRSEP2 spool pieces will be established by tests of these spool pieces at Alden Research Laboratories prior to installation at HBRSEP2.

These will include tests of a full-scale model of the HBRSEP2 hydraulic geometry and flow disturbances. Since the uncertainty values for the Caldon LEFM CheckPlusTM system were not finalized at the time of the licensees submittals, the plant-specific power measurement uncertainty calculations used bounding values. The licensee stated that final acceptance of the site-specific uncertainty analyses will occur after the completion of the commissioning process to confirm that the actual performance in the field meets the uncertainty bounds established for the instrumentation as described in the licensees submittals.

Thus, based on the licensee responses to the questions above, the NRC staff finds that the licensee has sufficiently resolved the plant-specific concerns regarding LEFM CheckPlusTM system maintenance and calibration, hydraulic configuration, processes and contingencies for an inoperable LEFM CheckPlusTM system, and the methodology for the plant-specific calculations of the HBRSEP2 power measurement uncertainty. Therefore, the NRC staff finds the licensees responses adequate and acceptable.

3.3.2.2 System Software Quality Assurance (SQA):

The licensee stated that the LEFM CheckPlusTM system SQA activities will be conducted in accordance with the licensees Nuclear Generation Group (NGG) SQA program procedures,

and the system software verification and validation (V&V) will be documented in a software V&V report in accordance with NGG SQA program procedures. The adequacy of the LEFM CheckPlusTM system will be preserved through the use of existing programs and procedures.

Administrative control of software and hardware configurations will be maintained by the programs and procedures for configuration management, configuration control of plant digital systems, and the engineering change process. Corrective actions will be conducted in accordance with procedures governing the plant corrective action program and work management process. Reporting of deficiencies to the manufacturer will be conducted in accordance with procedures governing the operating experience program and NRC reporting requirements. Receipt and addressing of manufacturer deficiency reports will be performed in accordance with the operating experience program and vendor control program procedures.

The LEFM CheckPlusTM indications of feedwater flow and temperature will be displayed on a local display panel and transmitted to the plant process computer for use in the calorimetric calculation. This information will be directly substituted for the venturi-based flow indications and the RTD temperature indications currently used in the plant calorimetric calculation. The LEFM Check PlusTM units are also able to calculate bulk feedwater temperature with greater precision than is measured by the currently installed temperature instrumentation. Bulk feedwater temperature is determined based on a correlation between measured feedwater pressure and sound velocity. An improved feedwater pressure transmitter is being provided to further reduce feedwater temperature measurement uncertainty. The venturi-based feedwater flow measurement will continue to be used for other functions that it currently fulfills.

3.3.3 I&C Technical Summary Based on the NRC staff review of CP&Ls submittals on the LEFM CheckPlusTM system and the plant power calorimetric uncertainty, the NRC staff finds that the HBRSEP2 thermal power measurement uncertainty with the LEFM CheckPlusTM system is limited to 0.3 percent of RTP and can support the proposed 1.7-percent power uprate. The NRC staff also finds that CP&L sufficiently addressed the four additional questions outlined in the staff SER of Caldon Engineering Report ER-157P. The proposed changes to the instrumentation allowable values were calculated using an acceptable methodology and, therefore, are acceptable. The NRC staff, therefore, finds CP&Ls request for a 1.7-percent thermal power uprate to be acceptable with respect to I&C issues.

3.4 Radiological Analysis of Main Steamline Break, SG tube rupture, LOCA, Rod Cluster Control Assembly, and the Locked Rotor Accidents In their May 16, 2002, letter, the licensee submitted a request for a 1.7-percent power uprate based upon an evaluation that justified the radiological consequences of the uprate with analyses using the alternate source term (AST). This evaluation included analysis of the consequences of a main steamline break (MSLB), an SG tube rupture (SGTR), a LOCA, a single rod cluster control assembly (RCCA), an RCP shaft seizure (locked rotor), and a radioactive waste gas decay system leak or failure. Following the May 16, 2002, submittal, NRC staff discussions with the licensee indicated that a more timely staff review and approval of the power uprate amendment request might occur if approval was not dependent upon the use of the AST for postulated accidents. The NRC staff recommended using the guidance in Regulatory Issue Summary (RIS) 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications. Consequently, the licensee provided a

revised evaluation of the radiological consequences of postulated accidents for the 1.7-percent power uprate in a letter dated August 12, 2002.

3.4.1 Staff Assessment The NRC staff has reviewed the information contained in the August 12, 2002, submittal. The NRC staffs review concluded that the LOCA, RCCA, and the locked rotor accident analyses were conducted at 102 percent of the 2300 MWt power level.

The licensee provided additional information on September 6, 2002. In this submittal, the licensee indicated that the analysis of record for the SGTR utilizes a break flow and a steam release flow rate based upon an initial core power level of 102 percent of 2300 MWt. In addition, they indicated that the RCS pressure was not being changed as a result of the power uprate. While the average RCS temperature was increasing by 0.5 oF due to the power uprate, this temperature increase remains within the +/- 4 oF incorporated into the analysis of record.

The secondary side pressure in the analysis of record is 800 psia. This value remains conservative relative to the uprate condition. For the uprate, the secondary side pressure will be 806 psia. Based upon the above, the NRC staff has concluded that the SGTR is bounded by the analysis of record.

In its September 6, 2002, letter, the licensee indicated that the analysis of record for the MSLB included as sources of activity released during the MSLB the water boiled off from the affected SG, i.e., the SG with the break, plus the steam released from the two unaffected SGs during cooldown of the reactor. The analysis of record assumes that the secondary side pressure on the affected SG drops instantaneously to atmospheric while primary coolant pressure remains essentially at 2250 psia during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the break. Since the power uprate will not result in a change in RCS pressure and only a small change in primary and secondary thermal hydraulic conditions, i.e., pressure and temperature, there will not be a significant change in break flow resulting from the uprate. In addition, the fuel integrity acceptance criteria will continue to be met even at the power uprate conditions because the greatest challenge to the fuel relative to its DNBR and its centerline melt conditions continues to be at zero power.

Based upon the above, the NRC staff has concluded that the consequences of an MSLB at the proposed increased power level remains bounded by the analysis of record.

The curie contents of any waste gas decay tank are currently limited so that their release would not result in a whole body dose of 0.5 rem or greater. This limitation is irrespective of reactor power level. Consequently, the NRC staff is in agreement with the licensee that the consequences of the release or failure of a waste gas tank are not functions of the reactor power level, and that the potential consequences remain acceptable.

3.4.2 Radiological Analysis of LOCA, RCCA, and the Locked Rotor Accidents Summary The licensee has completed an evaluation that concluded operation of HBRSEP2 at the proposed 2339 MWt is bounded for approximately 95 percent of Cycle 22 (approximately 504 effective full-power days [EFPD]). This is an acceptable approach because the source term increases with burnup, but remains within the LOCA analysis of record (AOR). This evaluation was based on establishing an AOR burnup limit for Cycle 22 that accounts for operation at the proposed 2339 MWt reactor power level. Therefore, the existing AOR for the LOCA, MSLB, SGTR, Single RCCA Withdrawal, Radioactive Waste Gas System Leak or

Failure, and RCP Shaft Seizure (Locked Rotor) radiological accident analyses will bound operation at the proposed uprated power of 2339 MWt for at least 504 EFPD during current Cycle 22. However, operation beyond 95 percent (504 EFPD) of Cycle 22 at the uprated power level depends on the results of subsequent NRC staff review of the licensees AST analyses provided in the May 10, 2002, AST submittal. This review is expected to be completed early in the year 2003. Since the NRC staff accepts the licensees conclusion that the consequences remain bounded for approximately 95 percent of Cycle 22, and in order to permit CP&L to install flow measuring hardware during the current refueling outage, and permit operation at the proposed 2339 MWt reactor power level for a limited time, the NRC staff authorizes the amendment subject to the following license condition: Operation of H. B. Robinson Steam Electric Plant, Unit No. 2, is limited to 504 effective full-power days. This additional condition shall remain in effect until approval of a license amendment that removes this limitation. This license condition is being incorporated in the HBRSEP2 FOL in Appendix B, Additional Conditions.

3.5 Chemical and Volume Control System (CVCS) and Flow-Accelerated Corrosion (FAC)

Program 3.5.1 CVCS System Evaluation The CVCS provides for boric acid addition, chemical addition for corrosion control, reactor coolant cleanup and degasification, reactor coolant makeup, reprocessing of water letdown from the RCS, and RCP seal injection. During plant operation, letdown flow from the RCS cold leg flows through the shell side of the regenerative heat exchanger and then through the letdown orifices. The regenerative heat exchanger reduces the temperature of the reactor coolant and the letdown orifices reduce the pressure. The cooled low-pressure water leaves the containment and enters the auxiliary building. A second temperature reduction occurs in the tube side of the non-regenerative heat exchanger followed by a second pressure reduction due to the low-pressure letdown valve. After passing through one of the mixed bed demineralizers to remove ionic impurities, coolant flows through the reactor coolant filter and enters the volume control tank (VCT). The licensee determined that the power uprate condition will not result in a change of the RCS Tcold. Resizing of CVCS equipment will not be needed and there will be no change in the letdown and makeup requirement. However, there will be a slight increase in N-16 activity that will occur at the power uprate condition that will have a negligible effect on letdown/excess letdown line delay time requirements. Based on the information provided by the licensee, the NRC staff concludes that the CVCS will not be negatively impacted by the power uprate condition.

3.5.2 FAC The purpose of the FAC program at HBRSEP2 is to maintain the design margin in wall thickness. The program is implemented through plant procedures and fulfills the recommendations of Generic Letter (GL) 89-08, Erosion-Corrosion Induced Pipe Wall Thinning. In addition, the FAC program uses the CHECWORKS computer program to model FAC in piping systems. For the power uprate operating condition, the licensee identified feedwater heater components that may exhibit susceptibility to FAC. These components are currently modeled in the FAC program and the projections for these components in the power uprate operating condition are updated in the CHECWORKS model. In addition, the results of these models are factored into future pipe inspections and replacement plans.

The NRC staff requested additional information from the licensee regarding the predicted wear rate for the components most suspectible to FAC for the power uprate operating condition. The licensee provided a table listing the system, the component, the average and maximum percent changes in predicted wear rate, and the average and maximum change in predicted wear rate (mils/year). This table is documented in the licensees submittal dated August 12, 2002. The NRC staff reviewed this information and concluded that the percent change in predicted wear rate and the change in predicted wear rates (mils/year) were negligible and were evaluated to exceed Heat Exchanger Institute (HEI) guidelines.

3.5.3 Model 44F SG Evaluation HBRSEP2 has three Westinghouse model 44F SGs with alloy 600 thermally treated tubing. To assess the licensees evaluation of SG structural and leakage integrity, the NRC staff reviewed the effect that the power uprate would have on SG tube degradation and the SG inspection program including condition monitoring and operational assessments. The NRC staff also reviewed the applicable plugging criteria.

According to the licensee, changes in the reactor vessel outlet temperature (Thot) and the secondary-side pressure are key parameters in determining corrosion effects. These parameters are inputs into calculations used to determine tube integrity.

The licensee stated that it will not increase Thot (the hot leg temperature) after power uprate.

Industry experience has shown that, in general, a high Thot correlates with increased tube degradation. Therefore, limiting Thot to the pre-uprate range will ensure that the uprated operation will not cause the rate of overall tube degradation to increase. In addition, the licensee will disposition new degradation (unanticipated degradation) through the HBRSEP2 corrective action program and the performance of a root cause analysis.

Experience with power uprates at other plants has shown that a significant increase in steam flow (>5%) and a significant decrease in steam pressure (>100 psi) may affect flow-induced tube vibration and result in increased anti-vibration bar (AVB) wear. However, the 1.7-percent power uprating slightly increases the steam flow rate and slightly decreases the steam pressure (2.8 psi). The licensee concluded that the 1.7-percent power uprate will have a negligible impact on the projected AVB wear rate and will not significantly impact future tube wear at the AVBs. Based on its review, the NRC staff agrees with the licensees conclusion and finds it acceptable.

The licensees current SG program includes the identification and disposition of loose parts either by removal or monitoring. In addition, the SG program provides for the evaluation of the impact of loose parts through condition monitoring and operational assessments. Therefore, the NRC staff concludes that, based on the slight increase of the steam flow rate and slight decrease of the steam pressure, the 1.7-percent power uprate will have a negligible impact on tube wear caused by loose parts, and that the licensee has provided reasonable assurance that the challenges to SG tube integrity from secondary-side loose parts will be managed by the current site SG inspection program.

By letter dated May 16, 2002, the staff requested the licensee to provide a summary of its operational assessment for any active degradation mechanisms under power uprate conditions.

In its response dated September 6, 2002, the licensee stated that there are no active degradation mechanisms in HBRSEP2 SGs. The licensee stated that the SGs will continue to be assessed for degradation per site directives that meet the Electric Power Research Institute (EPRI) guidelines.

The power uprate will not require any physical changes to the SGs. The licensees evaluation of RCS components concluded, in Section 3.6.2.1, that operation at power uprate conditions was bounded by RCS design conditions. It will not impact RCS component stresses or cause fatigue usage factors to exceed allowable limits, or appreciably affect thermal expansion loads on the RCS supports, including the SGs. The NRC staff concurs with the licensees determination that the stresses generated from the uprated operating conditions are bounded by the original design condition and thermal expansion analysis. Although some of the original operating parameters are no longer applicable, the SGs and the SG supports were qualified to loads generated as part of the original analyses that are higher than those resulting from the uprated operating conditions.

Operation at higher power levels typically yields higher fluid velocities, higher vapor void fractions, a slight reduction in recirculation ratio, and possibly less margin to fluid elastic stability. The licensee performed an evaluation of the SG U-bend gap velocities at normal and uprated conditions using the ATHOS 3-D thermal hydraulic code (ATHOS3). The analysis is based on an assumed zero and 6-percent SG Tube Plugging (SGTP). The ATHOS3 study concluded that the changes in gap velocities and the drop in recirculation ratio resulting from the proposed power uprate are small (< 2%), indicating that the changes in fluid velocities and steam quality would be negligible. The licensee further stated the wear mechanism will not change and there is no significant impact on SG tube plug wear.

Both steam pressure and steam temperature in the SGs will decrease under power uprate conditions. The average heat transfer from the reactor core to each SG will increase. The licensee performed analyses to demonstrate that the SGs are capable of transferring the heat in addition to the heat from the RCPs and other non-nuclear heat sources. The higher flow rate and reduced steam density result in higher steam velocity. The integrity of the SG tubes will continue to be verified through periodic inspections and measurements performed in accordance with the SG Program developed to meet the NEI guideline in NEI-97-06. The increase in steam velocity will also result in an increase in steam hammer loads, which have been evaluated and determined to be negligible. The parameters at the current level of SGTP

(~0.1%) and a projection of parameters after the power uprate with 6-percent SGTP were analyzed and the results indicated continued acceptability of the SGs to support plant operations and analyses at the uprated power level.

On the basis of its review, the NRC staff concludes that the licensee has demonstrated that the SGs, SG supports, SG tubes, and the steam hammer loads at the uprated operating conditions will all continue to be at or within the original design limits and, therefore, are acceptable for the proposed 1.7-percent power uprate.

3.5.4 CVCS System and FAC Summary Based on the information provided in Sections 3.5.3, 4.6.5, and 3.8.3 of the submittal, as supplemented with the detailed table of the predicted wear rate from the CHECWORKS model, the NRC staff concludes that the proposed power uprate results in negligible effects on the CVCS and FAC. In addition, the NRC staff concludes that the power uprate condition will not

compromise the function of the CVCS nor adversely impact the piping systems managed by the FAC program. Also, based on the information the licensee provided, the NRC staff concludes that the power uprate will not have a significant impact on its SG tube structural and leakage integrity.

3.6 Structural and Pressure Boundary Integrity of the Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP) Systems 3.6.1 Technical Evaluation The staff reviewed the HBRSEP2 power uprate amendment as it relates to the structural and pressure boundary integrity of the NSSS and BOP systems. Affected components in these systems included reactor coolant loops (RCLs), in-line equipment and supports, the reactor pressure vessel (RPV), core support structures (CSSs), reactor vessel internals (RVIs), SGs, control rod drive mechanisms (CRDMs), RCPs, and pressurizer.

The governing normal plant transients are limited by administrative controls and/or process limits and are therefore not impacted by the power uprate. For more severe transients, the evaluations were initially based on values equal to or greater than 102-percent reactor power.

Therefore, the thermal-hydraulic transients described in the HBRSEP2 UFSAR are not adversely impacted by the power uprate.

For piping attached to the RCS hot leg, thermal expansion stresses and loads resulting from uprate conditions are bounded by existing analyses and are not impacted. Therefore, the current pipe whip/jet impingement loads are not changed, and no new break locations are postulated. The subcompartment pressurization due to a LOCA in the RCS hot leg will increase slightly, but the change is considered negligible. Therefore, the stress analyses presented in the HBRSEP2 UFSAR remain bounding for power uprate conditions.

The licensee addresses the observation of Primary Water Stress Corrosion Cracking (PWSCC) of the CRDM nozzles in the reactor vessel head in Section 3.6.2.8 of Ref. 1. RCS Thot will increase as a result of the power uprate by approximately 0.9 °F with 0 percent SGTP, and by approximately 1.3°F with 6 percent SGTP. Temperature increases of these magnitudes will not significantly increase the PWSCC susceptibility of the CRDM nozzles. Consequently, PWSCC of the CRDM nozzles will not be impacted by the increase in RCS Thot resulting from the power uprate. The licensee plans an inspection of the CRDM nozzles during the next refueling outage (RO-21).

The NRC staff's SE concerning the effects of the power uprate on the pertinent components is provided below.

3.6.2 Reactor Vessel The proposed power uprate will result in changing the design parameters provided in Table 3.6-1 (Ref. 1). The staff questioned that the RPV was not explicitly addressed in Section 3.6.2 of Ref. 1, Reactor Coolant System Component Assessments. The licensee responded by Ref. 3 that the original RCS design conditions for Tcold and Thot temperatures, as listed in Table 3.6-1, show that the reactor vessel equipment specification bounds the uprated operating conditions. By comparison of the current design parameters and the corresponding

revised parameters for use in the power uprate analysis, the NRC staff concurs with the licensees response for the RPV at HBRSEP2.

3.6.3 Reactor Core Support Structures and Vessel Internals The licensee indicated in Table 3.6-1 of Ref. 1 that the temperature changes resulting from the proposed 1.7-percent power uprate are bounded for the reactor internals by the revised RCS design conditions. Flow rates will not increase under power uprate conditions and, therefore, will have no effect on the reactor. The effects of increased fluence due to the power uprate were also determined to have a negligible effect on the material properties of the RVIs.

For the reactor vessel supports, the temperature changes resulting from the proposed power uprate are also bounded by the revised RCS design conditions, as shown in Table 3.6-1. The effects of increased fluence due to power uprate have a negligible effect on the material properties of the reactor vessel supports.

On these bases, the NRC staff concurs with the licensee's conclusion that the current design of RVIs and supports continues to be in compliance with the licensing basis codes and standards for the power uprate conditions.

3.6.4 CRDMs The licensee stated that RCS operating conditions following the power uprate are well within the design conditions of the CRDM. CRDM coil temperature could increase by approximately 1.3 °F due to the power uprate-related increase in Thot. The forced air cooling of the CRDM coil stacks maintains the coils at approximately 392 °F. Since the design temperature of the coils is 450 °F, an increase in the coils temperature of 1.3 °F will continue to provide significant margin to the design temperature. The Rod Position Indicator (RPI) coils are located above the CRDM coils and therefore are less affected by the increase in Thot.

Based on the above, the NRC staff concurs with the licensee's conclusion that the CRDM and RPI coil temperatures will not be adversely impacted by the power uprate.

3.6.5 RCPs The licensees evaluation of RCS components concluded, in Section 3.6.2.1, that operation at power uprate conditions was bounded by RCS design conditions. Power uprate will not impact the stress or fatigue of any RCS components, or thermal expansion loads on the RCS supports, including the RCPs. The NRC staff concurs with the determination that loads and stresses generated from the uprated operating conditions were bounded by the original design conditions and thermal expansion analysis. Although some of the original operating parameters are no longer applicable, the RCPs and the RCP supports were qualified to loads generated as part of the original analyses that are higher than the uprated operating conditions. There is no change in the RCP flow, net positive suction head (NPSH) availability, or RCP power requirements due to power uprate. Thus, the licensee determined that the power uprate has no impact on the ability of the RCPs to provide circulation of the reactor coolant. The existing evaluation of RCP flywheel missiles is dependent upon several parameters, none of which are impacted by the power uprate. The NRC staff concurs with the licensee's conclusion that the proposed power uprate will not impact the RCPs or RCP motors.

3.6.6 Pressurizer The licensee evaluated the structural adequacy of the pressurizer and components at limiting locations in the pressurizer, the pressurizer spray line and nozzle, and the pressurizer surge line and nozzle at the uprated conditions. The evaluation was performed by comparing the key parameters in the HBRSEP2 pressurizer full-power condition presented in the stress current report against the revised design conditions for the proposed power uprate condition.

The licensees structural evaluation concluded that the power uprate will not require any physical changes to the pressurizer and that operation at power uprate power conditions was bounded by design conditions and will not impact the stress, load, or fatigue of any RCS components and supports, including the pressurizer. Operating temperature, pressure, and water level will not change under power uprate conditions. The change in RCS fluid mass due to the Tavg increase from 575.4 °F to 575.9 °F is small, and the pressure/temperature response rate change is viewed as insignificant.

The resulting difference between the pressurizer and the RCS hot leg temperature change is less than 1 percent from current to power uprate conditions. This is considered negligible in view of pressurizer surge line and nozzle thermal stratification. There is no change in auxiliary spray and RCS cold leg temperatures due to power uprate. Consequently, pressurizer spray line and nozzle thermal stratification will not be impacted.

Based on its review of relevant sections of the Ref. 1 evaluation, and the reasoning presented therein, the NRC staff agrees with the licensees conclusion.

3.6.7 RCS Attached Primary Piping and Supports The licensee evaluated the NSSS piping and supports by reviewing the existing design-basis analysis against the power uprate conditions with regard to the design system parameters, transients, the LOCA dynamic loads, and the thermal stratification in the RCS piping caused by valve leakage and turbulent penetration.

Section 3.6.2.1 of Ref. 1 indicated that the values for Thot and Tcold under power uprate conditions are bounded by the revised RCS design condition values shown in Table 3.6-1.

Temperature data in Table 3.6-1 indicates that the Thot for the revised RCS design condition (604.6 °F) is slightly higher than that of the uprated operating conditions (604.1 °F or 604.5 °F).

However, the Tcold for the revised RCS design conditions (546.1 °F) is slightly lower than that of the uprated operating conditions (547.6 °F or 547.3 °F). On the basis of temperatures from Table 3-6.1, the NRC staff requested the licensee to justify the above-stated determination.

The licensees response in Ref. 3 indicated that the values for Thot and Tcold under power uprate conditions are bounded by the combination of the original and revised design condition values shown in Table 3.6-1. The combination of the original design conditions and the revised design conditions creates an envelope that bounds the Tcold conditions for power uprate (Tcold original >Tcold uprated conditions > Tcold revised). For Thot, both the original and the revised Thot are greater than the uprate Thot. Therefore, thermal expansion loads and stresses for the uprated conditions are bounded by existing analyses, and the power uprate conditions will not adversely affect RCS component stress and fatigue. The NRC staff found the response and the conclusion acceptable.

The licensee stated that the potential effects of valve leakage, in view of thermal stratification, will not change appreciably as a result of the power uprate since the temperature changes

resulting from the power uprate are small. RCS temperatures and flow will not change significantly as a result of the proposed power uprate such that the potential for, and effects of, turbulent penetration during power uprate operations are not affected.

On the basis of its review of the licensees submittal, and for the reasons set forth above, the NRC staff concurs with the licensees conclusion that the effect of power uprate on the attached RCS primary piping and supports is negligible.

3.6.8 BOP Systems The licensee performed heat balance calculations at an NSSS power level of 2309 MWt and 2348 MWt core power and evaluated the adequacy of the BOP systems using the data from these heat balance calculations. Feedwater flow rate and velocity, and feedwater system pressure drop will increase during full power at uprated operation. However, they remain within acceptable limits for the Main Feedwater Regulation Valve (MFRV).

The licensee indicated that the feedwater heater system pressure, temperature, and flow rate will change slightly. The feedwater heaters were evaluated on a revised heat balance reflecting power uprate conditions, and the results were compared to the design guidelines of the HEI. The licensee identified some feedwater heater components that marginally exceed HEI guidelines. It was determined, based on the slight change of feedwater heater parameters and favorable experience under current operation conditions, that the feedwater heater will not be adversely impacted by the power uprate.

The proposed power uprate results in a nominal condensate system volumetric flow increase of 2-3 percent and a slight increase in system pressures and temperatures. The condensate pumps are able to provide the NPSH required at the FW pumps with substantial margin.

The licensee evaluated the main turbine auxiliary and secondary cooling water loads. The power uprate will result in small increases to system heat load for some systems and will require adjustments to cooling water flow to some components. However, the service water and secondary system designs bound these changes. The licensee also evaluated the impact of power uprate on the following systems: condensate polishing, main condenser and condenser vacuum, moisture pre-separator and moisture separator reheater, extraction steam, heat drains and vent, circulating, and main turbine. The systems and components were found acceptable and did not require modifications to operate at uprated power.

The licensee also performed an evaluation of piping associated with these systems to determine the effects of the proposed power uprate and concluded that these piping systems remain acceptable and satisfy the design-basis requirements in accordance with the design-basis criteria. The HBRSEP2 plant piping and related supports will remain within the Code-allowable stress limits. This includes the feedwater hydraulic loads, feedwater piping thermal stratification, and main steam isolation/check valve trips.

3.6.9 BOP Valves In Section 4.5 of the amendment request, the licensee assessed the impact of the 1.7-percent power uprate on conclusions established as a result of implementation of GL 89-10, Safety-Related MOV Testing and Surveillance and GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related MOVs. The proposed power uprate will not change

the current population of 58 valves and will not require any increases to the existing maximum opening and closing limits for valves in the program.

The licensee assessed the current evaluation of GL 95-07, Pressure Locking and Thermal Binding (PLTB) of Safety-Related Gate Valves. The proposed power uprate will not change the population of valves subject to the concerns raised in GL 95-07. Susceptible valves have already been modified or need not be modified such that the power uprate will not impact their performance.

The evaluations performed to satisfy GL 96-06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions were also reviewed.

Conclusions of the existing GL 96-06 documentation and evaluations will remain valid under the proposed power uprate.

On the basis of the above, the NRC staff finds the licensee's conclusions that the power uprate will have no adverse effects on safety-related valves, and that the conclusions of the HBRSEP2 GL 95-07, and GL 96-06, as well as GL 89-10 programs, remain valid for the power uprate condition.

3.6.10 Structural and Pressure Boundary Integrity of the NSSS and BOP Systems Summary On the basis of its review of the licensees submittal and for the reasons stated above, the NRC staff concurs with the evaluations performed by the licensee for the NSSS and BOP piping, components and supports, the reactor vessel and internal components, the CRDMs, SGs, RCPs, and the pressurizer. The NRC staff finds the licensee's evaluation to be acceptable by the licensing Code of record and the original design basis and, therefore, concludes the foregoing components are acceptable for HBRSEP2 uprate operations at the proposed core power level of 2339 MWt.

3.7 NSSS/BOP Fluid Systems Interface To evaluate the adequacy of the BOP systems, the licensee compared the existing design-basis parameters and assessed the safety of operations under power uprate conditions.

The licensee concluded that the components are adequately sized for the power uprate and that the current design basis is still valid for these systems and interfaces. The systems and components analyzed were as follows.

The NSSS/BOP Fluid Systems Interface components include: (1) Main Steam System, (2) Feedwater System, (3) AFW System, (4) SG Blowdown System, (5) Component Cooling Water System, (6) SWS, and (7) Ultimate Heat Sink. The BOP systems that were analyzed for the uprated conditions include: (1) Main Feedwater System, (2) Feedwater Heater System, (3) Condensate System, (4) Condensate Polishing System, (5) Main Condenser and Condenser Vacuum System, (6) Moisture Pre-Separator and Moisture Separator Reheater, (7) Extraction Steam System, (8) Heater Drains and Vent System, (9) Circulating Water System, (10) Main Turbine, (11) Turbine Component Cooling Water, (12) BOP Piping and Supports, and (13) BOP I&C Systems. These BOP major systems and component piping systems, together with the RCS piping and supports, were evaluated for the effects resulting from the revised NSSS parameters (RCS temperatures, steam temperature and steam flow rate) and the heat balance at 2348 Mwt (2339 MWt core power, plus 9 MWt non-nuclear heat sources). The details of the NRC staff evaluations of individual systems and components are as follows.

3.7.1 Main Steam System At 100-percent RTP, normal SG operating pressure is currently approximately 810 psia.

Normal SG operating pressure at 100-percent RTP will decrease slightly under uprated operating conditions, and will remain within the system design pressure with operating margin.

The following subsections summarize the evaluation of the major Main Steam (MS) System components relative to the revised power uprate conditions. The major MS System components include the SG Main Steam Safety Valves (MSSVs), SG Power-Operated Relief Valves (PORVs), Main Steam Isolation Valves (MSIVs), Main Steam Check Valves, Main Steam Dump Valves, and the Moisture Separator Reheater Safety Valves.

3.7.1.1 MSSVs The MSSVs must have sufficient capacity so that main steam pressure does not exceed 110 percent of design pressure (the maximum pressure allowed by the ASME Boiler and Pressure Vessel Code) for any anticipated transients.

The continued acceptability of the MSSVs under power uprate conditions is established by the accident analysis detailed in Chapter 15 of the UFSAR. These accident analyses model the installed MSSVs, including the pressure drop from the SGs to the MSSVs, during various plant transients. The analysis is performed at 102 percent of current RTP, and demonstrates that peak MS System pressure is maintained below 110 percent of the system design pressure under power uprate conditions. Consequently, the MSSVs have sufficient capacity to support the proposed power uprate.

3.7.1.2 SG PORVs The SG PORVs provide a means for decay heat removal and plant cooldown by discharging steam to the atmosphere whenever the MSIVs are closed, or when the condenser is not available. Under such circumstances, the SG PORVs, in conjunction with the AFW System, permit the plant to be cooled down from the PORV pressure setpoint to the point where the Residual Heat Removal (RHR) system can be placed in service. The SG PORV setpoint is below the lowest MSSV setpoint.

During plant cooldown, the SG PORVs are either automatically or manually controlled. When in the automatic mode, each SG PORV controller automatically compares steamline pressure to the pressure setpoint, which is manually set by the plant operator. The SG PORV automatic setpoint can be lowered as desired to conduct a cooldown and/or to remain at nominal hot standby temperature and pressure.

The SG PORVs are designed to provide a means of decay heat removal and plant cooldown following a loss of condenser vacuum during full-power operation. As a result, functional capacity requirements for the SG PORVs are limited by the SG steaming rate that can be maintained by AFW system flow into the SGs. AFW system capacity and flow rate requirements are not impacted by the proposed power uprate. Consequently, the licensee concluded that the SG PORVs have sufficient capacity to support the proposed power uprate.

The SG PORVs also operate in conjunction with the Steam Dump System to increase the ability of the plant to respond to transient events such as a secondary load rejection or turbine

runback. As a result of the power uprate, rescaling of the setpoint modules for the SG PORVs will be necessary to ensure that the SG PORVs open prior to exceeding steam dump capacity.

However, the licensee stated that the power uprate will not adversely impact the function or adequacy of the SG PORVs. Based on its review, the NRC staff agrees with the licensees justification and finds the licensees conclusions acceptable.

3.7.1.3 MSIVs and Main Steam Isolation Bypass Valves The MSIVs are located outside containment and downstream of the MSSVs. The MSIVs function to isolate forward steam flow to the main steam header. Isolation of main steam flow in the event of reverse flow is provided by the Main Steam Check Valves. The MSIVs are required to close within 5 seconds following receipt of a closure signal. The ability of the MSIVs to close within the required time is maintained with the power uprate.

The MSIV Bypass Valves are used to warm the main steamlines and equalize pressure across the MSIVs prior to opening the MSIVs. The MSIV Bypass Valves perform their function at no-load and low-power conditions where the operating conditions will not be affected by the power uprate. Consequently, the NRC staff accepts the licensees analysis that the proposed power uprate will have no impact on operation of either the MSIVs or the MSIV Bypass Valves.

3.7.1.4 Main Steam Check Valves The Main Steam Check Valves are located downstream of the MSIVs and prevent reverse flow in the main steamlines in the event of an upstream MSLB. The ability of the main steam check valves to close will not be adversely impacted by the proposed power uprate. The power uprate-related increase in main steam forward steam flow and velocity will tend to increase the force holding the Main Steam Check Valve discs out of the flow stream and against the stop, which is the preferred operating position and is not detrimental to valve operation. The dynamic loading on the Main Steam Check Valves will also not be adversely affected under MSLB conditions. Based on the above, the NRC staff finds the licensees evaluation of the Main Steam Check Valves acceptable.

3.7.1.5 Main Steam Dump Valves and Steam Dump Control System To reduce the probability of unwarranted reactor trips, the Main Steam Dump System is designed with the capacity to reject approximately 39 percent of the uprated full load main steam flow. A total of approximately 56 percent of the uprated full load main steam flow can be rejected when the Main Steam Dump System is operated in conjunction with the SG PORVs.

This capacity provides a means to absorb limited load rejections on the turbine-generator that occur more rapidly than the reactor power level can be reduced. In conjunction with the SG PORVs and the MSSVs, the Main Steam Dump System is designed to prevent the MS System from exceeding its design pressure during all phases of operation. The Main Steam Dump System also provides a means for removal of stored heat and decay heat from the RCS during heatup and cooldown.

The increase in steam flow under power uprate conditions will result in a slight increase in required steam dump capacity. The installed steam dump capacity will remain unchanged.

However, there is sufficient steam dump capacity currently installed to satisfy Main Steam Dump System requirements under uprated power conditions.

The Steam Dump Control System is actuated by coincidence of a large load change (as sensed by main turbine 1st stage pressure), and a high error signal between Tavg and a programmed reference value (Tref) for RCS average temperature. Predominantly as a result of replacement of the high-pressure main turbine rotor, but also partially due to the power uprate, the pressure verses power relationship for main turbine 1st stage pressure will change. This change will necessitate rescaling of the main turbine 1st stage pressure input to the steam dump controls.

The NRC staff agrees with the licensees analysis and finds its conclusion that the operation of the steam dump control system will not be adversely impacted by the power uprate acceptable.

3.7.2 Feedwater System The Main Feedwater (FW) System, in conjunction with the Condensate and Heater Drains and Vent Systems, must automatically maintain SG water level within a programmed band during steady-state operations and limited transients and events. The major components of the FW system are the MFRVs, Main Feedwater Header Section Valves, Feedwater Bypass Valves, and the Main Feedwater Pumps.

The range of revised NSSS performance parameters under power uprate conditions results in a nominal feedwater volumetric flow increase of approximately 2-3 percent during full-power operations. The higher feedwater flow rate has an impact on system pressure drop, which will increase as a result of the power uprate. The system has been evaluated to accommodate the increased pressure drop and flow requirements for the proposed power uprate. The licensee concluded that the components are adequately sized for the power uprate and that the current design basis is still valid. For the reasons set forth above, the NRC staff finds the licensees evaluation appropriate and acceptable.

3.7.3 AFW System The AFW System supplies feedwater to the secondary side of the SGs when the normal feedwater system is not available, thereby maintaining the SG heat sink. The AFW System provides an alternate source of feedwater to the SGs during normal unit startup, hot standby, and cooldown operations, and functions as an ESF System. The AFW System is required to remove decay heat during transients and accidents. The minimum flow requirements for the AFW System are dictated by the accident analyses that have been performed at 102 percent of current RTP.

The licensee has evaluated the functionality of this system and concluded that the power uprate will result in a slight increase in the total amount of water that must be provided from the Condensate Storage Tank (CST) for the SBO event.

Based on the information provided by the licensee and the experience gained from prior NRC staff reviews of power uprate applications for similar PWR plants, as set forth in Section 3.2.2.7 of this SE, the NRC staff finds the licensees evaluation acceptable.

3.7.4 SG Blowdown System The SG Blowdown System is used in conjunction with the chemical addition and sampling systems to maintain the chemical composition of the SG shell-side water within the specified limits. The blowdown system also controls the buildup of solids in the bottoms of the SGs. The blowdown flow rates needed during plant operation are based on chemistry control and tubesheet sweep criteria to control the buildup of solids.

Neither condenser leakage nor secondary water makeup quality will be impacted by the proposed power uprate. Consequently, the blowdown rate needed to address dissolved solids will not be impacted. For the blowdown flow control valves, based on the revised range of NSSS parameters associated with the power uprate, the operating range of these valves is adequate to preclude any impact on blowdown flows due to the power uprate.

As a result, the blowdown rates to control secondary chemistry and particulates will not be significantly impacted by the proposed power uprate. Based on the information provided by the licensee, as set forth above, the NRC staff finds the licensees evaluation appropriate and conservative.

3.7.5 Component Cooling Water System The Component Cooling Water (CCW) System provides an intermediate cooling loop for removing heat from reactor plant auxiliary systems and transferring it to the SWS during plant operation, plant shutdown, and the post-accident recovery period. During normal operation, one CCW pump and one CCW heat exchanger have sufficient capacity to transfer the design heat load from the components served. The power uprate-related increase in spent fuel pool (SFP) cooling load will not adversely impact the CCW System since sufficient CCW capacity is maintained.

The power uprate will result in a small increase in heat loads for the CCW System. CCW System heat removal requirements for the RHR heat exchangers and other CCW System loads at the power uprate conditions are bounded by existing analyses. Based on these analyses, the licensee stated that no changes or modifications to the CCW System are required to support the power uprate. Based on the information provided by the licensee, as set forth above, the NRC staff finds the licensees evaluation appropriate.

3.7.6 SWS The SWS provides cooling water from Lake Robinson to various safety-related and non-safety-related components during power operation, plant shutdown, and the post-accident recovery period. Heat rejection to the SWS will increase slightly during normal operation as a result of the proposed power uprate. However, the SWS design pressure and temperature will not be exceeded. Additionally, SWS flow demands will increase during normal operation as a result of increased turbine-generator heat loads. The increased heat load and flow demands are within the design capacities of the SWS pumps, heat exchangers, and control valves.

The timing and conduct of system alignments by operators during normal startup, standby, and cooldown will not be affected by the power uprate. Letdown and decay heat loads are manually controlled during these major evolutions. Thus, minor increases in primary and secondary system stored energy and the increase in decay heat load will not translate into a perceivable

increase in SW flow requirements or primary system cooldown and heatup times. Although the SWS will experience slightly higher heat loads during normal operation as a result of the power uprate, the existing system will continue to satisfy its normal and accident functions without requiring modifications to the system. Based on the information provided by the licensee, as set forth above, the NRC staff finds the licensees evaluation appropriate and conservative.

3.7.7 Ultimate Heat Sink Plant waste heat is rejected to the Lake Robinson Ultimate Heat Sink (UHS) during normal operation and accident conditions via the open-cycle Circulating Water System and SWS.

Water is taken directly from the lower end of the lake through a submerged inlet to an intake structure, and pumped through an underground conduit system for plant use. Water is discharged back to the lake near its upper end through a 4.2-mile-long discharge canal.

The waste heat to the UHS during normal power operation will increase in approximately the same proportion as the requested power uprate of approximately 1.7 percent. The expected increase in average UHS temperature is less than 0.2EF following the power uprate. The licensee is subject to the monitoring requirements of the Environmental Protection Agency, as delineated under the National Pollutant Discharge Elimination System (NPDES) program. The current NPDES permit specifies limitations on discharge temperature, dam release temperature, and circulating water system flow. These discharge limits will not change as a result of the proposed power uprate, and HBRSEP2 will continue to comply with the NPDES permit.

At the minimum allowable level of 218 ft. mean sea level (MSL), the UHS provides a 22-day supply of cooling water to the SWS pumps for accident mitigation under worst-case local meteorological conditions. The existing design-basis analysis of record for the UHS remains bounding with respect to the minimum available coolant inventory and level following power uprate. Based on the information provided by the licensee, as set forth above, the NRC staff finds the licensees evaluation appropriate and conservative.

3.7.8 Condensate Polishing System The Condensate Polishing System will experience increased condensate mass flow rate as a result of the proposed power uprate. The increased flow rate will also result in an increased pressure drop across the system, decreased inlet pressure, and a negligible change in inlet temperature. Both the inlet pressure and inlet temperature following the power uprate remain well below maximum allowable values. The condensate mass flow rate will increase due to the power uprate, which results in a corresponding increase in service velocity (flux) for the ion exchange resins. Polisher flux will remain below the maximum service velocity for the ion exchange resins. The increased system pressure drop resulting from the increased condensate mass flow rate will also remain below the maximum allowable pressure drop for the system. Therefore, the NRC staff finds the condensate polishing system acceptable for the power uprate.

3.7.9 Main Condenser and Condenser Vacuum System The licensee stated that the heat load and steam flow to the Main Condenser under power uprate conditions will result in a slight increase in condenser backpressure. The increased backpressure and flow are within acceptable limits and are bounded by the condenser design.

The Condenser Vacuum System will also be adequate to support operations at the uprated

power level. No modifications will be required to either the Main Condenser or Condenser Vacuum System. Based on the above, the NRC staff finds this system will perform satisfactorily at the uprated condition.

3.7.10 Moisture Pre-Separator and Moisture Separator Reheater The Moisture Pre-Separators and Moisture Separator Reheaters (MSRs) will experience slight increases in steam flows, drain flows, and operating pressures and temperatures under power uprate conditions. These increases are acceptable and continue to provide sufficient margin to design values. Additionally, the installed relief capacity of the MSR safety valves is sufficient to accommodate the MSR relief valve capacity needed at power uprate conditions.

Consequently, the Moisture Pre-Separators and MSRs are adequate to support operations at the power uprate level and no modifications will be required to the Moisture Pre-Separators or MSRs. Based on the above, the NRC staff finds this system will perform satisfactorily at the uprated condition.

3.7.11 Extraction Steam System The Extraction Steam System transmits steam from the high-pressure and low-pressure main turbines to the shell side of the feedwater heaters. During normal operation, steam from the high-pressure turbine is used to heat feedwater flowing through the 5th and 6th point feedwater heaters, and steam from the low-pressure turbine is used to heat feedwater flowing through the 1st through 4th point feedwater heaters.

Extraction steam flow will increase by approximately 2 percent as a result of the proposed power uprate for the 6th point feedwater heater. Based on the licensees analysis, the system piping and components (e.g., non-return valves) were found to be acceptable for operation at the uprated conditions. The NRC staff finds the licensees analysis acceptable for the Extraction Steam System.

3.7.12 Circulating Water System The Circulating Water (CW) System is an open-loop cooling system that provides cooling water for the main condenser of the turbine-generator unit. The cooling water is taken from and discharged to Lake Robinson.

CW System flow will remain essentially unchanged following power uprate. The increased levels of rejected heat resulting from increased turbine exhaust flow at uprated conditions will increase the CW outlet temperature by approximately 0.2EF and will cause a slight but acceptable increase in main condenser backpressure. The increase in CW outlet temperature due to the increased heat load is bounded by the CW System design and can be accommodated by Lake Robinson. No modifications to the CW System or its components are required for the proposed power uprate. Based on the above, the NRC staff finds the licensees analysis acceptable.

3.7.13 Main Turbine The capability of the Main Turbine to perform at power uprate conditions has been evaluated.

This evaluation included consideration of the throttle valves, high-pressure and low-pressure turbines, and associated auxiliary equipment (i.e., lube oil cooling, non-return valves,

electro-hydraulic control, etc.), including the MSRs and MSR relief valves. The evaluation also considered the effect of power uprate on the turbine missiles analysis. The main turbine and auxiliary equipment components are adequate to support operation at the power uprate level without requiring equipment modifications. Additionally, the turbine missiles analysis remains bounding under power uprate conditions. Therefore, the NRC staff finds the licensees analysis conservative and acceptable.

3.7.14 Turbine CCW Cooling water for secondary-side components, including main turbine auxiliary systems and components, is provided by the turbine building loop of the SWS. The SWS removes heat from secondary-side plant auxiliary systems and discharges it to Lake Robinson. The SWS is discussed in more detail in Section 3.7.6.

The main turbine auxiliary and secondary cooling water loads were evaluated at power uprate conditions. The power uprate will result in small increases to system heat load for some systems and will necessitate adjustments to cooling water flow to some components. However, the SWS and secondary system designs bound these changes. Based on the above, the NRC staff finds this system will perform satisfactorily at the uprated condition.

3.7.15 BOP Piping and Supports 3.7.15.1 BOP Structural Analysis An evaluation of piping systems associated with the Main Steam, Extraction Steam, Feedwater, High-Pressure Heater Drains, and Condensate systems was performed to determine the effects of the proposed power uprate. The evaluation concluded that these piping systems remain acceptable. The piping systems will continue to satisfy the design-basis requirements in accordance with the applicable design-basis criteria under the temperature, pressure, and flow rate effects that will result from power uprate conditions. The licensee states that the HBRSEP2 plant piping and related support systems will remain within Code-allowable stress limits. Accordingly, the power uprate is acceptable with respect to the BOP structural analysis.

3.7.15.2 Feedwater Hydraulic Analysis There are no specific commitments for HBRSEP2 related to waterhammer. The piping configuration for the Main Feedwater System is consistent with the Westinghouse recommended configuration for feedwater connections to the SG to reduce the effects of waterhammer. The licensee states that the SG piping configuration is not modified by the proposed power uprate. Accordingly, the power uprate is acceptable with respect to the feedwater hydraulic analysis.

3.7.15.3 Feedwater Thermal Stratification A finite element analysis of the 16-inch Main Feedwater System piping to determine the effects of thermal flow stratification was previously performed. The proposed power uprate does not impact this analysis because the initial temperature for the feedwater piping and the temperature of the cold auxiliary feedwater used in the analysis bound the uprated power conditions. Accordingly, the power uprate is acceptable with respect to the feedwater flow stratification.

3.7.15.4 MSIV/Main Steam Check Valve Trip Analysis The worst case for the MSIV/Main Steam Check Valve Trip internal stress analysis has been determined to occur at 53 percent of current licensed power. Consequently, the proposed power uprate will not impact the effects of a MSIV/Main Steam Check Valve Trip.

3.7.16 Heater Drains and Vent System The Heater Drains and Vent System and associated equipment were evaluated to ensure the ability of the system to function under power uprate conditions. Heater drain design parameters were reviewed and compared against power uprate conditions. Based on the licensees analysis, it is concluded that acceptable margins exist for operation at uprate power conditions.

The licensees analysis concluded that post-power uprate pressures and temperatures remain bounded by the existing design for the Heater Drains and Vent System and its components.

The heater drain control valves have been analyzed and have sufficient margin to operate at the increased flow rates that will result from the power uprate. Based on the above, the NRC staff finds the power uprate acceptable with respect to this system.

3.7.17 NSSS/BOP Fluid Systems Interface Summary Based on the information provided by the licensee, which is described above, and the experience gained from the NRC staff reviews of power uprate applications for similar PWR plants, the NRC staff finds the licensees evaluation conservative. The NRC staff concurs with the licensee and finds that operation at the proposed 1.7-percent uprate power level will have negligible impact on the operation of these BOP systems. Therefore, the NRC staff finds these systems acceptable for the power uprate.

3.8 RCS Component Assessments 3.8.1 Background The NRC staff reviewed the information pertaining to the effect of the requested power uprate on the structural integrity of the RVIs. Maintenance of the structural integrity of the RPV internals is required in order to demonstrate that the functional requirements of the RPV internals are met. These functional requirements include core support and ECCS performance aspects. As such, the structural integrity of the RPV internals is linked to regulatory requirements in 10 CFR 50.46 regarding ECCS performance and maintaining a coolable core geometry. Additional criteria regarding the evaluation of the structural integrity of RPV internals may be found in Ref. 9, ASME Code Sections III and XI, and other standards used in the NRCs review of the original licensing basis of HBRSEP2.

3.8.2 Regulatory Assessment The NRC staff reviewed the information pertaining to the continued applicability of Leak-Before-Break (LBB) to piping systems at HBRSEP2 that had been approved for LBB prior to the requested power uprate. Application of LBB is used to address the elimination of dynamic effects associated with pipe rupture from a facilitys licensing basis. This consideration is related to the requirements in Appendix A to 10 CFR Part 50, GDC 4 or, for those plants licensed prior to the development of Appendix A to 10 CFR Part 50, similar requirements that

were imposed during the staffs review of the facilitys operating license. HBRSEP2 was licensed prior to the development of Appendix A to 10 CFR Part 50. In evaluating the technical basis for a licensees application of LBB, the NRC staff references NUREG-1061, Volume 3, (Ref. 12) and Standard Review Plan (SRP) Section 3.6.3 (Ref. 9).

The NRC staff reviewed the information in the licensees submittal (Ref. 1) pertaining to the HBRSEP2 RPV pressure-temperature (P-T) limit curves and RPV surveillance capsule program and withdrawal schedule. This review also involved evaluating the licensees proposed modifications to HBRSEP2 TS Figures 3.4.3-1 and 3.4.3-2. The NRCs regulatory requirements related to the establishment of RPV P-T limit curves for any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, are given in Appendix G to 10 CFR Part 50, which also references, as incorporated in 10 CFR 50.55a, the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G.

Additional guidance for the NRC staffs review of RPV P-T limit curves is provided in SRP Section 5.3.2 (Ref. 9), Ref. 10, and Ref. 11. The NRCs regulatory requirements related to the establishment of a facilitys RPV surveillance capsule program and withdrawal schedule are given in Appendix H to 10 CFR Part 50, which also references the guidance in American Society for Testing and Materials Standard Practice E 185 (Ref. 13). SRP Section 5.3.1 (Ref. 9) also applies.

3.8.3 Technical Evaluation With regard to section 3.6.2.8 of the licensees submittal, the NRC staff reviewed the information pertaining to the effect of the requested power uprate on the potential for PWSCC of CRDM nozzles. Maintenance of the leakage and structural integrity of the CRDM nozzles is directly related to requirements in Appendix A to 10 CFR Part 50, GDC 14 or, for those plants licensed prior to the development of Appendix A to 10 CFR Part 50, similar requirements that were imposed during the staffs review of the facilitys operating license. Guidance regarding an acceptable evaluation of the potential for PWSCC (and the performance of appropriate inservice inspections related to the evaluation) has been provided in NRC Bulletin 2002-02, Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs, and the NRC staff evaluation of submittals from the industry and individual licensees to this bulletin.

The NRC staff reviewed information pertaining to the effect of the requested power uprate on the structural integrity of RCP flywheels and the potential for missile generation as a result of their failure. The evaluation of RCP flywheel integrity is related to Appendix A to 10 CFR Part 50, GDC 1 and GDC 4 or, for those plants licensed prior to the development of Appendix A to 10 CFR Part 50, similar requirements that were imposed during the staffs review of the facilitys operating license.

With regard to the information on protection against pressurized thermal shock submitted by the licensee in their July 25, 2002, supplemental response, the NRC staff reviewed against the regulatory requirements established in 10 CFR 50.61. As with P-T limits, Branch Technical Position (BTP) MTEB 5-2 (Ref. 11) may also apply with respect to the determination of initial, unirradiated properties of RPV materials.

With regard to the information on RPV material upper shelf energy analyses submitted by the licensee in their July 25, 2002, supplemental response, requirements related to RPV material upper-shelf energy (USE) are given in Appendix G to 10 CFR Part 50. Additional guidance for

the NRC staffs review of USE analyses is provided in RG 1.99, Revision 2 (Ref. 10), SRP Section 5.3.2 (Ref. 9) and the BTP (Ref. 11). Appendix K to Section XI of the ASME Code and RG 1.161 (Ref. 14) may also be used as guidance when USE equivalent margins analyses are necessary.

3.8.4 RCS Components Evaluation In Sections 3.6.1 and 3.6.2.3 of Attachment II to the licensees May 16, 2002, submittal (Ref. 1),

the licensee concluded that the LBB characteristics of the RCS piping will remain valid under power uprate operating conditions. The continued applicability of the LBB concept for operating conditions that bound the power uprate operating conditions was also demonstrated by Westinghouse evaluations performed in support of license renewal and documented in WCAP-15628 (Ref. 8).

Based on the changes in pressure, temperature, and operating loads expected to result from the proposed 1.7-percent power uprate, the NRC staff agrees with the licensees conclusion that the effect of the proposed power uprate on the facilitys existing LBB evaluations will be insignificant. However, the NRC staff notes that due to recent events concerning PWSCC of Inconel 82/182 material, the NRC staff is in the process of examining the significance of this issue with respect to existing LBB evaluations. Currently, the NRC staff is evaluating what licensee actions, if any, are necessary to ensure that the technical bases for existing LBB approvals remain valid, and any concerns regarding the effect of PWSCC on existing LBB evaluations will be resolved separately from this power uprate submittal. As noted above, the NRC staff also expects the impact of the proposed power uprate on the PWSCC susceptibility of any Alloy 82/182 materials in lines approved for LBB at HBRSEP2 to be insignificant due to the small projected increase in Thot.

With regard to the topic of the RPV P-T limits, the licensee concluded in Section 3.6.2.1 that the RCS P-T limit curves of TS 3.4.2, Figures 3.4.3-1 and 3.4.3-2, will be re-designated from 24 effective full-power years (EFPY) of operation to 23.96 EFPY to reflect a conservative projection of the increase in neutron fluence associated with the power uprate. This projection will ensure that the requirements of 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, will continue to be met following the proposed power uprate. In addition, the NRC staff also reviewed related information regarding the licensees updated RPV fluence information in their July 25, 2002, supplemental response. Based on the NRC staffs approval of the current HBRSEP2 P-T limit curves for 24 EFPY and the aforementioned updated RPV fluence information, the NRC staff has concluded that the existing HBRSEP2 P-T limit curves will continue to be acceptable based on meeting the requirements of Appendix G to 10 CFR Part 50 for operation up to at least 23.96 EFPY after the requested power uprate is implemented. In fact, the NRC staff noted that this action to reduce the period of applicability of the HBRSEP2 P-T limit curves was very conservative given the overall reduction in projected RPV fluence values that were cited in the licensees July 25, 2002, RAI response. Therefore, the changes to HBRSEP2 Unit 2 TS Figures 3.4.3-1 and 3.4.3-2 are acceptable.

With regard to the HBRSEP2 RPV surveillance program and capsule withdrawal schedule, the licensee concluded in Section 3.6.2.1 that there will be no impact to the reactor vessel coupon withdrawal schedule and inspection frequency as a result of the power uprate. Based on the revised RPV fluence information submitted in the licensees July 25, 2002, supplemental

response and the NRC staffs review of the current HBRSEP2 surveillance capsule withdrawal schedule, the NRC staff agrees with the licensees conclusion since the requirements of Appendix H to 10 CFR Part 50 will continue to be met.

With regard to the licensees evaluation of the continued integrity of the RPV internals, in Section 3.6.2.2 the licensee stated that the temperature changes resulting from the proposed power uprate are bounded for the RVIs by the revised RCS design conditions. The effects of flow on the RVIs were evaluated. Flow rates will not increase under power uprate conditions and will have no effect on the RVIs. Similarly, the effects of increased fluence due to the power uprate were also determined to have a negligible effect on the material properties of the RVIs.

Based on the information provided by the licensee regarding changes to operating temperature, flow rates, and neutron fluences that result from the power uprate, the NRC staff agrees that the integrity of the RPV internals will be maintained such that the licensees ability to meet the regulatory requirements in 10 CFR 50.46 regarding ECCS performance and maintaining a coolable core geometry will not be adversely impacted.

With regard to the issue of RCP flywheel integrity and the potential for missile generation as a result of flywheel failure, the licensee stated in Section 3.6.2.9 that the existing evaluation of RCP flywheel missiles is dependent upon the maximum overspeed of the turbine generator in combination with the postulated crack size and crack growth in the flywheel, none of which are impacted by the power uprate.

Based on the information provided by the licensee regarding the operational changes that will result from the proposed power uprate as described above, the NRC staff agrees with the licensees conclusion. Therefore, the NRC staff concludes that with regard to RCP flywheel integrity, the requirements of Appendix A to 10 CFR Part 50, GDC 1, and GDC 4, or similar requirements that were imposed during the staffs review of the facilitys operating license, will continue to be met.

With regard to the effect of the requested power uprate on the potential for PWSCC of CRDM nozzles, the licensee concluded in Section 3.6.2.8 (Ref. 1) that there are several factors that can contribute to PWSCC in these nozzles, one of which is the hot leg temperature (RCS Thot).

The licensees analysis showed that this will increase as a result of the power uprate by approximately 0.9 EF with zero percent SGTP, and by approximately 1.3 EF with 6-percent SGTP. Temperature increases of these magnitudes will not significantly increase the PWSCC susceptibility of the CRDM nozzles. Consequently, PWSCC of the CRDM nozzles will not be impacted by the increase in RCS Thot resulting from the power uprate. An inspection of the CRDM nozzles is planned for the next refueling outage (RO-21).

In the licensees July 25, 2002, supplemental response, a clarification regarding the final statement above was also provided:

Section 3.6.2.8 of the amendment request provides a conclusion pertaining to the effects of the higher RCS temperature on the issues in NRC Bulletins 2001-01 and 2002-01. The statement pertaining to the planned inspection activities for the upcoming refueling outage (RO-21) is a reiteration of commitments previously made in response to NRC Bulletins 2001-01 and 2002-01. The statements in Section 3.6.2.8 are clarified as follows:

The details pertaining to reactor vessel head inspection activities are provided in the HBRSEP, Unit No. 2, responses to NRC Bulletins 2001-01 and 2002-01.

Based on the insignificant impact of the slight increase in RCS Thot resulting from the power uprate, these inspection activities are not expected to be affected by the change in reactor operating conditions proposed in this license amendment request.

Since HBRSEP2 has already been categorized by the NRC staff as a high susceptibility plant for CRDM PWSCC, the NRC staff agrees with the licensees conclusion that the small increase in Thot that will result from the proposed power uprate will not have a significant effect on the inspections that the NRC staff will expect the licensee to perform on the RPV head. The NRC staff will continue to evaluate the RPV head inspections proposed by the licensee in response to NRC Bulletins 2001-01, 2002-01, and 2002-02, and will resolve any additional concerns that may arise with the proposed inspections in the context of our review of their Bulletin responses.

Therefore, with regard to this submittal, the NRC staff concludes that the licensee has provided sufficient information to demonstrate that the proposed power uprate will not significantly affect the potential for RPV head penetration degradation due to PWSCC.

Finally, regarding the PTS and USE analyses for the HBRSEP2 RPV, the licensee concludes the following in their July 25, 2002, supplemental response.

The fluences for the HBRSEP, Unit No. 2 vessel beltline materials for the current and the projected uprate operation are summarized in the following table. The current EOL [end of license] fluences are docketed and referenced in [the NRCs] Reactor Vessel Integrity Database (RVID), and the uprate fluences were projected using ENDF/BVI [Evaluated Nuclear Data File - Brookhaven, 6th edition] cross-sections and comply with Regulatory Guide 1.190.

Material EOL Fluence (n/cm2) EOL Fluence (n/cm2)

(Current) (Uprated Operation)

Intermediate Shell Plates 4.80 x 1019 3.67 x 1019 Upper Circumferential Weld 1.80 x 1019 1.57 x 1019 (W5214) & Upper Shell Plates Lower Circumferential Weld 2.00 x 1019 1.67 x 1019 (34B009) & Lower Shell Plates Axial Welds (Heat 86054B) 3.93 x 1019 2.73 x 1019 Since there are no changes to the material chemistries or the initial RTNDT(U), the currently docketed PTS values for each of the vessel materials remains bounding in accordance with 10 CFR 50.61. The proposed power uprate has no effect on the vessel PTS evaluation.

The upper shelf energy (USE) values are determined by reducing the unirradiated USE by an amount of predicted decrease as a function of fluence and copper content per Regulatory Guide 1.99, Revision 2, Position 2.2 (Reference 10). There is no change in the chemistries, and the current fluences

bound the predicted uprate fluences as shown in the response to Question (1).

Therefore, the current USE values are conservative and power uprate operation has no effect on the USE evaluation.

HBRSEP2 has end-of-license fluence values that account for the proposed power uprate and were calculated using a methodology adhering to guidance of RG 1.190. Those values are acceptable and they are not affected by the interim change of the P-T curves. Therefore, the values of the pressurized thermal shock remain valid.

The NRC staff has evaluated the information provided by the licensee and additional information submitted previously to the NRC and contained in the staffs Reactor Vessel Integrity Database. Based on the revised fluence values noted in the table above, the NRC staff independently confirmed that the HBRSEP2 RPV materials would continue to meet the PTS screening criteria requirements of 10 CFR 50.61 and the USE requirements of Appendix G to 10 CFR Part 50 through end of life.

3.8.5 RCS Component Summary The NRC staff has reviewed information provided in Sections 3.6.1, 3.6.2.1, 3.6.2.2, 3.6.2.3, 3.6.2.8, and 3.6.2.9 of the licensees May 16, 2002, submittal (Ref. 1), the proposed modifications to the HBRSEP2 TS Figures 3.4.3-1 and 3.4.3-2, and the licensees July 25, 2002, RAI response. Based on the foregoing, the NRC staff has concluded that sufficient information regarding the continued acceptability, with respect to the applicable NRC regulations, of the HBRSEP2 RPV, RPV internals, and LBB-approved piping has been provided to support NRC approval of a 1.7-percent power uprate for this unit. Further, the NRC staff finds the revised P-T limit curves (TS Figures 3.4.3-1 and 3.4.3-2) to be acceptable for up to 23.96 EFPY of operation for HBRSEP2.

3.9 Containment/Fire Protection Systems 3.9.1 Containment Performance Evaluation The licensee states that the containment analyses discussed in UFSAR 6.2.1 were performed at a power level of 2346 MWt (102 percent). However, due to slight changes in RCS conditions as well as the increased Feedwater Regulation Valve (FRV) flow capacity, the containment response to a LOCA and an MSLB under the proposed power uprate conditions was reevaluated. The evaluation determined that there are no changes in the plant parameters following the proposed uprate that would change the results or conclusions of the LOCA analysis presented in UFSAR 6.2.1.

The scope of MSLB evaluations performed to determine the impact of changes associated with the power uprate was limited to Hot Full Power (HFP) cases, since the hot zero power (HZP) cases would not be affected by these changes. The results of the analyses showed that the HZP Steamline Check Valve failure case continues to produce the limiting containment pressure response with a peak pressure of 41.85 psig; therefore, the changes associated with the proposed power uprate will not cause a postulated pipe failure in containment (LOCA or MSLB) to exceed the containment design pressure (42 psig).

3.9.2 Containment High-High Pressure The licensee stated that the proposed power uprate will not change the analysis values used in the UFSAR Chapter 15 accident analysis, or the analysis values used in the containment analysis. Additionally, the power uprate will not alter the limiting environmental conditions at the pressure transmitters, which are located outside containment, will not alter environmental conditions for the electronics, and will not alter the calibration of the equipment. Therefore, there is no impact on the instrument uncertainty for this function.

Since the analytical values and the uncertainty do not change, neither the TS Nominal Trip Setpoint, nor the TS Allowable Value for Containment High-High Pressure actuation of the Containment Spray, Phase B Isolation, or Steamline Isolation functions will change as a result of the power uprate. The TS limits on containment pressure will ensure that the Containment High-High Pressure setpoint is sufficiently above the normal containment pressure during operation.

3.9.3 Fire Protection Systems 3.9.3.1 Regulatory Basis The Fire Protection Program at HBRSEP2 is based on NRC criteria, National Fire Protection Association (NFPA) standards, Institute of Electrical and Electronic Engineers (IEEE) standards, and other industry codes. The program complies with Appendix A of BTP APCSB 9.5-1 (Ref. 11), dated August 23, 1976.

The objective of the Fire Protection Program is to minimize both the probability and consequences of postulated fires. The probability and consequences of such fires are minimized by a combination of design features, procedural controls, and personnel training, including a well-trained fire brigade.

3.9.3.2 Fire Protection Evaluation The licensee states that the procedural controls include provisions to ensure that plant modifications and design changes are controlled in order to ensure that plant structures, systems, and components continue to meet their performance and functional objectives. Plant procedures provide written instructions that describe the modification process and the means for documenting the changes and activities needed to support the power uprate. As a part of this process, consideration of the effects the modification may have on the Fire Protection Program is required.

The combustible equipment and changes to the plant systems, structures, and components that are being installed or modified to support the proposed power uprate have been evaluated with respect to their impact on the Fire Protection Program. Fire loading margins will not be challenged or exceeded by the power uprate-related plant modifications. Consequently, there is no impact to the Fire Protection Program as a result of the proposed power uprate.

Part 50, Appendix R of 10 CFR, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, requires that fire protection be provided for structures, systems, and components required for safe shutdown. The HBRSEP2 safe shutdown analysis and methodology is described in plant document FPP-RNP-300, Rev. 6 (Ref. 15). In order to

satisfy the Appendix R requirements for ensuring adequate core cooling, and also to maintain a minimum inventory in the SGs, it is currently necessary that the AFW System be initiated within a specified time.

The licensee has analyzed the plant fire protection design basis above and based on evaluation of the power uprate, the current AFW System initiation time is not impacted. The licensee states that the proposed power uprate will have no impact on the methodology or implementation of the HBRSEP2, Appendix R safe shutdown analysis, and consequently will not adversely impact the ability of the plant to achieve and maintain safe shutdown conditions.

The existing system designs are adequate to accommodate the increased decay heat removal associated with operation at higher power levels. Therefore, the NRC staff finds the proposed power uprate acceptable with respect to fire protection requirements.

3.9.4 Containment/Fire Protection Systems Summary Based on the review of the licensees analysis as set forth above, the NRC staff finds that the results are reasonable, conservative, and therefore acceptable with respect to the containment and fire protection systems.

3.10 Human Factors and Other Issues 3.10.1 Human Factors Regulatory Basis This evaluation is limited to the operator performance aspects resulting from the increased allowable maximum power level. It includes changes to operator actions, human-system interface changes, and to procedures and training resulting from the change in maximum power level.

The staffs guidance for this review includes Information Notice 97-78, Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times, and NUREG-0800, Standard Review Plan, Chapter 18 (draft), Human Factors Engineering.

3.10.2 Human Factors Evaluation The licensee stated that the power uprate will lead to minor changes in several plant parameters, including the 100-percent value for RTP, RCS delta temperature, 1st stage turbine pressure, turbine governor valve positions, SG pressure, and main steam and feedwater flows.

Therefore, the licensee concluded that the proposed power uprate is expected to have a limited effect on the manner in which the operators control the plant during normal operations and transient and emergency conditions.

The licensee stated that there will be minor changes to plant parameters displayed in the control room due to the power uprate. Those parameters that are determined to be outside of their existing indicating bands are addressed within design packages that include plant changes, such as span and scaling, due to the proposed power uprate. With regard to changes to control room controls, displays, and alarms, the licensee stated that a control room alarm is added due to installation of the LEFM Check PlusTM System to indicate conditions that could adversely affect availability of the LEFM Check PlusTM system instrumentation. Operator response to this alarm will be provided within an approved Annunciator Panel Procedure (APP).

The APP will specify the actions required upon a loss of LEFM Check PlusTM System instrumentation. The NRC staff finds that the licensees response is satisfactory because the licensee has adequately identified the changes that will occur with regard to alarms, displays, and controls as a result of the power uprate and adequately described how these changes will be accommodated.

The licensee also stated that necessary changes to Normal and Abnormal Operating Procedures, Emergency Operating Procedures, and Off-Normal Procedures will be made in accordance with the plant-approved process for plant modifications and will be in place prior to use at the uprated power conditions.

Administrative controls to prevent operation above the licensed power level will continue to be provided in Plant Operating Manual procedures. These controls include a statement that steady-state reactor power shall be maintained at or below the licensed power level, and management expectations to minimize (instantaneous) temporary operation above the licensed power level. Plant operating procedures will be revised to indicate that temporary operation above the licensed power level shall be limited to approximately 0.3 percent of RTP, consistent with the reduced power measurement uncertainty prior to operation at uprated power levels.

Instrumentation will automatically actuate to reduce power should licensed power levels be exceeded for any significant duration such that safety limits might be approached. The NRC staff finds that the licensees response is satisfactory because the licensee has adequately addressed the question of operator actions that are sensitive to the power uprate.

The proposed power uprate will require changes to the simulator to ensure that it continues to accurately reflect plant status and physical appearance (hardware), and simulation of plant response (software). These changes will range from simple modifications to process temperatures and flow rates, to plant responses to accidents and transients.

The licensee stated simulator training will be provided on power uprate-related changes to the plant that affect operator performance prior to operating at uprated power levels. Training will also be provided on power uprate-related changes to plant procedures that affect operator performance prior to their use at the uprated power conditions. Changes to the training simulator that are made due to the proposed power uprate will be performed consistent with ANSI/ANS-3.5-1998, and the simulator will be validated in accordance with ANSI/ANS-3.5-1998, Section 4.4, "Simulator Testing."

Hardware and software changes to the simulator are implemented through plant-approved change processes. The hardware and software changes, including changes involving plant process computer inputs that affect operator performance, will be completed prior to operation at the uprated power level. Simulator revalidation is performed in accordance with ANSI/ANS-3.5-1998, Section 4.4, "Simulator Testing."

3.10.3 Human Factors Summary The NRC staff finds that the licensee has satisfactorily addressed these human factors areas associated with the proposed power uprate. The NRC staff further finds that the power uprate should not adversely affect simulation facility fidelity, operator performance, or operator reliability. Therefore, the NRC staff finds the proposed power uprate acceptable with respect to human factors issues.

3.11 TS Changes The licensee submitted to the NRC the proposed TS changes (Refs. 1 and 2) in its support of safe operation of the HBRSEP2 at a maximum power level of 2339 MWt. Following is the NRC staff review of the TS changes.

3.11.1 TS 1.1 - Definition for Rated Thermal Power The TS defines 2339 MWt (increased from 2300 MWt) as the RTP. The TS change is acceptable since the power level of 2339 MWt is considered as the RTP in the acceptable transient and accident analysis.

3.11.2 TS Figure 2.1.1 Decrease in High Nuclear Flux Trip Shown in the Reactor Core Safety Limits Curve The current analysis value for the high nuclear flux trip is 118 percent of the rated power. To reflect the power uprated from 2300 to 2339 MWt, the trip setpoint indicated on TS Figure 2.1.1-1, Reactor Core Safety Limits, is revised from the current value of 118 percent of the rated power to 116 percent of uprated power level. With the revised setpoint, the absolute value (in terms of MWt) for the high flux trip setpoint is slightly less than that used in the supporting analysis for the power uprate application and is conservative. Therefore, the NRC staff concludes that the change is acceptable.

3.11.3 TS Table 3.3.1 Increase in the Value of Tavg Used to Calculate the Overtemperature Delta T (OTDT) and Overpressure Delta T (OPDT) RPS Trips The value of Tavg for the OTDT and OPDT reactor trip equations provided in TS Table 3.3.1-1 is revised from 575.5 oF to 575.9 oF. The TS changes reflect the change in power uprate operating conditions. The licensee indicated that the OTDT and OPDT reactor trips were credited in the analysis of record for two events: the loss of external load event and the uncontrolled RCCA bank withdrawal at power event. As a result of its assessment, the licensee indicated that the TS changes affect the timing of the reactor trip insignificantly. In addition, the analysis of record bounds the effect of the slight increase in Tavg on the calculated minimum DNBR. Based on the licensees assessment results, the NRC staff concludes that the analysis of record remains bounding. Therefore, the TS changes are acceptable.

3.11.4 TS Figures 3.4.3-1 and 3.4.3 Reactor Coolant System Heatup and Cooldown Limitations The licensee proposed to limit the period of applicability of the P-T curves by .04 EFPY of operation from the current 24.0 to 23.96 EFPY. Assuming an 18-month cycle, the additional energy to be produced by the plant due to the power uprate will be 58.5 MWtY while the energy decrease corresponding to .04 EFPY is 93.56 MWtY. Therefore, the proposed adjustment is conservative. This relationship will be similar to the 18-month cycle whether the cycle is 15 or 20 months long.

In addition, the licensee submitted WCAP-15805, the HBRSEP2 capsule X analysis report (Ref. 6). WCAP-15805 was based on a methodology in compliance with the guidance of RG 1.190. The updated fluence values indicate that the existing values are conservative. In Ref. 4, the licensee explained that this is due to: (1) conservatism in the original calculations

that were based on measurement adjustments, and (2) evolving low leakage loadings.

Accordingly, the NRC staff concludes that the existing limits are conservative and provide additional support to the validity of the P-T limit curves for 23.96 EFPY.

Because of the conservatism of the proposed 23.96 EFPY, TS Figures 3.4.3-1 and 3.4.3-2 are acceptable.

3.11.5 TS 3.7.1 - Decrease in the Power Level Associated with the Required Action for an Inoperable Main Steam Safety Valves Operability of all the MSSVs ensures that the secondary system pressure is limited to less than 110 percent of its design pressure during the most severe transient. With less than full MSSVs capacity available, operation may be allowed at reduced power levels. Required Action A.1 of TS 3.7-1, Main Steam Safety Valves (MSSVs), references actions that are tied to a thermal power of < 51 percent RTP. This value is revised from 51 percent of the original RTP of 2300 MWt to 50 percent of the uprated power of 2339 MWt for operating conditions with one inoperable safety valve on any operating SGs and moderator temperature coefficient zero or negative. The proposed operating power level (in terms of MWt) is lower than the value of the current TS, and is more restrictive. Therefore, the TS change is acceptable.

3.11.6 Setpoint allowable value of Engineered Safety Feature Actuation System (ESFAS)

Instrumentation TS Table 3.3.2-1 Function The change in the power uprate operating conditions also requires a change in the setpoint allowable value of Engineered Safety Feature Actuation System (ESFAS) Instrumentation TS Table 3.3.2-1 Function 1.e, "Safety Injection - Steamline High Differential Pressure Between Steam Header and Steamlines." The licensee stated that containment response evaluations show both an overly low value (lower setpoint) and an overly high value (higher setpoint) of "Safety Injection - Steamline High Differential Pressure Between Steam Header and Steamlines." Therefore, the licensee has proposed a band for the setpoint range allowable values from #116.24 and $83.76 psig instead of the current value # 108.95 psig.

In their response of July 25, 2002, the licensee stated that the need for this change arose from a sensitivity analysis performed by Westinghouse using the revised values for the trip setpoint (100 +/- 60 psi). The sensitivity analysis showed that an MSLB at HZP with an electrical bus failure, and an MSLB at 102-percent power with an MFRV failure, result in a slightly higher containment pressure, of the order of 0.03 psig and 0.08 psig, respectively. This resulted from an earlier ESFAS actuation. The licensee stated that this will not cause exceeding the current design value of containment design pressure of 42 psig. (See also section 3.9.) According to Westinghouse, the greatest sensitivity to the steamline and steam header high differential pressure signal lies with the AFW start time, and since steamline breaks are sensitive to main feedwater and auxiliary feedwater delivery rates and times, it was clear that AFW start time would have an effect. Due to the conclusions of this sensitivity analysis, the licensee stated that the containment analysis has been revised to assume a bounding actuation setpoint of 100 +/- 60 psi. Since the proposed TS upper bound of 116.24 psig and lower bound of 83.76 psig are bounded by the containment analysis assumptions, and also the proposed setpoint band is conservative (proposed ranges from 83.76 to 116.24 as compared to the present range from 0 to 108.95 psig), the NRC staff found this TS change to be acceptable (See section 3.9.1).

The licensee stated that the proposed changes to the instrumentation setpoint allowable values were developed using HBRSEP2 engineering design procedure EGR-NGGC-0153, "Engineering Instrument Setpoints." The licensee also stated that EGR-NGGC-0153 is based on ISA-S67.04, Part 1, 1994, "Setpoints for Nuclear Safety-Related Instrumentation," which is endorsed by NRC RG 1.105, Rev. 3, "Setpoints for Safety-Related Instrumentation." The Instrumentation and Controls Section of EEIB, NRR, found the setpoint methodology used to determine the setpoint allowable value to be acceptable, while the Containment Performance Section, SPLB, NRR, found the rest of the analysis in terms of containment spray actuation setpoint and peak containment pressure to be acceptable at the uprated power. The NRC staff, therefore, finds the proposed TS changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes a surveillance requirement. 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 (67 FR 56319). 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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: S. Sun M. Mitchell B. Marcus B. Fu C. Lauron J. Hayes N. Trehan K. C. Chang L. Lois R. Subbaratnam Date: November 5, 2002

7.0 REFERENCES

1. Letter from B. L. Fletcher III (CP&L) to NRC, H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Request for Technical Specifications Changes to Increase Authorized Reactor Power Level, May 16, 2002.

2. Attachments III and IV to Reference 1, A Markup of the Affected TS Pages and Retyped Pages for the Proposed TS.

3. Letter from B. L. Fletcher III (CP&L) to NRC, H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Response to Request for Additional Information on Amendment Request to Increase Authorized Reactor Power Level (TAC NO. MB5106), July 25, 2002.

4. Letter from B. L. Fletcher III (CP&L) to NRC, H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Response to Request for Additional Information on Amendment Request to Increase Authorized Reactor Power Level (TAC NO. MB5106), September 6, 2002.

5. Letter from B. L. Fletcher III (CP&L) to NRC, H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Supplement to Request Regarding Increase of Authorized Reactor Power Level, August 12, 2002.

6. WCAP-15805, Analysis of Capsule X from the Carolina Power and Light Company H.B.

Robinson Unit 2 Reactor Vessel Radiation Surveillance Program by T. J. Laubham et al., Westinghouse Electric Company LLC, dated March 2002.

7. Memorandum from L. J. Callan (NRC) to the Advisory Committee on Reactor Safeguards (NRC), Agency Program for High Burnup Fuels, dated July 6, 1998.

8. Westinghouse WCAP-15628, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for H. B. Robinson Unit 2 Nuclear Power Plant for the License Renewal Program, July 2001.

9. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 3.9.3, ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures.

10. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials.

11. Attachment to Reference 9, Fracture Toughness Requirements.

12. NUREG-1061, Volume 3, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks.

13. American Society for Testing and Materials Standard Practice E 185, Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

14. Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessel with Charpy Upper-Shelf Energy Less Than 50 Ft-Lb.

15. FNP-RNP-300, Rev. 6, 10 CFR 50 Appendix R Section III G Safe Shutdown Component/Cable Separation Analysis.

16. Caldon Engineering Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Using the LEFM Check System.

17. Caldon Engineering Report ER-157P,Supplement to Topical Report ER-80P: Basis for a Power Uprate with LEFMUTM or LEFM CheckPlusTM System.

Table 1 - Summary - Non-LOCA and LOCA Transients Analyses Event Description UFSAR Bounding Does the power uprate Disposition of the Accident/Transient analysis Staff conclusion Event Reference affect the previous analysis ?

FWS Malfunctions that 15.1.3 No Analysis in the UFSAR remains valid Accept decrease FW Temperature FWS Malfunctions that 15.1.3 and 15.4.1 No RCCA bank withdrawal from a subcritical or Accept increase FW Flow low-power condition UFSAR bounds Increase in Steam Flow 15.1.3 Yes Previously Approved through TS Amendment Accept No. 154 Inadvertent Opening of a 15.1.3 and 15.1.5 No, severity of the Bounded by the excess load event or by Accept SG Relief or PORV accident is not increased MSLB after Rx trip Main Steamline Break 15.1.5 No effect on the hot zero- UFSAR results of analysis for the MSLB Accept (MSLB) power conditions event at hot zero-power conditions remain valid Loss of External Electric 15.2.2 The effect of the power Small increase in RCS average temperature Accept Load, Turbine Trip, Loss uprate on the timing of from 575.5 oF to 575.9 oF has an insignificant of Condenser Vacuum and reactor trip is effect Other Events Resulting in insignificantly small Turbine Trip, or Inadvertent Closure of MSIV Loss of Non-Emergency 15.2.2, 15.3.1,15.2.6 No Previously accepted through TS Amendment Accept AC Power to the Station and 7 87 and UFSAR analysis remains valid Auxiliaries Loss of Normal Feedwater 15.2.7 Yes Previously accepted UFSAR analysis remains Accept valid Feedwater System Pipe 15.2.8 as bounded by Changes resulting from Previously accepted through TS Amendment Accept Break 15.1.5 the power uprate do not 87 and UFSAR analysis remains valid affect the severity of the feedwater line break Loss of Forced Reactor 15.3.1 Yes Small increase in RCS average temperature Accept, UFSAR analysis Coolant Flow from 575.5 oF to 575.9 oF has an insignificant remains bounding effect Reactor Coolant Pump 15.3.2 Yes Small changes in core operating conditions Accept, UFSAR analysis Shaft Seizure will not affect the fuel failure estimate remains bounding

Event Description UFSAR Bounding Does the power uprate Disposition of the Accident/Transient analysis Staff conclusion Event Reference affect the previous analysis ?

Reactor Coolant Pump 15.3.2 Yes Bounded by the locked rotor event Accept Shaft Break Uncontrolled RCCA Bank 15.4.1 No Uprate has no effect on the core power Accept, UFSAR analysis Withdrawal from assumption. Input parameters have remains bounding Subcritical or Low Power negligible effect on transient Uncontrolled RCCA Bank 15.4.2 Yes Small increase in RCS average temperature Accept Withdrawal at Power from 575.5 oF to 575.9 oF has an insignificant effect and timing of reactor trip is insignificantly affected Single Rod Cluster Control 15.4.3.1 Yes The small changes in core operating Accept. The staff Assembly (RCCA) conditions as a result of the power uprate will concludes that the analysis Withdrawal at Full Power not affect the fuel failure estimate of record for this event remains valid and acceptable for the uprated power.

Rod Cluster Control 15.4.3.2 and .3 Yes Small increase in RCS average temperature Calculated DNBRs and fuel Assembly (RCCA) from 575.5 oF to 575.9 oF has an insignificant centerline temperatures are Misalignment effect on the center line melting temperature not adversely affected by the power uprate conditions. Accept since 15.4.3.2 and .3 remain bounding for uprate.

Startup of an Inactive 15.4.4 No This event is not analyzed. Operation in this Maintain current Reactor Coolant Loop at configuration is not permitted status-quo an Incorrect Temperature Chemical Volume Control 15.4.6 No Existing Basis remains unchanged Accept System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant

Event Description UFSAR Bounding Does the power uprate Disposition of the Accident/Transient analysis Staff conclusion Event Reference affect the previous analysis ?

Inadvertent Loading and 15.4.7 Yes Small increase in RCS average temperature Accept Operation of a Fuel from 575.5 oF to 575.9 oF has an insignificant Assembly into the effect on the thermal properties of the RCS Improper Position primary and secondary sides. Previously accepted through TS Amendment 141 and UFSAR analysis remains valid RCCA Ejection Accident 15.4.8 No Previously approved part of TS Amendment Accept No. 141 and bases are still valid Inadvertent Operation of 15.5.1 No The UFSAR conclusions will not be changed Accept Emergency Core Cooling by the power uprate System CVCS Malfunction that 15.5.2 Yes Bounded by the decrease in boron Accept Increases Reactor Coolant concentration event. Does not affect the Inventory severity of the event relative to the bounding cases Inadvertent Opening of a 15.6.1 No None of these parameters are affected by the Accept Pressurizer Safety or power uprate Power Operated Relief Valves Small-break LOCA 15.6.2 Yes The UFSAR analysis remain valid for the Accept power uprate Steam Generator Tube 15.6.3 Yes Parametric values studied remain Accept Rupture conservative relative to the uprate condition Large-Break Loss-of- 15.6.5 Yes Limiting cases of large-break LOCA events Accept Coolant Accident show that the results satisfy the acceptance criteria of 10 CFR 50.46

Mr. J. W. Moyer H. B. Robinson Steam Electric, Carolina Power & Light Company Plant, Unit No. 2 cc:

Mr. William D. Johnson Mr. C. T. Baucom Vice President and Corporate Secretary Supervisor, Licensing/Regulatory Programs Carolina Power & Light Company Carolina Power & Light Company Post Office Box 1551 H. B. Robinson Steam Electric Plant, Raleigh, North Carolina 27602 Unit No. 2 3581 West Entrance Road Ms. Karen E. Long Hartsville, South Carolina 29550 Assistant Attorney General State of North Carolina Ms. Beverly Hall, Acting Director Post Office Box 629 N.C. Department of Environment Raleigh, North Carolina 27602 and Natural Resources Division of Radiation Protection U. S. Nuclear Regulatory Commission 3825 Barrett Dr.

Resident Inspectors Office Raleigh, North Carolina 27609-7721 H. B. Robinson Steam Electric Plant 2112 Old Camden Road Mr. Robert P. Gruber Hartsville, South Carolina 29550 Executive Director Public Staff - NCUC Mr. P. T. Cleary 4326 Mail Service Center Plant General Manager Raleigh, North Carolina 27699-4326 Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Mr. Virgil R. Autry, Director Unit No. 2 South Carolina Department of Health 3581 West Entrance Road Bureau of Land & Waste Management Hartsville, South Carolina 29550 2600 Bull Street Columbia, South Carolina 29201 Mr. Chris L. Burton Director of Site Operations Mr. Terry C. Morton Carolina Power & Light Company Manager H. B. Robinson Steam Electric Plant, Performance Evaluation and Unit No. 2 Regulatory Affairs CPB 7 3581 West Entrance Road Carolina Power & Light Company Hartsville, South Carolina 29550 Post Office Box 1551 Raleigh, North Carolina 27602-1551 Public Service Commission State of South Carolina Mr. John H. ONeill, Jr.

Post Office Drawer 11649 Shaw, Pittman, Potts, & Trowbridge Columbia, South Carolina 29211 2300 N Street NW.

Washington, DC 20037-1128 Mr. B. L. Fletcher III Manager Regulatory Affairs Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550-0790