ML20087B573

Safety Evaluation Supporting Amends 203 & 203 to Licenses DPR-32 & DPR-37,respectively

ML20087B573
Person / Time
Site:	Surry
Issue date:	08/03/1995
From:	NRC (Affiliation Not Assigned)
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ML20087B571	List: ML20087B566 ML20087B573
References
NUDOCS 9508080184
Download: ML20087B573 (32)

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UNITED STATES y-NUCLEAR REGULATORY COMMISSION E

f WASHINGTON, D.C. 2066H001 p'

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 203 TO FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 203 TO FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated August 30, 1994, as supplemented by letters dated February 6, February 13, February 27, March 23, March 28, April 13, April 20, April 28, May 5, and June 8,1995, the Virginia Electric and Power Company (VEPC0 cr the licensee) submitted a request to revise Facility Operating Licenses Nos. DPR-32 and DPR-37 for the Surry Power Station, Unit Nos. I and 2, respectively, to increase the currently licensed core power level of 2441 megawatts thermal (MWt) to an uprated core power level of 2546 MWt. The amendment application also submitted a number of changes to the Technical Specifications (TS) to impleme7t uprated power operation. The supplemental letters provided clarification or amplification of the analyses in the August 30, 1994, submittal and were not outside the scope of the original Federal Reaister notice.

2.0 DISCUSSION Surry Units 1 and 2 are currently licensed for operation at a reactor core power level of 2441 MWt.

The licensee undertook a program to uprate the Surry units to a maximum reactor core power level of 2546 MWt, approximately a 4.3%

increase. At the core uprate power, the generator gross output for each unit, based on the predicted heat balance, will be about 852.3 megawatts electrical (MWe). The engineering studies supporting the core uprate have been performed in accordance with Westinyhouse Topical Report WCAP-10263, entitled "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor" and dated January 1983.

The licensee's letter of August 30, 1994, submitted a " Licensing Report for Operation with a Core Rated Power of 2546 MWt, Surry Power Station Units 1 and 2" which provided supporting documentation and analyses for the proposed changes.

The licensing basis assessment included a review of the accident analyses, component and system design, Emergency Operating Procedures; TS, and appropriate sections of the Updated Final Safety Analysis Report (UFSAR). The report evaluated the ability of plant systems, structures and components to operate within safe limits, during both normal and accident conditions. The scope of the licensee's review to support the proposed power uprating encompassed all aspects of the Surry Nuclear Steam Supply System (NSSS) design and operation affected by the increase. NSSS designs were reviewed to verify

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2 compliance at the increased power rating with licensing criteria and standards currently specified in the Surry operating licenses.

In accordance with 10 CFR 50.59, the licensee also identified potential unreviewed safety questions that may occur as a result of the increased power rating. The structural design of equipment was reviewed to assure that compliance would be maintained at the increased power rating with industrial codes and standards that applied when the equipment was originally built.

In addition, the review verified that NSSS components and systems will continue to meet functional requirements specified in the UFSAR at the increased power rating.

Currently approved analytical techniques were used for analyses at the increased power rating.

Also, the design of NSSS/ Balance of Plant (B0P) safety-related interfaces were reviewed for any impact at the increased power rating.

Based on the scope of the review as outlined above, the licensee states that the Surry units are capable, in their present design configuration, of operating at the proposed core power rating of 2546 MWt and an NSSS power rating of 2558 MWt without violating any of the design criteria or safety limits specified in the Surry UFSAR and currently required in Facility Operating Licenses DPR-32 and DPR-37.

In addition to evaluating the ability of the plants to perform at the new power level under steady state conditions, the licensee reevaluated all the design basis transients and accidents that the NRC staff uses to determine that adequate safety margins are maintained.

These analyses were performed by the licensee using staff-approved computer codes.

Those events which might challenge the departure from nucleate boiling ratio (DNBR) limits were evaluated using the Westinghouse WRB-1 critical heat flux correlation approved for the licensee's licensing use. ' Steady state instrument errors were considered in establishing the initial conditions, and 2% was added to the initial power to account for calorimetric error.

3.0 EVALUATION 3.1 Nuclear Desian and Core Thermal-Hydraulic Desian 3.1.1 Core Desian The staff issued License Amendment 116, dated January 6, 1988, allowing the Surry Units to use Surry Improved Fuel (SIF) in the Surry cores. The supporting analyses for this amendment accounted for the incremental replacement of the resident Westinghouse Low Parasitic fuel and considered the varied mixtures of the fuel types during the transitional period.

r The licensee currently uses an approved method of analysis, the Virginia Power Statistical DNBR Evaluation, to determine the departure from nucleate boiling.

The proposed design parameter changes after core uprating implementation were evaluated with respect to those assumed in the Statistical DNBR Evaluation Methodology.

The licensee determined that the method of analysis remains valid.

Likewise, the licensee also verified the relevant thermal-hydraulic items considered on a reload basis that could be affected by core uprating which are i

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i evaluations of core bypass flow rate, core thermal limits, axial power J

distribution effects and retained DNBR margin. The core thermal limits were the only items affected by the uprate because a power increase increases the total temperature rise across the core. The licensee stated that these effects are modelled in the reload core design and that new thermal limits for the uprated condition were generated. The staff finds the licensee's findings acceptable.

3.1.2 Reactor Coolant System (RCS)

The licensee's submittal proposes to increase the steam generator tube plugging limit from 0% to 7%, which was assumed in the licensae's NSSS and component evaluations for the power uprate analyses. The NSSS accident analyses and core thermal hydraulic assessments assume 15% tube plugging. The i

staff concludes that the lower flow is acceptable for the uprated power because it is considered in the technical justifications for the power uprating.

In its discussion of RCS topics, the licensee did not identify any other items that could affect the power uprating. We find that the adequacy of the RCS design is not affected by the power uprating.

3.1.3 Overoressure Protection It is required that pressurizer safety valves be designed with sufficient capacity to prevent the pressurizer pressure from exceeding 110% of design pressure following the maximum expected RCS pressure transient.

For purposes of analytical justification, this event is specified to be a 100% load rejection resulting from a turbine trip with concurrent loss of main feedwater. No credit is taken for operation of the RCS relief valves, steam line relief valves, steam dump system, pressurizer level control system, or direct trip on turbine trip. Reactor scram is initiated by the first safety-grade signal from the reactor protection system.

By letter dated May 5, 1995, the licensee documented the existing RCS overpressure analysis for Surry and verified the continued applicability of the existing analysis for operation at uprated conditions.

The licensee stated that Westinghouse WCAP-7769, Revision 1, " Overpressure Protection for Westinghouse Pressurized Water Reactors," June 1972, remains valid for the Surry units in the uprated conditions.

In its uprating report, the licensee also describes it's analyses, which agree with the WCAP-7769 cunclusions regarding the adequacy of the sizing of the Surry safety valves. The limiting event was identified as the Complete Loss of External Electrical Load and is documented in Section 3.5.8 of the Surry Core Uprating Licensing Report. The results of these analyses demonstrate that 94% of the total pressurizer safety valve capacity is required to mitigate the peak pressure of 2745 psia, which occurs in the cold leg 10.2 seconds into the event.

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Based on the above information, the staff concludes that, for operation at powers up to the proposed uprated power, the pressurizer safety valve capacity at Surry is adequate to meet the original Surry licensing requirements.

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3.1.4 Auxiliary Feedwater and Residual Heat Removal The staff review and approval of the Surry auxiliary feedwater system (AFW) l was granted by letters dated November 17, 1980, and April 27, 1992.

The transients that identify the limiting single-failure and minimum flow requirements for Surry's AFW system are loss of all AC to station auxiliaries and loss of normal feedwater. The existing analyses for the these events assumed the uprated power conditions and are included in the Surry Core Uprate Licensing Report, Sections 3.4.3 and 3.8.1.

The licensee stated that, as a result of the analyses, the limiting scenarios for single failure and required flow were not changed. The staff concludes that, since the Surry AFW flow capacity exceeds cooling requirements for the uprated power, cooldown time to residual heat removal (RHR) cut-in conditions would not be significantly affected.

The RHR System cooldown analysis was not significantly affected by the uprating.

The licensee indicated that the system still has the ability to bring the plant to cold shutdown (<;200*F) in the uprated condition. With two RHR pumps in operation, cooldown to 140*F can be achieved 11.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after shutdown. With one pump in operation, cooldown to 200*F can be achieved in 21.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The staff agrees that the RHR system can continue to perform its intended function in the uprated condition.

3.1.5 Emeraency Core Coolina System (ECCS)

The licensee's submittal identifies no adverse impact to ECCS operability or vulnerability to single failure resultant from the power uprating.

ECCS performance analyses were evaluated using the Westinghouse large break loss-of-coolant accident (LBLOCA) 1981 Evaluation Model (EM) with BASH. The small break analyses were performed with the Westinghouse NOTRUMP small break LOCA (SBLOCA) EM.

Both were approved for use by the licensee at the uprated power and increased steam generator tube plugging limit and demonstrate conformance with the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K, respectively.

The licensee stated that the current LBLOCA analysis, performed in March 1994, assumes the uprated power and 15% steam generator tube. plugging and results in a calculated peak cladding temperature (PCT) of 2120*F and metal-water reaction levels of less than 1.0% core-wide and 8.67% local.

These values are within the limits specified in 10 CFR 50.46(b) (1 - 3) of 2200*F, less than 1%

core-wide metal-water reaction and 17% local metal-water reaction, respectively, and assure that the core would remain amenable to cooling as required by 10 CFR 50.46(b)(4). Meeting the long-term cooling requirement of 10 CFR 50.46(b) is assured for the uprated power by the continued acceptability of the Surry ECCS design.

5 The SBLOCA analyses for the uprated power with 15% steam generator tube plugging calculated a PCT of 1852*F, with less than 1.0% core-wide metal-water reaction and 3.2% local metal-water reaction. The results meet the requirements of 10 CFR 50.46(b) (1 - 3) and are bounded by the LBLOCA analysis results.

Satisfaction of 10 CFR 50.46(b)(4) and (5) for SBLOCA analyses is similar to that for LBLOCA analyses. The staff concludes t::at the ECCS analyses provided in support of the power uprate are in compliance with 10 CFR 50.46 and Appendix K.

The Surry ECCS design is therefore adequate for the uprated power.

3.1.6 Transient Analyses VEPC0 indicated that they reanalyzed or reevaluated all Surry UFSAR Chapter 14 events considering the uprated power. They concluded that certain events did not require reanalyses either because (1) the events do not apply to the plant's present licensed configuration (i.e., Malpositioning of Part-Length Control Rod Assemblies, Startup of Inactive Reactor Coolant Loop) or (2) because existing analyses address the events for uprated conditions (i.e.,

Turbine-Generator Overspeed, Main Steam Line Break, Excessive Heat Removal Due to Feedwater System Halfunctions, Loss of Normal Feedwater, Control Rod Assembly Ejection, SBLOCA, and LBLOCA).

The licensee justified not reanalyzing by showing that assumptions for the uprated conditions would not significantly affect the existing analyses or that the existing analyses assumed the uprated power.

Based on their assessment, the licensee concluded that reanalyses of Uncontrolled Control Rod Assembly Withdrawal from a Subcritical Condition and at Power, Control Rod Assembly Drop / Misalignment, Chemical and Volume Control System Malfunction, Excessive Load Increase, Loss of Reactor Coolant Flow, Locked Rotor, Loss of Electrical Load, and Steam Generator Tube Rupture events were warranted.

The licensee reanalyzed these events using currently approved methodologies. We find the licensee's reasons and conclusions acceptable.

3.1.6.1 Uncontrolled Control Rod Assembly Withdrawal From a Subcritical Condition The licensee reanalyzed the rod withdrawal from subcritical event using the RETRAN computer code and the associated Virginia Power reactor system transient analysis methodology.

RETRAN calculates nuclear power, core heat flux, average fuel, and cladding and coolant temperatures. The detailed core 1

thermal-hydraulics analysis was performed using the COBRA computer code to generate the minimum departure from nuclear boiling ratio (MDNBR). The major difference between the previous analysis and the current reanalysis is that the licensee assumes that all three reactor coolant pumps (RCPs) are in operation.

The licensee determined that the core and the RCS are not adversely affected because the peak thermal core and coolant temperatures in the departure from nuclear boiling (DNB) limiting case are well below their nominal full power values. The staff finds the licensee analysis and conclusion acceptable.

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6 3.1.6.2 Uncontrolled Control Rod Assembly Withdrawal at Power The rod assembly withdrawal at power event required reanalysis due to the change in the Overtemperature and Overpower oT protection setpoints.

The event was reanalyzed using RETRAN and COBRA computer codes.

The licensee concluded that (1) the DNBR remains above the 95/95 DNBR design limit, (2) the most limiting RCS pressure is below 110% of design pressure, and (3) the most limiting main steam pressure is less than 110% of the main steam design pressure. Therefore, the staff finds that, in the uprated condition, the Surry plants are able to mitigate the consequences of the uncontrolled control rod assembly withdrawal at power event.

3.1.6.3 Control Rod Assembly DroD/Misalianment The control rod assembly drop / misalignment event required reanalysis at the uprated condition to generate new plant-specific core thermal limit lines for reload analyses.

These new limits were generated by completing the transient and thermal-hydraulic portion using staff-approved methodology. The nuclear analysis will be completed when the reload design calculations are completed for the uprated core.

The staff finds the licensee's analysis acceptable.

3.1.6.4 Chemical and Volume Control System Malfunction The chemical and volume control system (CVCS) malfunction event centers around boron dilution accidents.

Boron dilution is a manual operation that is under strict administrative controls limiting rate and duration of dilution.

It was necessary to reanalyze the event because the uprated power affects the Overtemperature and Overpressure oT setpoints.

The licensee reevaluated the administrative controls and alarms to ensure that there remains at least a minute margin from positive indication of a dilution in progress to loss of shutdown margin for corrective operator action during MODES 1 through 4.

For MODES 5 and 6 the licensee ensured that primary grade water will be isolated from the RCS 15 minutes after a planned dilution.

The licensee determined that the existing alarms and administrative controls still allow enough time for operator intervention in a credible boron dilution event. The staff finds the licensee's findings acceptable.

3.1.6.5 Excessive load Increase The limiting scenarios for an excessive load increase event are initiated at full power; therefore, it was necessary to reanalyze the event for the uprated condition. The analysis was completed using LOFTRAN and Westinghouse standard non-statistical thermal-hydraulic methodology.

Four cases were analyzed to demonstrate plant response following a 10% step load increase.

The cases included the reactor at the beginning-of-life, manually and automatically controlled and end. af-life and manually and automatically controlled.

In all cases the minimum DNBR remained above the design limit value.

Therefore, the staff finds the plant response to excessive load increase in l

J the uprated condition acceptable.

7 3.1.6.6 Loss of Reactor Coolant Flow The loss of reactor coolant flow is the limiting DNB event and was therefore reanalyzed in the uprated condition.

The analysis includes two cases for loss of reactor coolant flow, loss of three out of three RCPs, from 100% power, due to (1) an undervoltage condition and (2) a frequency decay condition.

The analysis included a transient simulation using RETRAN for thermal-hydraulic plant response.

The results were used as input into the COBRA computer code for a detailed thermal-hydraulic analysis to compute the DNB margin. The licensee reported that both cases resulted in minimum DNBRs with considerable margin to the DNBR limit value.

Therefore, the staff finds the plant response to loss of reactor coolant flow acceptable in the uprated condition.

3.1.6.7 Locked Rotor The locked rotor event is characterized by the rapid loss of circulation in one Reactor Coolant loop due to the seizure of the RCP. The plant response is simulated by using the RETRAN transient analysis code and the COBRA IIIC/MIT detailed thermal-hydraulics.

In this analysis, the unaffected RCPs were assumed to trip, on low coolant flow, two seconds after generation of the reactor trip signal. The licensee stated that this assumption is consistent with the RCP trip assumption made in the locked rotor analysis for the uprating of its North Anna plants (approved in an NRC Safety Evaluation report dated August 25, 1986). The licensee stated that in the North Anna analysis no rods had an MDNBR less than the statistical DNBR design limit and the RCS and main steam peak pressures remained below 110% of design pressure. The staff finds these results acceptable.

3.1.6.8 Loss of Electrical Load The complete loss of electrical load was analyzed with the core characteristic of the conditions at beginning of cycle and with power at 102% of the uprated core power.

The licensee used RETRAN and COBRA computer codes to simulate the plants' response,to the transient. The licensee stated that the results of the analysis were (1)'the peak RCS and main steam pressures remained below the associated design pressures and (2) the MDNBR remained above the 95/95 DNBR design limit.

The staff finds that Surry in the uprated condition can still mitigate the consequences of a complete loss of electrlcal load.

3.1.6.9 Steam Generator (SG) Tube Ruoture The SG tube rupture event was analyzed for consequent radiological dose using staff-approved methodologies.

The thermal-hydraulic component of the accident was simulated with RETRAN.

The radiological dose was calculated using an NRC-approved methodology and the results were within the 10 CFR 100 limits.

A significant aspect of an SG tube rupture scenario is the immediate cooldown, I

from full power conditions, to pressure and temperature equilibrium between

8 the primary system and the shell side of the ruptured SG. The cooldown is accomplished through the use of the unruptured SGs fed by the auxiliary feedwater system (AFS). As identified in the discussion of the AFS, the staff concludes that the flow capacity of the AFS continues to exceed cooldown requirements for the uprated power, and that the cooldown times would not be significantly affected. Therefore, the conclusions for the SG tube rupture event for the present licensed power of 2441 MWt continue to apply for the uprated power. The licensee's analyses confirm this conclusion for the Surry units.

The licensee's uprating submittal assesses the impact of the power uprate on the results of the existing approved UFSAR Chapter 14 analyses. The transient analyses supporting the power uprating assumed an SG plugging level of 15%.

The staff reviewed these analyses and concluded that appropriate safety criteria were met.

Based on the above, the staff finds the referenced accident analyses acceptable to support operation at the uprated power with a limit of 7% SG tube plugging.

3.2 Systems. Structures. and Components Evaluation The staff reviewed the licensee's evaluations of the proposed power uprate on the structural and pressure boundary integrity of the NSSS and the BOP system, including the piping, components, their supports, the reactor pressure I

vessel (RPV), core support structures (CSSs), reactor vessel internals (RVIs),

SGs, control rod drive mechanisms (CRDMs), RCPs, and pressurizer.

As discussed in the following subsections, the staff finds that calculated stresses and fatigue usage factors as stated by the licensee remain within ine Code-allowable limits, and concludes that the proposed power uprate does not have any significant impact on the structural and pressure boundary integrity l

of the NSSS and B0P systems.

3.2.1 Reactor Vessel The proposed power uprate increase of 105 MWt is approximately 4.3% over the currently licensed level. The licensee reported that the power increase will result in a change in the RCP/RVIs inlet temperature.

The outlet temperature remains unchanged. There were no significant changes in thermal transients and LOCA blowdown forces as a result of the power uprating. The RCP/RFI inlet temperature change is a reduction from 543.0*F to 540.4*F. The licen',ee evaluated the design and operation of the regions of the reactor vessel affected by the temperature change and fluence, based on the proposed uprated core power. The evaluation included a review of the reactor vessel design specifications, stress report', and fracture mechanics analyses.

The regions of the reactor vessel affected by the temperature change include l

the RPV (main closure head flange, studs, and vessel shell), CRDM nozzles, core support pads, and the instrumentation tubes. The evaluation of the maximum ranges of stresses and cumulative fatigue usage factors was performed for the critical components at the core power uprated conditions. The l

9 licensee found that the core power uprate has insignificant effect on the components ih the affected regions of the reactor vessel. The staff agrees with the licensee's conclusion that the maximum stres:es and cumulative l

fatigue usage factors, shown in Tables 4.1.1-1 and 4.1.1-2, respectively, of the August 30, 1994 submittal, remain within the allowabic American Society of Mechanical Engineers (ASME) Code limits.

As documented in Topical Reports BAW-2192 dated December 1993 and BAW-2178 dated April 1993, the staff approved the licensee's analysis that demonstrated equivalent margins of safety to those represented by the requirements of 10 CFR 50, Appendix G and the ASME Code for the material's upper shelf energy.

The staff has reviewed the licensee's assessment and concludes that the integrity of the reactor vessel will not be significantly affected by the new operating conditions since the analyses remain valid at the uprated core power level.

3.2.2 Reactor Core Soonort Structures and Vessel Internals By letters dated February 13, 1995, and April 20, 1995, the licensee provided the additional information requested by the staff, with regard to the evaluation of the reactor vessel core support and internal structures. The licensee reported that the effect on the core support and internals structures caused by the power uprate represents less than a 3*F reduction in the core inlet temperature and a less than 5% increase in total heat generated in the core barrel, baffle, thermal shield, and upper and lower core plates.

The licensee's evaluations of the components affected by the temperature reduction and increased heat generation demonstrated that the effects were negligible.

Further, the licensee concluded there is no concern with regard to flow-induced vibration problems, based on the experience of several other similarly designed three-loop reactors that are licensed and are successfully operating at higher core powers. The licensee and its NSSS vendor verified that the design criteria, including stress limits, applied to the original CSS and RVI remain valid at the proposed uprated conditions.

3.2.3 Control Rod Drive Mechanisms (CRDMs)

The licensee evaluated the adequacy of the CRDMs by reviewing the Surry j

current CRDM design specifications and stress report to compare the design l

basis input parameters with the operating conditions at the uprated core power.

The evaluation concluded that the original desjgn basis thermal and structural analyses are bounding for the core power uprate.

Based on its review, the staff concurs with the licensee's conclusion, for the power uprated conditions, that the current design of CRDMs continues to be in compliance with the codes and standards under which the plant was licensed.

3.2.4 Steam Generators (SGs)

The licensee's analysis of the effects of the uprate on the SGs was divided

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into three major areas:

structural, thermal-hydraulic, and corrosion i

evaluations.

In response to staff inquiries, VEPC0 provided additional

10 information regarding the SG aspects of the uprate in submittals dated March 23, 1995, April 13, 1995, and June 8, 1995.

The licensee analyzed the Westinghouse Model 51F SGs for the proposed uprate conditions. The uprated SGs differ from the current conditions in the following parameters:

steam pressure decreases only 1 pound per square inch (PSI), feed temperature increases 5.5*F, and steam flow increases 5.6% to 11.26 million pounds per hour (7% plugging assumed), while steam temperature is essentially constant.

For maintaining tube structural integrity, the licensee stated that the tube plugging criteria will be based, in part, on the limit in the ASME Boiler and Pressure Vessel Code (ASME Code)Section XI, Paragraph IWB-3521, which states that the depth of a flaw shall not exceed 40% of the tube wall thickness.

This limit was calculated based on a generic evaluation of Westinghouse and Combustion Engineering SGs.

The licensee stated that this generic evaluation is applicable to their SG.

The licensee also performed a calculation in accordance with ASME Section III, Paragraph NB-3324.1, which demonstrated that the tube minimum wall thickness for the uprated conditions is 0.020-inches (40%) using the margin of safety of three from Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes."

Including a 20% wall thickness allowance (0.010-inches) for eddy current uncertainty and potential flaw growth, the tube repair limit was verified to be 40%.

This tube repair limit is consistent with the current repair limit in the TS.

Concurrent with the increased steam flow associated with the power uprate is an increase in the void fraction in the U-bend region. Although this increase may increase the potential for vibrations leading to mechanical degradation (e.g., tube wear), the licensee indicated this would not affect the long-term integrity of the Model 51F SGs. To date, the Surry units have experienced only minor wear at the anti-vibration bars (AVBs), resulting in five tubes being plugged in Unit I and two tubes in Unit 2.

Although the increased potential for tube wear at the AVBs associated with power uprate is small, the licensee committed to notify the NRC of any significant increase in the number of wear-related indications found during the three refueling outages after the power uprate.

The licensee indicated that there are no changes to the design transient as a result of the core power uprate. The stress levels and the cumulative fatigue usage factor's of the critical SG components continue to remain in compliance with the requirements of the ASME Code,Section III through the Winter 1976 Addendum.

The licensee valuated the potential for SG corrosion with the power uprate.

The licensee previously completed chemical cleaning on Units 1 and 2 to remove corrosion product buildup.

Improved chemistry controls and removal of copper from the secondary plant are expected to significantly reduce deposits. The proposed power uprate should have minimal adverse effect on corrosion product buildup.

Based upon the information provided, the staff concludes that the licensee's analysis of SG performance for operation at the proposed power uprate

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demonstrates an acceptable level of quality _and safety. The licensee's

commitment to notify the NRC-(in the SG plugging report required by TS 4.19F)

'l of any significant increase in the number of wear-related indications found during the'three refueling outages after the power uprate provides additional assurance that a~ subsequent increase in wear-related indications will be i

detected and its significance assessed.

The staff concludes that the current Surry Model 51F SGs are acceptable for the proposed power uprate.

3.2.5 Reactor Coolant Pumos I

i The licensee evaluated the RCPs by comparing the design specifications with-the proposed uprated conditions. At the core power uprate, the RCS pressure remains unchanged. There are no changes to the design thermal transients.

j The small reduction (3*F) in the RCP inlet temperature has insignificant effect on the pressure boundary stresses.

Based on its review, the staff concurs with the licensee's conclusion that the current design of the RCPs, when operating at the proposed power uprated conditions, will remain in l

compliance with the requirements of the codes and standards under which thc Surry units were originally licensed.

l 3.2.6 Pressurizer The licensee evaluated the adequacy of the pressurizer and components including the pressurizer spray nozzle, safety and power operated relief

'l valves, the pressurizer surge line, safety valve discharge piping and the pressurizer relief tank for operation at the'uprated conditions. The i

evaluation was done by reviewing the existing Surry pressurizer stress report and design basis analyses of the pertinent pressurizer components and piping.

1 The licensee found that the uprate conditions are bounded by those used in the original pressurizer analyses. Based on its review, the staff concurs with the licensee's conclusion that the existing pressurizer and components remain l

adequate for the proposed core power uprate.

1 3.2.7 NSSS Pioina and Pipe Sucoorts i

The proposed power uprate of the Surry station will increase the temperature difference across the RCS. This includes a reduction of the RCS cold leg

. temperature (from 543.0*F to 540.4*F) while the hot leg temperature remains unchanged. The RCS loop pressure will not change as a, result of tha proposed core power uprate. At Surry, the existing design basis thermal and fatigue analyses ~ of the NSSS system piping and supports were reviewed by the licensee,-

in comparison with the uprated power conditions, with respect to the design system parameters and transients. The licensee concluded that the existing i

~ design basis stress analyses for the RCS system piping and supports and systems connecting to the RCS system remain valid for the power uprated conditions. The staff finds that the reduction (less than 3*F) of the cold leg temperature and the increase of the temperature difference across the RCS system, which affect the NSSS piping, are too insignificant to impact the design basi's analysis of the piping and pipe support. Therefore, the existing NSSS piping and supports, the primary equipment nozzles, the primary equipment

12 supports, and the branch lines connecting to the primary loop piping will remain in compliance-with the requirements of the design bases criteria as defined in the UFSAR, and are acceptable for the power uprate.

4 3.2.8 B0P Pioina The licensee evaluated the adequacy of the B0P piping systems by comparing the existing design bases parameters with the core power uprate conditions shown in Table 4.5-1 of the August 30, 1994 submittal.

The comparison indicated t

that for most piping systems, the design temperature and pressure are bounding for the power uprate.

For piping systems, such as the 3rd Point Extraction Steam piping and the 5th Point Heater Drain piping, in which the temperature exceeds the design requirements of the system, the licensee performed the stress analyses for the proposed core power uprated conditions. The evaluation concluded that the 5th Point Heater Drain piping nozzle loads exceed the allowable limit.

In the February 13, 1995, response to the staff's request for additional information, the licensee committed to reinforce the affected nozzle for operation at the uprated power level.

In addition, the licensee reviewed the design bases pipe break analyses to evaluate the effects of the uprate conditions on the pipe break locations, jet thrust, and jet impingement forces that were used in the plant hazard analyses and the design of pipe whip restraints. The review verified that the existing postulated pipe break locations are not affected by the power uprate since the design bases piping analyses will not change due to the power uprate. The current design bases for jet thrust and jet impingement forces due to postulated pipe breaks for these systems are not affected by the uprate since the systems do not experience a pre'ssure increase as a result of the core conclusion that the original design analyses for the pipe break locations, jet power uprate. Based on its review, the staff agrees with the licensee's thrust, jet impingement, and pipe whip restraints are unaffected by the power j

uprate.

Based on the above evaluation, for all the secondary-side systems reviewed,

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the staff concurs with the licensee's conclusion that the power uprate has no j

significant impact on the B0P design bases.

B0P systems are evaluated in section 3.6 of this evaluation.

3.3 CONTAINMENT INTEGRITY ANALYSES The licensee has performed containment integrity analyses at uprated power to ensure that the maximum pressure inside the containment will remain below the containment building design pressure of 45 psig if a design bases loss-of-coolant accident (LOCA) or main steam line break (MSLB) inside containment should occur during plant operation.

The analyses also established the pressure and temperature conditions for environmental qualification and operation of safety-related equipment located inside the containment to ensure that the present conditions remain bounding. The peak pressure is also used as a basis for the containment leak rate test pressure to ensure t. hat dose limits will be met in the event of a release of radioactive material to containment.

13 The licensee stated that the containment functional analyses included the assumption of the most limiting single active failure and the availability or unavailability of offsite power, depending on which resulted in the highest containment temperature and pressure. Also, conservative assumptions for core, coolant, and containment initial conditions were used in the analyses.

Based on its review, the staff concurs with the licensee's conclusions.

3.3.1 LOCA Containment Intearity Analyses The Surry containment design involves normal operation at subatmospheric pressure and use of multiple spray systems that return the containment to subatmospheric conditions within one hour after the design basis accident.

The licensee stated that the existing analyses employ the Stone and Webster LOCTIC model for generation of mass and energy releases and for containment integrity and safeguards pump net positive suction head (NPSH) analyses.

For uprated operation, mass and energy analyses were performed for a postulated double-ended rupture in the hot leg and pump suction piping that results in maximum mass and energy release rates.

The licensee gained margin for the uprating analysis using Westinghouse mass and energy evaluation models with assumptions that account for key Surry plant characteristics.

For the blowdown, reflood and post-reflood phases of the pump suction break, the Westinghouse model documented in WCAP-10325 was used.

For the reflood and post-reflood phase of the hot leg break, mass and energy releases were calculated using the model WCAP-8264.

Section 3.6.1.4 of the Surry Core Uprating Licensing Report states that the WCAP-8264 model was used because of l

its capability to calculate reflood and post-reflood phase transient mass and energy release data.

The focus of the WCAP-10325 evaluation model is on the pressure and temperature response of containment. Since generic studies confirm that there is no reflood peak for a hot leg break, the WCAP-10325 model reflood capability was not pursued by Westinghouse. The WCAP-8264 evaluation model remains a valid analytic tool that has been reviewed and approved by the NRC.

The licensee stated that the uprating analyses used the assumptions to ensure that the mass and energy releases are conservatively calculated, thereby maximizing energy release to containment. The analyses used an allowance of

+4*F in RCS maximum operating temperature, a margin of 3% in RCS volume, an allowance of +2% in core rated power, an allowance of +30 psi in maximum RCS pressure, a margin of +15% in core stored energy, a maximum containment backpressure equal to design value, 0% SG tube plugging level for maximum RC volume and fluid release, maximum heat transfer area and reduced coolant loop resistance for increased break flow.

The uprating analyses used the 1979 American Nuclear Society (ANS) decay heat model and revised method of calculating ' mass and energy releases from a broken primary coolant loop post-LOCA to include steam water mixing. The analyses were performed for minimum and maximum safety injection' cases to conservatively bound potential alignments.

The licensee stated that, for uprating analysis, Westinghouse has used the same methodology and assumptions (except the Surry plant specific data) as approved on the dockets for numerous dry containment plants such as Hillstone 3, Beaver Valley 2, and Indian Point Unit 2 for mass and energy

14 release calculations. The staff finds the use of the Westinghouse computer

-models and assumptions for calculating mass and energy releases to the containment acceptab.le.

The mass and energy release data from Westinghouse's postulated LOCA analyses were used to calculate containment pressure and temperature responses with the use of the Stone and Webster computer code LOCTIC, which was also used in the original analysis.

Key containment-assumptions used for calculating maximum pressure and temperature were, as per TS, a maximum containment initial air pressure of 10.3 psia, a maximum containment initial temperature of 125'F and 100% relative humidity, a maximum service water temperature of 95*F, and a maximum refueling water storage tank water temperature of 45'F.

The Surry containment is maintained at a subatmospheric air partial pressure between 9.0 and 10.3 psia consistent with TS Figure 3.8-1 depending upon the cooldown capability of the Engineered Safeguards equipment.

For the postulated pump suction double ended rupture (PSDER) break, the Surry uprating analyses calculated a containment peak pressure of 44.44 psig.

For the hot leg double ended break (HLDER), the uprating analyses calculated a containment peak pressure of 43.7 psig. These peak pressures remain below the containment design pressure of 45 psig, which is used for containment leak rate testing as per 10 CFR Part 50 Appendix J.

The peak containment temperature during LOCA was calculated to be 275.6*F for the HLDER break and j

274.6*F for the PSDER break.

These temperatures remain bounded by the current peak temperature used for equipment qualification. The staff finds the use of the LOCTIC computer code and above key assumptions acceptable.

3.3.2 MSLB Containment Intearity Analysis The licensee indicated that analysis of the. containment response following a steamline rupture was not included in the original NSSS safety analyses performed for Surry Units 1 and 2.

The containment integrity design assessment was based upon the calculations of effects following a large break LOCA accident. The impact of an MSLB on containment response was evaluated for Surry during the analyses performed for elimination of the boron injection tank in 1983 and it was concluded that:

(1)

Transient temperature and pressure behavior for containment design is established from analysis of a postulated design basis break LOCA.

(2)

LOCA temperature transient results are used for post-accident equipment qualification as allowed by IE Bulletin 79-OlB and its supplements. For a PWR MSLB inside containment " equipment qualified for a LOCA temperature is considered qualified for a MSLB environment in plants with automatic spray systems not subjected to disabling single component failures." The Surry spray systems meet this condition.

(3)

MSLB transient temperature and pressure behavior is established from an analysis of design basis MSLB for Beaver Valley Unit 1, which bounds the expected response for Surry Units 1 and 2.

i l

15 (4)

The transient temperature and pressure for a Surry design basis LBLOCA bounds that obtained in the Beaver Valley Unit 1 MSLB analysis.

The proposed uprating has no effect upon the first two issues listed above, since these are associated with application of regulatory requirements and are independent of plant operating conditions.

For the uprating evaluation, it is only necessary to assess items 3 and 4 above concerning the relative severity of Surry and Beaver Valley analyses.

The Beaver Valley analysis was concluded to be bounding primarily because of a much greater assumed AFW flowrate than is credible for Surry. The licensee indicated that the assumed flowrate to the faulted SG for Beaver Valley is approximately four times the Surry value of 400 gpm.

Since the proposed uprating involves essentially no change in SG operating parameters, the previous conclusion that Surry MSLB containment response is bounded by Beaver Valley remains valid. The staff finds this acceptable.

Validation of item 4 requires demonstrating that the revised Surry LOCA analysis containment response bounds the expected Surry MSLB response. The licensee stated that the mass and energy releases from the revised LOCA analysis were compared with the corresponding MSLB values used in the previous evaluation. The revised Surry LOCA analysis results bound the Surry MSLB results documented in a November 30, 1983 submittal. Therefore, the containment response for power uprate will remain bounded by the postulated LOCA analysis.

Based on the above discussion, the staff finds the licensee analyses for dete mining the containment peak pressure and temperature for design basis LOLA rnd MSLB acceptable as the methodology and assumptions used for calc iating mass and energy release and for calculating pressure and temperature t ansients have been previously used for plants of similar design to meet the requirements of Standard Review Plan (SRP) Section 6.2.1.3 for mass and energy analyses and Section 6.2.1.1.A for dry PWR containment integrity peak pressure analyses.

The proposed change for power uprate will not affect the containment integrity as the calculated peak containment pressure of 44.44 psig remains below the containment design pressure of 45.0 psig.

3.3.3 Deoressurization Analysis The licensee has indicated that the depressurization analysis was performed to show that the containment can be returned to subatmospheric conditions within I hour and remain subatmospheric thereafter. The limiting transient for depressurization results from the PSDER break and the failure of one train of an engineered safety feature (ESF) to actuate.

The other initial containment parameters that were used result in the most conservative depressurization analysis.

The analysis results indicate that containment pressure will become subatmospheric in 2820 seconds and remain subatmospheric thereafter.

Therefore, the proposed uprate will not affect the Surry containment depressization criterion of I hour.

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16 3.3.4 Safeauards Pumos Net Positive Suction Head Analysis The licensee indicated that the mass and energy release rates for the double-ended pipe suction break with minimum safety injection were found to be limiting for sump temperature calculations. The LOCTIC containment evaluation model was used to calculate the sump temperature profile. The minimum initial containment pressure was used. Also, assumptions were made to concentrate as much energy as possible in the containment sump. The results of these assumptions is a minimum steam phase heat inventory and a corresponding minimum heat removal. The licensee indicated that the power uprate does not affect the minimum nat positive suction head (NPSH) requirements for the ECCS system. The minimum NPSH available for inside and outside recirculation spray pumps is 13.0 ft available versus 10.2 ft required for inside pumps and 10.0 ft available versus 9.1 ft required for outside pumps. The minimum NPSH available for low head safety injection pumps is 17.0 ft versus 15.8 ft required.

Based on the above review, the staff finds the power uprate acceptable with regard to containment sump temperature as it does not adversely affect the safety-related equipment.

3.3.5 LOCA Short-Term Subcompartment Analysis The existing containment integrity analyses include structural analyses of the pressurizer cubicle, SG cubicle, and reactor vessel cavity. The licensee indicated that the impact of uprating containment subcompartment pressurization was evaluated and it has been determined that the proposed conditions will cause negligible increase in cubical pressurization and that i

the current short-term LOCA mass and energy releases remain bounding for the uprating conditions. Based on the above, the staff concludes that the uprating is acceptable as the subcompartment pressure loading analysis from 1

high-energy-line ruptures remain bounded by the current UFSAR analysis.

3.3.6 Combustible Post-LOCA Gas Control Analysis l

The licensee indicated that the existing containment post-LOCA hydrogen generation calculation was reviewed to determine the impact of the proposed uprate conditions.

It was concluded that the key parameter values assumed in 3

the hydrogen generation calculation bound the proposed uprated conditions.

In I

particular, a core power of 2597 MWt was assumed, which is 102% of the uprated power. The hydrogen generation calculation also assumes that 2% of the core zirconium reacts to produce hydrogen. This is conservative, since the LOCA-ECCS analyses described in Sections 3.4.5 and 3.4.6 of the Surry Core Uprate i

Licensing Report meet the 1% limit specified in 10 CFR' 50.46. Based on the above, the staff finds the power uprate acceptable as the hydrogen generation remains bounded by the current analysis.

i Based on the above, the staff has concluded that the licensee has evaluated and assessed the various energy release analyses and that all potentially available sources of energy from postulated accidents have been included in the reassessment.

We further conclude that the Surry power uprate will have no adverse effect on containment integrity.

17 3.4 Control Systems and Instrumentation In Section 4.6 of the August 30, 1994 submittal, the licensee presented the results of the assessment it performed to determine the effect of power uprate on the B0P instrumentation and control valves.

In Section 4.7, the licensee discussed its evaluation to validate the existing and proposed setpoint values for actuation of Reactor Protection Systems (RPS) and ESF Systems functions.

Many of the proposed changes to the TS involved revisions to the RPS trip and interlock setpoints. A discussion of each specific change to the TS is presented at the end of this Safety Evaluation.

In its submittal of August 30, 1994, the licensee provided evaluation results of the effect of power uprate on the plant control systems. This evaluation determined that the first point feedwater heater drain level control valves needed to be replaced to enhance plant control.

The licensee replaced these valves in Unit I and has committed to replace them in Unit 2 prior to power uprating. Westinghouse also assessed the plant operating margin and control system capabilities for the licensee based on the RCS operating parameters and identified minor changes to the control system setpoints for the steam dump control, pressurizer level control, and the reference temperature for the rod control system.

The licensee also validated the safety-related instrumentation system setpoints to assure that protective circuits will continue to perform their design functions with the revised setpoints for the control systems. The licensee's submittal of August 30, 1994 did not identify the instrument setpoint methodology applied, and Table 4.7.1-1 did not properly identify the TS changes associated with these instruments.

However, by letter dated April 28, 1995, the licensee provided this information.

The staff reviewed the April 28, 1995 letter and determined that the licensee's setpoint methodology is the same as that previously approved, and the changes to the TS are acceptable.

The proposed setpoint changes are intended to maintain the existing margins between operating conditions and the reactor trip setpoints. These new setpoints also do not significantly increase the likelihood of a false trip or failure to trip upon demand.

Therefore, the existing licensing basis is not affected.

The staff concludes that the licensee's instrument setpoint methodology and the resulting setpoint changes incorporated in the TS for power uprate are consistent with the Surry 1 and 2 licensing basis and are, therefore, acceptable.

3.5 Electrical Systems Evaluatign 3.5.1 Loss of Reactor Coolant Flow Incident A loss of reactor coolant flow incident can result from a mechanical or electrical failure in an RCP or from an interruption in the power supply to these pumps.

If the reactor is at power at the time of the incident, the immediate effect is a rapid increase in coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly.

18 The protection implemented for the loss-of-coolant flow incident, as documented in Section 14.2.9.1 of the Surry Power Stations (SPS) Updated Final Safety Analysis Report (UFSAR), includes the following reactor trip circuits:

(1) low voltage or frequency on the RCP power supply buses (2)

RCP supply circuit breaker opening (3) low reactor coolant flow As a result of core uprating, the protection implemented for the loss-of-coolant flow incident was clarified, as documented in Section 3.5.6 of the Surry Core Uprate Licensing Report transmitted by letter dated August 30, 1994, to include the following reactor trip circuits:

(1) low reactor coolant flow (2)

RCP motor circuit breaker opening (3) low voltage on pump power supply busses (4) low frequency on pump power supply busses (opens RCP supply breakers)

Of these, only the low reactor coolant flow reactor trip is assumed in the analysis.

The low frequency and low voltage signals are not credited for reactor protection, but are assumed to trip the RCPs at their appropriate setpoints.

I In addition, the licensee documented, on page 2.3-7 of the Surry Power Station proposed TS changes (also transmitted by letter dated August 30,1994),that the low-flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more RCPs. The accident analysis conservatively ignores the undervoltage and underfrequency reactor trips and assumes reactor protection is provided by the low-flow reactor trip. The undervoltage and underfrequency reactor trips are retained as back-up protection.

Based on the above inforeation, and a subsequent licensee response to an NRC request for additional information dated February 27, 1995, the staff concluded the following:

~

(1)

The original circuit design for reactor trip on 10w reactor coolant flow meets the IEEE 279 criteria and thus the single failure criterion. The circutt design will continue to meet these requirements following core uprating.

TS requirements for the design will also continue in force.

(2)

For protection of the reactor, circuitry for tripping the RCPs, other devices, and/or the reactor on underfrequency/undervoltage conditions are not credited in the Surry Station accident analysis for the Loss of Reactor Coolant Flow Incident. The original licensing basis for the 6

r

=,

19 Surry Station, thus, did not require this circuitry to meet IEEE 279 criteria. The licensing basis following core uprate will continue not to require this circuitry to meet IEEE 279 criteria.

The staff concludes that the proposed changes to the TS clarify licensing basis requirements that will continue to be applicable after core uprate. The changes do not alter current TSrequirements and are, therefore, considered acceptable.

3.5.2 Electrical Systems Criterion 17 of 10 CFR Part 50, Appendix A, requires that an offsite electrical system be provided which has sufficient capacity and capability to permit safety systems to perform their required functions for all modes of plant operation.

The design basis for the Surry Station requires that the electrical systems be designed to supply electrical power to all essential unit m ipment during normal operation and under accident conditions. The staff a ncluded that this design basis requirement is the same as that stated above h Criterion 17 of 10 CFR Part 50, Appendix A.

The Surry Station electrical system includes circuits that supply power to the safety buses through switchyard and reserve station service transformers. These circuits make up the offsite electrical system at the Surry Station. Criterion 17 of 10 CFR Part 50, Appendix A, and the design basis for the Surry Station require that these circuits (i.e., the offsite system at the Surry Station) have scfficient capacity and capability to permit safety systems to perform their required functions for all modes of plant operation.

For core uprating, the licensee reanalyzed the Surry Station offsite system to assure that system components are sized for the increased worst case load requirement.

The licensee concluded that the worst loadings are less than offsite system component ratings.

Based on the results of the licensee's reanalysis, the staff concludes that the offsite system design will have sufficient capacity and capability following core uprate to permit safety systems to perform their required function for all modes of plant operation. The design therefore meets the requirements of Criterion 17 of 10 CFR Part 50, Appendix A, or the design basis for the Surry Station, as described above, and is acceptable.

3.6 Balance of Plant Systems Evaluation 3.6.1 Auxiliary Feedwater System The licensee stated that existing analyses used for the design of the AFW system were based on a reactor power of 2597 MWt, which is approximately 102%

of the proposed uprated power level.

Therefore, the core uprating has no impact on the AFW system.

Based on its review, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the AFW system.

20 3.6.2 Main Steam System The licensee evaluated the effects on the steam and power conversion system due to plant operations at the proposed uprated power level. The licensee stated that the steam flow for each SG steam header resulting from plant operations at the proposed uprated power level will be 3,753,333 lb/hr, which is less than 1.2% above the design flow of 3,711,100 lb/hr for the main steam trip valve and main steam non-return valve. The small increase in steam flow will produce a less than 1 psi increase in pressure drop across the valves.

These valves remain adequate for the uprated condi}F are essentially unchanged ions.

The present SG main steam outlet design conditions of 785 psia and 516 as a result of plant operations at the proposed uprated power level.

The licensee stated that steam flow for plant operations at the proposed uprated power level plus a 2% margin is 11,485,200 lb/hr. The main steam safety valves have a total relieving capacity of 11,527,362 lb/hr.

Therefore, these main steam safety valves remain adequate for the uprate conditions.

Based on its review, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the main steam system.

l 3.6.3 Extraction Steam Systern The extraction steam system is designed to provide steam at various pressures and temperatures to preheat condensate and feedwater as it flows from the main condensers to the SGs.

Since the extraction steam system does not perform any i

safety-related function, the staff has not reviewed the impact of plant l

operations at the proposed uprated power level on the extraction steam system design and performance.

3.6.4 Auxiliary Steam Systu)

)

i The auxiliary steam system, which is designed to supply low-pressure saturated steam throughout the station for auxiliary services does not perform any safety-related function. Therefore, the staff has not reviewed the impact of j

plant operations at the proposed uprated power level on the auxiliary steam system design and performance.

3.6.5 Condensate and feedwater Systems The licensee had evaluated the condensate and feedwate'r systems for the plant operations at the proposed uprated power level.

The licensee stated that the condensate and feedwater systems satisfy their design bases for plant operations at the proposed uprated power level.

Since these systems do not perform any safety-related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on the design and performance of these systems.

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e 21 l

3.6.6 Feedwater Heaters The licensee had evaluated the feedwater heaters for plant operations at the proposed uprated power level. The licensee stated that the feedwater heaters satisfy their design bases for plant operations at the proposed uprated power level.

Since these feedwater heaters do not perform any safety-related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on their design and performance.

3.6.7 Main Turbine The licensee stated that a safety evaluation for main turbine operations at the proposed uprated power level was performed by Westinghouse.

The evaluation included performing calculations and a detailed review of all critical components with respect to design acceptance criteria to verify the mechanical integrity under the conditions imposed by the power uprating. The evaluation showed that there would be no increase in the probability of turbine overspeed or associated turbine reissile production due to plant operations at the proposed uprated power level.

Therefore, the licensee concluded that the turbines could continue to be operated safely at the proposed uprated power levels.

1 Based on its review of the licensee's rationale and the review of power uprate applications for similar PWR plants, the staff agrees with the licensee's conclusion that plant operations at the proposed uprated power level will have an insignificant or no impact on the main turbines.

3.6.8 Moisture Seoarator and Hiah-Pressure Heater Drain Systems The licensee evaluated the moisture separator and high-pressure heater drain systems for plant operations at the proposed uprated power level. The licensee stated that the moisture separator and high-pressure heater drain systems sacisfy their design bases for plant operations at the proposed uprated power level.

Since these moisture separator and high-pressure heater drain systems do not perform any safety-related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on their design and performance.

3.6.9 Low-Pressure Heater Drain System The licensee evaluated the low-pressure heater drain system for plant operations at the proposed uprated power level. The licensee stated that the low-pressure heater drain system satisfies its design bases for plant operations at the proposed uprated power level.

Since this low-pressure heater drain system does not perform any safety-related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on its design and performance.

3.6.10 Circulatina Water System / Ultimate Heat Sink The circulating water system provides cooling water for the main condensers and the service water systems of both units. The system is designed to take

1 1

~

22' water from the James River (which serves as the ultimate heat sink) on the east end of the site and to discharge to the James River on the west end of the site. The licensee evaluated the effects of plant operations at the proposed uprated power level on the circulating water system and ultimate heat sinkapd_statedthat-theoutletcirculatingwater.temperatureincreasesless than 1 F at uprated conditions and that the increase in rejected heat to the James River is within the current discharge permit limits.

Based on its review of the licensee's rationale, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the circulating water system and ultimate heat sink.

3.6.11 Service Water System The service water system (SWS) is designed to supply cooling water to various non-safety-related components and heat exchangers in the turbine, reactor, and radwaste buildings during normal plant operation, and to supply cooling water to emergency core cooling system components and other essential equipment during a loss of off-site power event and/or a LOCA. The licensee, having performed evaluations, stated that the SWS as designed will supply sufficient water to remove the additional heat loads resulting from plant operations at the proposed uprated power level.

j i

Based on its review, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the SWS.

f 3.6.12 Component Coolina Water System The component cooling water (CCW) system provides cooling water to selected components (e.g. fuel pool cooling system, RHR system, RCP motor coolers, l

chemical and volume control system, etc.) during normal operation and is not required for the operation of engineered safety features following a LOCA or MSLB accident. The system is a closed loop system which serves as an intermediate barrier between the SWS and potentially radioactive systems in order to eliminate the possibility of an uncontrolled release of radioactivity.

The licensee evaluated the effects of plant operations at the proposed uprated power level on the CCW sys}em. The licensee stated that.the only increase in heat load (from 20.12 x 10 Btu /hr to 20.2'x 10' Btu /hr) during normal plant operationatupratedconditignswasfromtheCVCSandthattheCCWsystemhas adequate capacity (50.3 x 10 Btu /hr) to accept this minimal increase in CVCS heat load.

Based on its review, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the CCW.

3.6.13 Bearino Coolina Water System The bearing cooling water system, which is designed to supply cooling water for turbine generator bearings, does not perform any safety-related function.

t

. - -... ~. -...

23 Therefore, the staff has not reviewed the impact of plant operations at the.

proposed uprated power level on the bearing cooling water system design and performance.

3.6.14 Fuel Pool Coolino System The licensee stated that spent fuel pool heat loads were evaluated for plant operations at the uprated power level. The results of the evaluation indicate that the original analyses associated with decay heat rate, time-to-boil, evaporation from boiling, and the associated consequences are still valid due to conservatism used in the original analyses.

Plant operations at the proposed uprated power level do not change the design aspects and operations of the spent fuel pool cooling system.

For this reason and from its experience in reviewing power uprate applications for other plants, the staff concludes that plant operations at the proposed uprated t

power level will have an insignificant or no impact on the spent fuel pool cooling system. However, an issue associated with spent fuel pool cooling adequacy was identified in NRC Information Notice No. 93-83, " Potential Loss of Spent Fuel Pool Cooling Following a loss of Coolant Accident (LOCA)," dated October 7, 1993, and in a 10 CFR Part 21 notification, dated November 27, 1992. The staff will address this issue for the licensee as part of the generic evaluation process.

3.6.15 Plant Heatina. Ventilation. and Air Conditionina (HVAC) Systems The licensee indicated that plant HVAC systems (including fuel building, auxiliary building, and cortrol room HVAC systems) were evaluated for plant operations at the uprated power level.

The results of the evaluation indicate that power uprating will have no impact on these systems.

Plant operations at the proposed uprated power level do not change the design aspects and operations of the HVAC systems.

For this reason and from the experience gained in reviewing power uprate applications for other plants, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the plant HVAC systems.

1 3.7 Eauioment Oualification Inside and Outside Containment The licensee evaluated the effects of plant operations at the proposed power level on qualified equipment, including safety-related electrical equipment and mechanical components.

i With regard to the radiological dose used for equipment qualification (EQ),

the licensee stated that the existing dose used for EQ was calculated based on a reactor power level of 2546 MWt.

Therefore, the existing EQ is still valid for plant operations at the proposed power level.

With regard to the temperatures and pressures used for qualifying equipment inside containment, the licensee stated that the results of existing calculations remain bounding for those temperatures and pressures resulting from plant operations at the proposed power level.

24 With regard to high energy line break analyses which support equipment environmental qualification outside containment, the licensee stated that the existing calculations remain bounding for plant operations at the proposed power level.

l Since the EQ parameters affected by the proposed changes remain bounded by the values determined by the existing analy.;es, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the EQ of electrical equipment and mechanical components and, therefore, is acceptable.

3.8 Analyses for ComDliance With 10CFR50. Aooendix R j

The licensee reviewed the analyses, that were performed for the Appendix R evaluation (10 CFR 50 Appendix R Report, Surry Power Station Units

{

l and 2, Revision 11, November 1993), for potential impact from plant j

operations at the proposed power level.

The licensee stated that the existing j

analyses associated with Appendix R fire protection assumed conditions that bounded those associated with plant operations at the proposed power level and, thus, remained applicable.

Since plant operations at the proposed uprated power level do not change the design aspects and operations of the fire suppression or detection systems, the staff concurs with the licensee's conclusion that plant operations at the proposed uprated power level will have an insignificant or nu impact on the Appendix R evaluation previously performed.

3.9 Radwaste Systems 3.9.1 Gaseous Waste System The gaseous waste system is designed to provide adequate radioactive decay storage for the waste gases and, in addition, provides long-term holdup of these gases. The licensee stated that the Surry fission product source terms used for the gaseous waste system design are based on a reactor power of 2546 MWt; therefore, plant operations at the proposed uprated power level will have an insignificant or no impact on the plant gaseous waste system.

Based on its. review, the staff agrees with the licensee that plant operations at the proposed uprated power level will have an insignificant or no impact on the plant gaseous waste system.

3.9.2 Liouid and Solid Waste System The liquid waste system collects and processes radioactive liquid waste.

The radwaste facility evaporator system processes the normal liquid waste generated by the two units.

The radwaste facility also has an ion-exchange j

system available for use during high liquid waste generation periods or as a backup to the evaporator system.

The spent resin and solid waste systems process solid waste for shipment offsite.

The radwaste facility also has a solidification system capable of processing waste, including evaporator concentrates and spent resins.

The licensee stated that the radwaste facility

25 was recently, installed to increase the plant radwaste processing capability and that the impact of the uprate will be within the current capacity of the radwaste facility.

1 Based on its review, the staff concludes that plant operations at the proposed uprated power level will have an insignificant or no impact on the plant liquid and solid waste system.

3.10 Eadioloaical Conseauences Analysei The staff reviewed the potential increase in design basis accident (DBA) radiological consequences due to the power uprate. The licensee evaluated the impact of the proposed amendment to show that the applicable regulatory acceptance criteria continue to be satisfied for the uprated power conditions.

i t

In conducting this evaluation, the licensee evaluated the effect of the power uprate on the DBA radiological consequences. The original licensing DBA source terms for Surry were considered.

The licensee also evaluated control room habitability under DBA conditions.

3.10.1 Desian Basis Accidents The licensee stated that the original radiological consequence analyses could not be exactly reconstituted. Therefore, the analyses were performed using the methodology described in the UFSAR with the original licensing basis assumption of 2546 MWt (105 percent of current power level).

For the uprate analyses, the core radionuclide inventory was based on a power level of 2605 MWt. The licensee's doses are within the dose reference values stated in 10 CFR Part 100 and the SRP.

The calculated control room operator doses are within the limits to control room operators given in General Design Criterion (GDC) 19.

The events evaluated for uprate were the LOCA, MSLB, steam generator tube rupture (SGTR), locked rotor accident (LRA), the fuel handling accident (FHA),

and the waste gas decay tank (WGDT) rupture.

The whole body and thyroid dose were calculated for the exclusion area boundary (EAB), the low population zone (LPZ), and the control room.

The plant-specific results for power uprate remain well below established regulatory limits.

The analysis was based on 105 percent of the uprated power, using methodologies currently approved by the NRC. After reviewing the information submitted by the licensee, the staff concludes that for the uprated power, the analyzed consequences of DBAs will remain within the limits of 10 CFR Part 100 and GDC 19 and are, therefore, acceptable.

The control room operator doses were estimated using the methodology given in the SRP Section 6.4.

These computed offsite and control room operator doses are well within *.he acceptance criteria given in SRP Section 15.7.4 and GDC 19, respectively.

26 CONCLUSION Based on our review of the licensee's major assumptions, the methodology used in the licensee's dose calculations, and the staff's original Safety Evaluation, the staff finds that the offsite radiological consequences and control room operator doses at the uprated power level of 2546 MWt will continue to remain below 10 CFR Part 100 dose reference values and the GDC 19 dose limit. Therefore, the staff concludes that the licensee's request to uprate the authorized maximum reactor core power level by 4.3 percent to 2546 MWt from its current limit of 2441 MWt is acceptable.

3.11 Chanaes to the Ooeratina Licenses The licensee has proposed the following changes to the operating licenses for Surry, Units 1 and 2, to authorize an increase in licensed power from 2441 MWt to 2546 MWt:

Operating License No. OPR-32 (Unit 1), DPR-37 (Unit 2) - Condition 3.A:

Maximum Power Level is being revised to read "The licensee is authorized to operate the facility at steady state reactor core power levels r.ot in excess of 2546 megawatts thermal."

Operating License No. OPR-32 (Unit 1), DPR-37 (Unit 2) - Condition 3.N:

This condition, which refers to control room dose calculations which are being superseded by analyses contained in the license amendment application, is being deleted.

These changes are consistent with the previously discussed analyses and the NRC staff finds them acceptable.

3.12 Chanaes to the Technical Specifications A number of changes to the TS will be necessary to accommodate the proposed power uprate.

The changes can generally be categorized as fcilows:

core rated power changes in the definitions and in the operating license and other changes associated with the revised plant nominal operating conditions changes that reflect the safety limits and limit'ing safety system settings associated with revised accident analyses (NSSS events, containment analyses, and radiological consequences analyses) changes in limiting conditions for operation associated with revised e

analyses or evaluations The specific changes to the TS are described below.

TS page 1.0.1 (TS 1.0. A) - Definition of RATED POWER - Revise to state 2546 MWt

a 27 TS page 2.1-3 (Basis for Figure 2, 1 - 1) - Delete sentence "The three loop j

operation...to 100% of design flow." This refers to densification effects no longer included in the analysis basis.

TS Figure 2.1 -1 (Reactor Core Thermal and Hydraulic Safety Limits) - Replace with figure of limits which reflect operation at uprated conditions (4.1-1).

TS page 2.1-4 (Basis for Figure 2.1-1) - Revise to reflect relationship between deterministic and statistical analysis basis incorporated into the figure.

TS page 2.1-5 (Basis for Reactor Control & Protection System) - Revise stated nominal RCS temperature to 573.0*F.

Delete footnote *, which allows Unit 2 Cycle 12 RCS nominal operating pressure to be reduced to 2135 psig.

The licensee determined that the temperature (T') value used in the equations for overtemperature AT and overpower AT needed to be changed from 574.4*F to 573.0*F.

The instrument setpoint was then determined based on the Westinghouse methodology documented in WCAP-8746-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," dated September 1986.

This methodology has been approved by the staff previously and, therefore, the proposed new TS setpoint changes are acceptable.

TS page 2.2-2 (Basis for TS 2.2) - Delete footnotes

• and **, which refer to reduced power-operated relief valves and high pressure reactor trip settings for Unit 2 Cycle 12.

TS page 2.3-2 (TS 2.3.A.2(b)) - High Pressurizer Pressure Reactor Trip -

Delete footnote *, which refers to a reduced trip setting for Unit 2 Cycle 12.

TS page 2.3-2 (TS 2.3.A.2(d)) - Overtemperature AT - Revise T' to equal 573.0*F, the proposed nominal RCS average temperature.

TS page 2.3-7 (Basis for Low Flow Reactor Trip) - This paragraph is being revised to emphasize that the low flow trip is the primary trip and that undervoltage and underfrequency trips are considered back-up protection. This revision reflects the assumptions of the revised complete loss-of-flow analysis.

TS page TS 3.1-1 (TS 3.1.A.l.a) - LC0 for Number of RCPs - Revise to read "A reactor shall not 6s brought critical with less than three pumps, in non-isolated loops, in operation." This change reflects the revised analysis of uncontrolled rod withdrawal from a subcritical condition.

TS page 3.1.3 (TS 3.1.A.3.b) - Pressurizer Safety Valve Lift Settings - Delete the footnote *, which refers to expanded pressurizer safety valve lift setting tolerance for the remainder of Cycle 10 and 11 for both units.

TS page 3.1-16 to 3.1-17a (Basis for Coolant Activity Limits) - The description on these pages has been rewritten, based upon the revised SG tube rupture radiological consequences analysis.

4 28 TS page 3.3-7 (Basis for Accumulator valves) - Delete the footnote *, which.

refers to a reduced nominal operating pressure for Unit 2 Cycle 12.

)

TS page 3.6-2 (TS 3.6.E) - SG Secondary 1-131 Activity - Delete the sentence I

"The iodine-131 activity in the secondary side of any steam generator, in an unisolated reactor coolant loop, shall not exceed 9 curies," and revise the next sentence to begin "The specific activity..." This change reflects the secondary activity assumed in the revised MSLB radiological analysis, which is a specific activity of 0.10 C1/cc. A separate limit on total activity is redundant and is not required by the revised analysis.

TS page 3.6-4 (Basis for Emergency Condensate Storage Tank Capacity) - Revise-3 basis statement to read "The specified minimum water volume in the 110,000-gallon protected condensate storage tank is sufficient for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of residual heat removal..." Add a sentence that reads "It is also sufficient to maintain one unit at hot shutdown for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, followed by a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> cooldown from 547'F to 350'F (i.e. RHR operating conditions)." This reflects the cooldown capability for operation at uprated conditions.

TS page 3.6-4 (Basis for Main Steam Safety Valve Capacity) - Revise the statement of main steam safety valve flow capacity to read "... total combined capacity of 3,842,454 pounds per hour at their individual relieving pressure; the total combined capacity of all fifteen main steam code safety valves is 11,527,362 pounds per hour." Revise the second sentence to read "The nominal power rating steam flow is 11,260,000 pounds per hour." These changes reflect a revised calculation of valve relief capacity and the increased nominal steam flow associated with uprating.

TS page 3.6-5 (Basis for SG Secondary Activity) - Delete the text starting with "The limit on steam generator..." through the sentence which ends with

... the specific iodine-131 limit would be.089 uCi/cc." Replace with the following:

The limit on steam generator secondary side Iodine-131 activity is based on limiting the inhalation dose at the site boundary following a postulated steam line break accident to a small fraction of the 10 CFR 100 limits.

The accident analysis, which is performed based on the guidance of NUREG-0800 Section 15.1-5, assumes the release of the entire contents of the faulted steam generator to the atmosphere.

These changes replace the previous description, which provided a basis for comparison between the total and the specific activity limits.

Since the total activity limit (TS 3.6.E)_ is being deleted and the revised analysis does not require a specific total activity, this discussion is not relevant.

TS page 3.7-26 (Table 3.7-4) - Recirculation Mode Transfer (RMT) - Revise the RWST Level-Low setting limits to be as follows:

211.25% and s15.75%. These revised settings reflect the values assumed in the LOCA containment analyses.

The setpoint value has been validated by the analyses as providing adequate

29 margin for containment depressurization while ensuring that the low head safety injection pcmps will have adequate net positive suction head for operation in sump recirculation mode.

The licensee dettrmined that, to meet acceptable analysis margins, the refueling water storage tank level setpoint for automatic RMT needed to be reduced from the current nominal value of 18.93% span to 13.5% span. The staff previously approved the setpoint methodology for establishing automatic RMT during an inspection at Survy in January 1993. This approval is documented in NRC Inspection Raport 93-01, dated February 23, 1993, and therefore, the proposed new TS setpoint is acceptable.

TS page 3.8-4 (Basis for Figure 3.8-1) - Revise the description of the figure characteristics and numerical ranges to be consistent with the replacement figure.

TS Figure 3.8-1 (Allowable Air Partial Pressure) - This figure, which presents operating limitations for containment air partial pressure, containment bulk average temperature, and service water temperature, has been revised in

. conjunction with the containment integrity analysis. The revised figure allows operation over a range of air partial pressure from 9.0 psia to 10.3 psia and over a temperature range of 75'F to 125'F.

TS page 3.10-7 (Basis for Activity Assumed in Fuel Handling Accident) - Revise the description to state that this accident has been analyzed based on the methodology in Regulatory Guide 1.25, assuming that 100% of the gap activity from the highest powered assembly is released after a 100-hour decay period following operation at 2605 MWt. This reflects the revised radiological dose consequences analysis.

Delete the footnote *, which compares the fuel rod gap activity of 15x15 and 17x17 demonstration assemblies. This information is not relevant to any fuel assemblies currently in Surry cores. Any potential future use of demonstration assemblies will be addressed on a case-specific basis.

1 TS page 3.12-12 (TS 3.12.F.1) - DNB Parameters - Revise the temperature limit to state " Reactor Coolant System T 5577.0*F."

This reflects the proposed y

nominal operating temperature, plus uncertainties, that have been accommodated i

in the revised thermal-hydraulic analyses (4.1-2). Delete footnote *, which refers to reduced nominal pressurizer pressure for Unit 2 Cycle 12.

TS page 4.1-10a (Table 4.1-28) - Minimum Frequencies for Sampling Tests -

Delete the words "9 Curie" in Notes (4) and (8).

This' reflects the deletion of the 9-curie total activity limit, leaving only a general reference to v

Specification 3.6.E, which contains the limit.

i TS page 4.4-3 (Basis for Containment Air Partial Pressure Limits) - Revise statement of containment pressure range to read "The containment is maintained i

at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon..." This figure presents the range assumed in the containment integrity analysis.

l

30 TS page 5.2-3 (IS 5.2.C.1) - Containment Systems - Revise the stated Recirculation Spray Subsystems flow to be "at least 3,000 gpm of water from the containment sump." This reflects the flowrate assumed in the revised containment analysis.

We find the above TS changes acceptable because they are appropriate to the i

uprated power and are supported by acceptable analyses.

l There are also several administrative type changes to the TS on some of the above pages.

These are described below.

They have been reviewed as part of the staff's evaluation and determined to the acceptable.

Deletion of references to two-loop operation in TS 2.3.A.2.(d) (TS p.

2.3-2) and 3.1.A.1.e (TS p. 3.1-3), as well as in various Basis discussions (TS pp. 2.1-3, 2.1-4, 2.1-5, and 2.3-7).

TS 3.3.A.11 prohibits two-loop power operation.

Capitalization of defined words and systems (e.g., Reactor Coolant System, OPERABLE, etc.) (Insert A on TS pp. 2.3-7, 2.1-4, 3.1-1, 3.1-3, 3.1-17, 3.6-2, 3.3-7 and 4.1-10A).

Terminology revisions for consistency (e.g., hot shutdown conditions versus HOT SHUTOOWN, FSAR versus UFSAR, etc.) (TS pp. 2.2-2, 3.6-2, 3.6-4, 4.1-10a, and 4.4-3).

Correction of typographical errors (TS pp. 2.3-2, 3.6-4, 3.10-7, and

=

4.1-10a).

i In addition to the above TS changes, the licensee changed the assumed value used in the uprating analysis for the following parameters:

High-High SG 1evel----Turbine Trip l

High-High SG 1evel----feedwater Isolation High Pressurizer level----Reactor Trip For each of these parameters, the licensee determined that there is sufficient difference between the revised analysis value and existing limiting setting to provide adequate margin to accommodate channel uncertainties. Therefore, no change in the existing limiting setting is required.

The proposed setpoint changes are intended to maintain,the existing margins between operating conditions and the reactor trip setpoints. These new setpoints also do not significantly increase the likelihood of a false trip or failure to trip upon demand.

Therefore, the existing licensing basis is not l

affected.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendments.

The State official

{

had no comment.

31

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on June 21, 1995 (60 FP 32356).

In this finding, the Comission determined that issuance of these amendments would not have a significant effect on the quality of the human environment.

6.0 CONCLUSION

The Comission has concluded, based on the considerations discussed above, I

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 f

activit,tes will be conducted in compliance with the Comission's regulations, and (3) '.'he issuance of these amendments will not be inimical to the comon defenta and security or to the health and safety of the public.

i Principal Contributors:

R. Clark, B. Buckley, S. Brewer, H. Garg, R. Goel, J. Knox, J. Minns, F. Orr, D. Shum, C. Wu Date:

August 3, 1995 l

i 7590-01

' UNITED STATES NUCLEAR REGULATORY COMISSIDH VIRGINIA ELECTRIC AND. POWER COMPANY DOCKET NOS. 50-280 AND 50-281 NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U.S. Nuclear Regulatory Commission (Commission) has issued Amendment Nos. 203 and 203 to Facility Operating License Nos. DPR-32 and DPR-37 issued to Virginia Electric and Power Company, which revised the License and the Technical Specifications for operation of the Surry Power Station, Unit Nos. I and 2 located in Surry County, Virginia. The amendments are effective as of the date of issuance.

The amendments modified the Licenses and the Technical Specifications to increase the authorized core power level for Surry, Units 1 and 2, from 2441 MWt to 2546 MWt.

The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations. The Comission has made appropriate findings as required by the Act and the Comission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment and Opportunity for Hearing in connection with this action was published in the FEDERAL REGISTER on December 16, 1994 (59 FR 65085). No request for a hearing or petition for leave to intervene was filed following this notice.

The Comission has prepared an Environmental Assessment related to the action and has determined not to prepare an environmental impact statement.

f

~

Based upon the environmental assessment, the Commission has concluded that the issuance of the amendment will not have a significant effect on the quality of the human environment (60 FR 32356).

For further details with respect to the action see (1) the application for amendment dated August 30, 1994, and supplemented February 6, February 13, February 27, March 23, March 28, April 13, April 20, April 28, May 5, and June 8, 1995, (2) Amendment Nos. 203 and 203 to License Nos. DPR-32 and DPR-37, (3) the Commission's related Safety Evaluation, and (4) the Commission's Environmental Assessment. All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street NW., Washington, DC, and at the local public document room located at the Swen Library, College of William and Mary, Williamsburg, Virginia 23185.

Dated at Rockville, Maryland, this 3rd day of August 1995.

FOR THE NUCLEAR REGULATORY COMMISSION

\\%ft (, QaucSy n

Bart C. Buckley, Senior Project Manager Project Directorate 11-1 Division of Reactor Projects Office of Nuclear Reactor Regulation

.