ML061310178

Technical Specification Change TS-06-09, Revision of Ultimate Heat Sink Temperature.

ML061310178
Person / Time
Site:	Watts Bar
Issue date:	05/08/2006
From:	Skaggs M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-WBN-TS-06-09
Download: ML061310178 (59)

Text

Tennessee Valley Authority, Post Office 2000, Spring City, Tennessee 37381-2000 Mike Skaggs Site Vice President, Watts Bar Nuclear Plant TVA-WBN-TS-06-09 MAY 0 8 2006 10 CFR 50.90 U. S. Nuclear Regulatory Commission Mail Stop: OFWN P1-35 ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:

In the Matter of ) Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) - UNIT 1 - TECHNICAL SPECIFICATION (TS) CHANGE TS-06-09, "REVISION OF ULTIMATE HEAT SINK (UHS) TEMPERATURE" Pursuant to 10 CFR 50.90, TVA is submitting a request for an amendment to WBN's License NPF-90 to change the Technical Specifications for Unit 1.

The proposed TS change (TS-06-09) will revise the limiting condition for operation for TS Section 3.7.9, "Ultimate Heat Sink." The maximum essential raw cooling water (ERCW) temperature limit associated with Surveillance Requirement (SR) 3.7.9.1 would increase from 85 degrees Fahrenheit (F) to 88 degrees F. This proposed change is based on evaluations of the ERCW system and the UHS functions and maximum temperatures that will satisfy the associated safety functions. Corresponding TS Bases changes for the temperature increase for UHS and a minor change in the maximum calculated containment pressure resulting from the increased UHS temperature are attached for information.

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U.S. Nuclear Regulatory Commission Page 2 MAY 0 8 2006 This change does not impact proposed TS change TVA-WBN-TS-05-09 currently under review by the NRC staff to increase the ice condenser ice weight due to replacement of the steam generators. The Loss of Coolant (LOCA) pressure analysis performed for TS change TVA-WBN-TS-05-09 used a UHS temperature of 88 degrees F.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Public Health. to this letter provides the description and evaluation of the proposed change. This includes TVA's determination that the proposed change does not involve a significant hazards consideration, and is exempt from environmental review. An Attachment to Enclosure 1 provides TVA's detailed justification for the subject proposal including a description of the revised WBN design basis.

Enclosures 2 and 3 contain copies of the applicable Unit 1 TS and TS Bases pages, respectively, marked-up to show the proposed changes.

It is TVA's judgment that Tennessee River conditions could approach temperature limits for the UHS this summer.

Therefore, TVA requests approval of this TS change by early July 2006 to mitigate a potential threat to continued plant operation. TVA requests that the implementation of the revised TS be within 45 days of approval.

TVA will continue to monitor river temperature predictions to further validate the necessity of approval of this TS change by the requested approval date. We will promptly notify the NRC of any significant developments.

U.S. Nuclear Regulatory Commission Page 3 MAY o 8 2006 To support the NRC staff in the review of this TS change request, TVA will make its staff available to meet with the NRC staff at their convenience to discuss the basis for increasing the UHS temperature limit.

A list of regulatory commitments is provided in Enclosure 4.

If you have any questions about this change, please telephone P. L. Pace at (423) 365-1824.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 8th day of May, 2006.

Sincerely, Mike Skaggs

Enclosures:

1. TVA Description and Evaluation of the Proposed Change Attachment - "Ultimate Heat Sink - 88 0 F Maximum Operating Temperature Evaluation," Watts Bar Nuclear Plant.

2. Proposed Technical Specifications Changes (mark-up)

3. Proposed Technical Specifications Bases Changes (mark-up)

4. List of Commitments cc: See Page 4

U.S. Nuclear Regulatory Commission Page 4 MAY 0 8 2006 Enclosures cc (Enclosures):

NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. Lawrence E. Nanney, Director Division of Radiological Health 3 d Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 DOCKET NO. 390 PROPOSED LICENSE AMENDMENT REQUEST WBN-TS-06-09 REVISION OF ULTIMATE HEAT SINK (UHS) TEMPERATURE DESCRIPTION AND EVALUATION OF PROPOSED CHANGE

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-90 for Watts Bar Nuclear Plant Unit 1. The proposed TS change will revise the limiting condition for operation (LCO) for TS Section 3.7.9, "Ultimate Heat Sink." The change will increase the maximum allowed essential raw cooling water (ERCW) temperature from 85 degrees Fahrenheit (F) to 88 degrees F for Surveillance Requirement (SR) 3.7.9.1. This proposed change is based on evaluations of the ERCW system and the UHS functions and maximum temperatures that will satisfy the associated safety functions.

Corresponding TS Bases changes are included for the temperature increase for UHS and a minor change in the maximum calculated containment pressure resulting from the increased UHS temperature.

2.0 PROPOSED CHANGE

The proposed change will revise WBN TS Section 3.7.9, "Ultimate Heat Sink," specifically, Surveillance Requirement (SR) 3.7.9.1. The limit in SR 3.7.9.1 will be increased from 85 degrees F to 88 degrees F. Corresponding TS Bases changes are also proposed. The TS and TS Bases changes affect the following sections and are illustrated by marked-up pages provided in Enclosures 2 and 3:

Technical Specification SR 3.7.9.1 is revised to increase the ultimate heat sink temperature limit from 85 degrees F to 88 degrees F.

Technical Specification Bases Corresponding TS Bases changes to sections B3.7.7, "Component Cooling System," B3.7.8, "Essential Raw Cooling Water," and B3.7.9, "Ultimate Heat Sink,"f are included for the temperature increase to 88 degrees F for the UHS. TS Bases sections B3.6.4, "Containment Pressure," and B3.6.6, "Containment Spray System," are also revised to reflect a minor change to the peak containment pressure resulting El-l

b from the containment pressure reanalysis using the increased ERCW temperature of 88 degrees F. Refer to the markups provided in Enclosure 3.

3.0 BACKGROUND

TVA's review of Tennessee River conditions indicate the UHS temperature this summer could challenge the Technical Specification limit of 85 degrees F. If this limit is exceeded, it could result in unit shutdown to comply with Technical Specification LCO 3.7.9.

Measurement of the UHS temperature in accordance with SR 3.7.9.1 takes into account the accuracy of the instrumentation used to measure the temperature. The SR 3.7.9.1 UHS temperature measurement is performed by averaging the temperature readings in the ERCW supply headers (lA, 1B, 2A, and 2B). To account for instrument inaccuracy in the measurement of the UHS temperature, if all four measurement points are available then the UHS limit is verified to be less than or equal to 84.8 degrees F. If less than four ERCW supply header temperature points are available then the UHS limit is verified to be less than or equal to 84.7 degrees F. If measurement and test equipment (M&TE) is used for the four ERCW temperature points, then the TS limit is adjusted by the error of the M&TE.

Therefore, TVA proposes that the maximum ERCW temperature limit in SR 3.7.9.1 be increased from 85 degrees F to 88 degrees F. The proposed change is based on evaluations of the ERCW system, UHS functions, and maximum temperatures utilizing existing margins in the ERCW design basis.

4.0 TECHNICAL ANALYSIS

The methodology and results of TVA's evaluation of the proposed UHS temperature increase of 3 degrees F above the existing design basis value of 85 degrees F to 88 degrees F are discussed in detail in the attached evaluation, "Ultimate Heat Sink - 88 degrees F Maximum Temperature Evaluation, Watts Bar Nuclear Plant." (Reference 1) A description of the ERCW and UHS Systems is provided in the WBN Updated Final Safety Analysis Report (UFSAR) Sections 9.2.1 and 9.2.5, respectively.

The effects of the proposed temperature increase have been examined in detail on equipment, components, systems, and safety analyses and have been found to be acceptable. An extensive review was performed of existing calculations, procedures, design criteria, system descriptions, and the UFSAR. The evaluation included several heat transfer calculations, a UHS drawdown calculation resulting from postulated breach of the Chickamauga Dam, a Residual Heat Removal (RHR) design basis cooldown analysis, and a design basis containment pressure response calculation for a postulated loss-of-coolant-accident (LOCA) occurring with UHS/ERCW intake water temperature of 88 degrees F. As discussed in the attached evaluation, the methods of evaluation varied and included sensitivity evaluations, bounding analyses, and reanalysis (e.g., for the containment pressure E1-2

analysis) to demonstrate that the cooling water requirements for the most demanding accident, shutdown, and cooldown conditions have been met.

Margin exists in many areas within the prior ERCW design basis analyses. ERCW flow rates, cooled medium (water and air) flow rates, heat exchanger fouling, heat exchanger tube plugging, equipment and room heat loads, heat load losses to ambient, assumed lake levels, and margin inherent in structural Codes (piping and supports), all offer various amounts of margin in the analyses. In most cases, only ERCW flow rate margin was utilized, and in some cases, heat load margins were used to evaluate acceptable ERCW system performance at 88 degrees F. Some specific systems, structures and/or components required a more in-depth evaluation to determine acceptability of minimum required performance. More limiting input assumptions were utilized in some evaluations such that additional margins were created, for example, use of a higher ERCW flow rate for a component supported by preoperational test data. The attached evaluation discusses when such techniques or different methods were used for the evaluations. With the exception of these ERCW component evaluations and the RHR cooldown analysis discussed below, no other changes in assumptions or methodologies were credited for the evaluation of the revised UHS temperature change.

An evaluation which addressed ERCW system flow margins to system components was conducted in April 2004. This evaluation and associated flow margins have been recently revalidated by assessing emergent issues or changes which have impacted the system or components served since April 2004. In some cases, the revalidation examined changes since the preoperational test program. These included ERCW pump performance and throttle valve position changes.

The revalidation results are summarized in section 6 of the attached evaluation. The evaluation concludes that the margins remain acceptable.

The impact of the proposed change on shutdown and removal of residual heat, including time requirements for achieving safe shutdown conditions, were assessed in Section 3.16 (RHR Impact) of the attached evaluation. The analysis concluded that acceptable system performance during normal plant cooldowns at 88 degrees F would be obtained. For the case of single train cooldown, in which TS compliance with the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to cold shutdown is dependent on single train availability of RHR, the cooldown time period could be achieved by ensuring spent fuel pool cooling is isolated for up to five hours and assurance that the remaining reactor coolant pump is secured no later than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after shutdown. Securing the last pump at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> is consistent with and bounded by operational practices that would occur during a loss of offsite power event, where all power is provided by the emergency diesel generators (EDGs). In this event, the reactor coolant pumps (RCPs) are not loaded on the emergency shut down electric boards.

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Approval of the proposed changes will result in a change of the WBN UHS design basis maximum temperature limit to 88 degrees F.

As discussed in Section 3.18 of the attached evaluation, TVA performed a re-analysis of the postulated effects of a non-flood failure of Chickamauga Dam (downstream of WBN). Although the analysis of UHS and ERCW performance at 88 degrees F did not quantitatively credit this analysis, the results indicate a higher minimum river water pool elevation (approximately seven feet greater than assumed for the original ERCW flow balance), thus providing additional flow capacity to ERCW components.

The UHS temperature increase to 88 degrees F resulted in a minor increase in the maximum calculated containment pressure due to a design basis LOCA. This change from 10.64 psig to 10.90 psig remains below the American Society of Mechanical Engineers (ASME)

Code design internal pressure of 13.5 psig. The containment accident reanalysis is discussed in the Section 3.4 of the attached evaluation. The containment functional design is discussed in UFSAR Section 6.2.1.

TVA's evaluation concluded that there is not a significant increase in the risk or consequences of normal operation, shutdown, or accident mitigation or danger to the public, equipment, or site personnel because adequate margins exist in the critical systems, structures, and/or components. The evaluation concluded that a Technical Specification change to allow continued plant operation up to 88 degrees F UHS temperature is acceptable.

Changes to Operating Requirements The attached evaluation identifies several changes to operating requirements in order take advantage of the higher temperature allowance for the UHS. These items will be implemented through the design change and TS change implementation processes and are discussed below.

1) The EDG's jacket water heat exchangers were shown to be marginal in their performance capabilities if standard raw water design fouling conditions were assumed. The evaluation concluded that expected actual fouling rates of the EDG heat exchangers would provide acceptable EDG cooling provided tube cleaning was performed annually during the spring time-frame.

This would assure relatively clean EDG heat exchangers during the late summuer time period of maximum UHS temperatures.

Therefore, TVA will revise the UFSAR to require EDG jacket water heat exchangers to be cleaned once each spring. These heat exchanger cleanings are currently in progress for this calendar year.

2) As discussed above, procedure changes will be necessary to address RHR shutdown cooling assumptions. TVA will revise the UFSAR to address single-train RHR cooldown restrictions for El-4

ERCW temperatures above 85 degrees F which consist of a minimum five hour isolation time for spent fuel pool cooling and a requirement to secure the remaining reactor coolant pump within 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after shutdown.

3) Revise System Descriptions, Design Criteria, and other key design documents and the UFSAR to implement operation above 852F and the new UHS design basis of 88 degrees F.

5.0 REGULATORY SAFETY ANALYSIS The proposed TS change will revise the limiting condition for operation for the UHS. The ERCW temperature limit will be increased from 85 degrees F to 88 degrees F. These proposed changes are based on evaluations of the ERCW system and the UHS functions and maximum temperatures and minimum river elevations that will satisfy the associated safety functions. The proposed changes will minimize the likelihood of a required unit shutdown as a result of slightly higher river temperatures in the summer.

5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to increase the UHS maximum temperature will not adversely alter the function, design, or operating practices for plant systems or components.

The UHS is utilized to remove heat loads from plant systems during normal and accident conditions. This function is not expected or postulated to result in the generation of any accident and continues to adequately satisfy the associated safety functions with the proposed changes. Therefore, the probability of an accident presently evaluated in the safety analyses will not be increased. The heat loads, that the UHS is designed to accommodate, have been evaluated with the higher temperature limit. The result of these evaluations is that there is existing margin associated with the systems that utilize the UHS for normal and accident conditions.

These margins are sufficient to accommodate the postulated normal and accident heat loads with the proposed changes to the UHS. Since the safety functions of the UHS are maintained, the systems that ensure acceptable offsite E1-5

dose consequences will continue to operate as designed.

The change in the maximum calculated containment pressure associated with the design basis loss-of-coolant-accident (LOCA) remains below the American Society of Mechanical Engineers (ASME) Code design internal pressure.

Therefore, the consequence of any accident will be the same as those previously analyzed.

Since the UHS safety function will continue to meet accident mitigation requirements and limit dose consequences to acceptable levels, TVA has concluded that the proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The UHS function provides accident mitigation capabilities and serves as a heat sink for normal and upset plant conditions; the UHS is not an initiator of any accident.

By allowing the proposed change in the UHS temperature requirements, only the parameters for UHS operation are changed while the safety functions of the UHS and systems that transfer the heat sink capability continue to be maintained. The proposed change does not impact the response of the systems and components assumed in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change has been evaluated for systems that are needed to support accident mitigation functions as well as normal operational evolutions. Operational margins were found to exist in the systems that utilize the UHS capabilities such that these proposed changes will not result in the loss of any safety function necessary for normal or accident conditions. The ERCW system has excess flow capacity that will accommodate the increased flows necessary for the proposed temperature increase.

While operating margins have been reduced by the proposed changes, safety margins have been maintained as assumed in the accident analyses for postulated events. The proposed change results in an increase in the maximum calculated containment peak pressure. However, the change in the El-6

maximum calculated containment peak pressure associated with the design basis LOCA is a small percentage of the margin between the current maximum calculated containment peak pressure and the ASME Code design internal pressure.

This aspect of the proposed change does not involve a significant reduction in a margin of safety.

Additionally, the proposed changes do not require any further modification of component setpoints or operating provisions that are necessary to maintain margins of safety established by the WBN design (the shutdown board room chillers were physically modified to operate properly at the 88 degree F UHS temperature). Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The Commission's regulatory requirements related to the content of the TS are contained in Title 10, Code of Federal Regulations (10 CFR), Section 50.36. The Ultimate Heat Sink (UHS) temperature satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) and is therefore included in TS 3.7.9.

General Design Criterion (GDC) 44, "Cooling Water," of Appendix A, "General Design Criteria," to 10 CFR Part 50, provides design considerations for the UHS. Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants,"

provides an acceptable approach for satisfying this criterion. The discussion below regarding Regulatory Guide 1.27 compliance demonstrates the ability to meet the recommendations of the regulatory guide and therefore satisfies the requirements of GDC 44.

As discussed in WBN UFSAR Section 9.2.5, UHS, Regulatory Guide 1.27 provides an acceptable approach for the design of the UHS. This guidance provides four criteria for an acceptable UHS function. These criteria include recommendations for sufficient cooling capability, integrity during postulated events, function availability and redundancy, and control by the TSs. TVA has evaluated the proposed changes and their impact on the UHS design based on the criteria in Regulatory Guide 1.27 and has determined that these recommendations continue to be met.

The cooling ability of the UHS, with the proposed increase in temperature, has been evaluated and verified to satisfy E1-7

the recommendations for heat removal considerations. The integrity and availability recommendations have not been affected by the proposed changes as the features are not being altered physically. The TS provisions are proposed to be changed but continue to meet the recommendation to provide actions in the event the function of the UHS cannot be satisfied. Therefore, operation of the WBN unit with the proposed TS change will not result in a deviation from the recommendations of Regulatory Guide 1.27.

Appendix A to 10 CFR 50, GDC 16, "Containment design," GDC 38, "Containment heat removal," and GDC 50, "Containment design basis," continue to be met with the proposed change.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 50.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

Refer to Section 7 of the attached evaluation for a complete listing of References. In addition, the following references are included.

1. "Ultimate Heat Sink - 88 degrees F Maximum Temperature Evaluation, Watts Bar Nuclear Plant."

2. Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants," Revision 1, March 1974.

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ATTACHMENT TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 Ultimate Heat Sink - 88 degrees F Maximum Operating Temperature Evaluation E1-9

WATrS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation BRIEF Watts Bar Nuclear Plant (WBN) utilizes the Tennessee River, Chickamauga Reservoir, as the ultimate heat sink (UHS). This is the source of the water used by the Essential Raw Cooling Water (ERCW), High Pressure Fire Protection (HPFP), and the Raw Cooling Water (RCW) systems. Allowable plant operation as a function of UHS temperature is regulated by Technical Specification 3.7.9 with limiting conditions for operation (LCO) tied to average ERCW supply header temperatures. The current design basis and TS limit for the UHS is 85QF.

Multipurpose river operation, coupled with hotter than normal summers, and below normal river flows have challenged and continue to challenge the UHS temperature limit prompting resolution of this critical operating condition. This evaluation reviews an increase of the design basis UHS temperature limit to 880F.

Effects of the proposed UHS temperature increase of 30F to 880F have been examined in detail on equipment, components, systems, and safety analysis and have been determined not to create any unsafe conditions. Some specific systems, structures and/or components required a more in-depth evaluation to determine acceptability of minimum required performance. In addition, certain key safety analyses, i.e.,

containment pressurization and plant cooldown analyses, were re-analyzed by Westinghouse at the higher UHS temperature (as an input assumption) of 88'F to assure adequate margins exists in the primary containment design. The analysis methodology itself did not change. The Loss of Downstream Dam (LODD) scenario was reanalyzed utilizing a minimum Watts Bar Dam discharge flow of 14,000 cubic feet per second. This is consistent with the flow that is currently guaranteed at the Sequoyah Intake.

This evaluation concludes that there is not a significant increase in the risk or consequences of operation, shutdown, or accident mitigation or danger to the public, equipment, or site personnel because adequate margins exist in the critical systems, structures and/or components. This conclusion is further corroborated by detailed calculation reviews, specific component calculation evaluations, review of material conditions, nuclear industry wide research, and standard Technical Specification developments which are described in the evaluation.

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WATrS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88 0F Maximum Operating Temperature Evaluation TABLE OF CONTENTS Cover Page Brief Table of Contents Revision Log 1.0 Definitions & Terms 2.0 Discussion 2.1 Background Information 2.2 Design configuration & Description 2.3 UHS Temperature 2.4 Temperature & river Level Discussion for Accident Mitigation 2.5 System Description - WBN ERCW System 2.6 Safety Functions 2.6.1 Phase A Containment Isolation 2.6.2 Phase B Containment Isolation 2.6.3 LOOP With Loss of One Shutdown Power Train 2.6.4 Flood Mode 2.6.5 Other Modes of Operation 2.7 Normal Functions 2.8 One Unit Only Operation 2.9 Loss of Upstream and / or Downstream Dam 2.10 Instrumentation and Control 2.11 Component Cooling System 3.0 Methodology and Results 3.1 Safety Analysis 3.2 Plant Operation 3.3 ERCW Design Basis Events 3.3.1 Appendix R Compliance 3.3.2 Loss of Downstream Dam 3.3.3 Loss of Offsite Power 3.3.4 Station Blackout 3.3.5 Loss of Diesel Power 3.3.6 A Critical Crack in Piping 3.3.7 An ERCW Pipe Break in the Turbine Building 3.4 Westinghouse Containment Reanalysis 3.5 Other Accidents or Events 3.6 Environmental Qualification Impacts 3.7 Fire Protection 3.8 Flood Mode 3.9 Spent Fuel Pool 3.10 Auxiliary Building Secondary containment Enclosure and Auxiliary Building Gas Treatment System 3.11 Raw Cooling Water 3.12 Environmental Analysis 3.13 Plant / System Condition Review 3.14 Technical Specification 3.15 Piping and Supports 3.16 Residual Heat Removal System Impact 3.17 Tritium Production 3.18 River Data 3.19 Deleted 3.20 Other Operating Experience 3.21 ERCW Flow Margins Page 2

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88'F Maximum Operating Temperature Evaluation TABLE OF CONTENTS 3.22 Tube Plugging 3.23 NPSH 3.24 Regulatory Impacts / Review 3.25 MCR Habitability 3.26 Radiological Impacts 3.27 Motor Operated Valves 3.28 ASME Section Xi 3.29 Instrumentation and Control 3.30 Auxiliary Feedwater System 3.31 Regulatory Program Review 3.32 Design Basis Control 4.0 Summary of Results 5.0 Recommendations 6.0 Revalidation Effort 7.0 References Page 3

WAITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 1.0 DEFINITIONS & TERMS AC - Air Conditioner, Air Conditioning Equipment AFW - Auxiliary Feedwater CCS - Component Cooling System CCP - Centrifugal Charging Pump CFS - Cubic feet per second CRDM - Control Rod Drive Mechanism CSD - Cold Shutdown, mode as described in the Technical Specification CSS - Containment Spray System CST - Condensate Storage Tank DBA - Design Basis Accident DBE - Design Basis Event EDG,DG, D/G - Emergency Diesel Generator ECCS - Emergency Core Cooling System EL, Elevation - Elevation above Mean Sea Level in feet ESF - Engineered Safety Features EQ - Equipment Qualification ERCW - Essential Raw Cooling Water

'F, degrees F - degrees Fahrenheit UFSAR - Updated Final Safety Analysis Report HPFP - High Pressure Fire Protection HSB - Hot Standby, mode as described in the Technical Specification Hx, HtX, HTX - heat exchanger INPO - Institute of Nuclear Power Operations LBLOCA - Large Break Loss of Coolant Accident LCO - Limiting Condition for Operation LER - Licensee Event Report LODD - Loss of Downstream Dam LOOP - Loss of Offsite Power MCR - Main Control Room MIC - Microbial Induced Corrosion MSLB - Main Steam Line Break NPDES - National Pollutant Discharge Elimination System NPSH - Net Positive Suction Head NPSHa - available NPSH PER - Problem Evaluation Report RCS - Reactor Coolant System RCP - Reactor Coolant Pump RCW - Raw Cooling Water RHR - Residual Heat Removal RSO&E - River System Operations and Environment Group RWST - Refueling Water Storage Tank SFP, SFPC&CS - Spent Fuel Pit Cooling and Cleanup System SER - Safety Evaluation Report, by NRC SSE - Safe Shutdown Earthquake STS - Standard Technical Specification SQN - Sequoyah Nuclear Plant TS, Tech Spec - Technical Specification TVA - Tennessee Valley Authority, Licensee UHS - Ultimate Heat Sink, Tennessee River UT - Ultrasonic Examination WBH - Watts Bar Hydroelectric Power Plant WBN - Watts Bar Nuclear Power Plant WO - Work Order Page 4

WAITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 2.0 DISCUSSION

2.1 BACKGROUND

INFORMATION The UHS temperature at WBN has previously approached the Technical Specification limit of 850F. If this limit had been exceeded it could have resulted in unit shutdown to comply with Technical Specification LCO 3.7.9. This evaluation compiles information from many sources into a general technical evaluation. TVA has performed an extensive review of ERCW, and other affected systems which would be impacted by an increase in river water temperature to 88 0F. The reviews were performed on the Design Criteria, System Descriptions, UFSAR, Technical Specifications (TS), Standard Technical Specifications (STS), Technical Specifications Bases, STS Bases, and river temperature related documentation. WBN operating data and procedures were reviewed, historical records were reviewed, and new calculations of the reservoir surface elevation drawdown transient after a postulated failure of Chickamauga Dam were created. Normal operation and accident conditions of the ERCW system and the systems it supplies with cooling water were considered with emphasis on the accident mitigation and safe shutdown cases. TVA, NRC, and industry guidelines, as well as specific Westinghouse analyses were utilized in the evaluation for impact of the ERCW operating temperature limit increase.

The conclusion of the review is that there is sufficient justification to increase the UHS upper temperature allowable limit from 850 F to 88 0F. Operational procedure guidelines will be enhanced, as required, in order to implement this limit. Modifications to the shutdown board room chiller compressors have been made to support the implementation of this change. Once the requested Technical Specification change is approved by the Nuclear Regulatory Commission, the UFSAR and ERCW design documentation will be revised to support the implementation of the Technical Specification change.

The WBN UHS consists of the water source and associated structures used to remove waste heat from the plant during all plant normal, shutdown, and accident conditions. The overriding safety function of the UHS is dissipation of residual and decay heat for a unit shutdown and for shutdown after a postulated accident.

At WBN, the UHS is comprised of a single water source, the Tennessee River. This includes the complex of TVA-controlled dams upstream of the plant intake, TVA's Chickamauga Dam (the nearest downstream dam),

and the plant intake channel. During normal operation, cooling water flows from the Chickamauga Reservoir through the plant intake channel to the intake pumping station. The intake channel is located on the inside of a bend in the river about two miles downstream of Watts Bar Dam. The intake channel extends about 800 feet from the edge of the reservoir through the flood plain along a line approximately perpendicular to the river flow. The bottom of the channel is at a sufficient depth to ensure direct flow from the main river channel to the pumping station during all low water levels, including the level resulting from a loss of downstream dam (LODD). A floating pontoon type structure is provided across the channel to serve as a barrier and discourage direct approach to the pumping station from the reservoir. The barrier is designed to make it virtually impossible to sink; however, if it were to sink, it could not block the channel to the extent of preventing the required flow from reaching the station. Water from the UHS is pumped to the plant by the ERCW and raw cooling water (RCW) pumps, and in certain events, the high pressure fire protection (HPFP) pumps. All of these pumps are located at the Seismic Category I intake pumping station. The intake pumping station design assures protection of the safety related ERCW pumps and HPFP pumps from the design basis flood. The ERCW pumps and HPFP pumps are capable of functioning under any plant design basis condition including a safe shutdown earthquake (SSE) plus LODD and a LOCA.

River flows vary according to a host of needs. The river system is operated to reduce flood damage, maintain a navigable waterway, supply power, enhance conditions for aquatic life, and supply water for drinking, recreation, and industry - including cooling water for TVA's thermal plants. River Systems Operation and Environment (RSO&E) maintains and follows a formal process for river control which, among other things, monitors and moderates the UHS temperature for the nuclear plants. It has been observed that in the area around WBN that a summer temperature excursion above the existing design basis temperature of 850 F is possible for a limited duration.

River temperatures are recorded at various site locations for compliance with environmental requirements under the WBN site specific National Pollutant Discharge Elimination System (NPDES) Permit.

Page 5

WARTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation Environmental Permit limits are typically reported in WBN daily management meetings with regard to condenser performance, electrical power output and associated economics of operation.

Operation of the closed loop Condenser Circulating Water (CCW) system provides transfer of waste plant heat from power operations to the environment through the cooling tower. The normal CCW waste heat rejection path at WBN is augmented by the Supplemental Condenser Circulating Water System (SCCW).

Makeup water from the Tennessee River is also added as required to replace losses through evaporation and blowdown by using the discharge from the ERCW and RCW cooling water systems. Blowdown discharges from the cooling tower basin are made under the conditions of the WBN site specific NPDES Permit.

2.2 DESIGN CONFIGURATION & DESCRIPTION WBN isa two unit (Unit 1 is operational, Unit 2 iscurrently deferred) Westinghouse pressurized water reactor (PWR) site that utilizes the Tennessee River, Chickamauga Reservoir, as the UHS. Chickamauga Dam is the downstream dam that provides primary control of Chickamauga Reservoir level on this portion of the Tennessee River. Watts Bar Dam is the upstream dam that provides primary flow into the Chickamauga Reservoir from the Watts Bar Reservoir. The Hiwassee River which is located downstream of WBN also supplies flow into the Chickamauga Reservoir. Various TVA divisions work together to operate the Tennessee River system and generate electric power for the region. The UHS and river system are discussed in the UFSAR chapter 2.

The intake channel extends from the intake pumping station into the Chickamauga Reservoir to the original riverbed and is dredged down to elevation 660 to provide free access to the river during low river level conditions. Both the normally exposed and submerged portions of the channel are dredged to sufficient width, riprapped on the sides, and seismically qualified to eliminate the possibility of channel blockage due to an earth or mudslide. The channel is monitored periodically and dredged as required to maintain free access to the river. Therefore, adequate water will be available to the ERCW pumps at all times and for all events including the LODD. Since the intake channel is seismically qualified, the occurrence of the SSE could significantly affect the UHS only by causing failure of the non-seismic downstream dam and/or upstream dams. For the resulting low and/or high reservoir event, water would still be available to the intake channel at all times. A seismically induced disturbance of the rock surfaces could only block a small percentage of the intake channel due to its high conservative width.

A tornado cannot disrupt the ERCW water supply to the intake station.

Protection of the intake channel and station against blockage or impact by river traffic is provided by its location. For all conditions of river navigation (up to water level 698 which corresponds to the 40 year flood level in Watts Bar Dam tail waters at which lock operation ceases), the grade elevation of the river flood plain through which the intake channel passes is such that even when the flood plain is submerged, sufficient depth will not exist for passage of any major river vessel at the intake channel or the intake structure location. In addition, due to the close proximity of the upstream dam, the possibility of a barge being accidentally released upstream and reaching the plant site would be extremely remote. However, if such an incident does occur, the barge would be carried away from and beyond the intake channel and station by the higher velocity water passing the plant on the outside of the river bend on the opposite side of the reservoir.

For reservoir levels which would provide sufficient water depth for a barge to approach the intake station, it is not considered credible that serious damage would be incurred as lock operation at the Watts Bar Dam would cease at elevation 698, thereby limiting any traffic on the Tennessee River that had not been stopped due to the flood. The location of the Intake Pumping Station in relatively stagnant, shallow water approximately 800 feet from the main river channel minimizes the potential of traffic on the river impacting the Intake Pumping Station.

TVA regulation of the Tennessee River is such that drought conditions will not jeopardize the UHS's capability; this is historically confirmed by the data in UFSAR Section 2.4.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 2.3 UHS TEMPERATURE The highest temperature measured at the tailrace of the Watts Bar Hydro Plant between 1959 and 2005 was approximately as shown on the following Figure 1:

FIGURE 1 Maximum Annual Water Temperature @WBH Tailrace, 1959 - 2005 85 ----.-------...

83. . . . . . . . .

81. .. . . .. .. . . . . .. .. . .. . . . . ..... .. .. . . .

0.. . . .a.

1. ... . . . .....

. 79 .-'-.8a. . . ti . .. y --. S...........

757- 4-''o'!;,....

. a .a a a a a a

.,5..

1955 1960 1965 1970 1975 1980 1985 1990 1995 2000 2005 201 Year Measured Highest temperature measured during the 1959 to 2005 period was approximately 83.51F in 2002. More recent temperature data was obtained from RSO&E for the WBN operating period of 1996-2005, and is shown in the following Figure 2. The temperatures shown are general river temperatures used for environmental compliance. The chart however shows the yearly water temperature pattern and span.

Page 7

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88°F Maximum Operating Temperature Evaluation FIGURE 2 RSO&E Monitoring Station Intake Water Temperature for WBN 90

- Maximum 1996-2005,

-- - - Minimum 1996-2005

-Average 1996-2005 ,,,

80- -Cn sL -

U-70 -

, - - - ,,-t-! <

0 I I II I I I z I I I .

• I I I I I *' 1e I I I I I I I h k 0

0. 60 - ---- ------r- --- r ---- r-- -- -I- -- -- T - - - :.r- -

E 0!

,. , ,, ,I , , , I I-U 50 -

> ' I r t I I I I ' 'I R~1 __________ ___ ___ _L ___ __ __I -- I- -- I-- -I- --

,, Iw I I 40 - I I I I I I I I I II I I 1 I _ I _I I f _

Ou J an F,. M.par r M Ju Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec This chart does not plot the temperatures recorded for the ERCW headers at WBN for UHS Technical Specification compliance. Temperatures measured in the ERCW supply headers are obtained by more accurate instrumentation than the tailrace temperatures. A comparison of the data from January 1,2006 through April 9, 2006 showed that the ERCW supply header temperatures were slightly lower (-0.5 to 1.00F lower) than the tailrace temperature.

FIGURE 3

WATYS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation There is a 60 hour6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (to a close approximation) drawdown following breach of the Chickamauga Dam to the minimum elevation. This was provided for use in evaluating the effect of a LODD. The ERCW/UHS design has always assumed that the LOCA-recirc/LOOP/Loss of Diesel Train condition is the most limiting design basis condition with a LODD. The ERCW initially provides cooling water to required safety related equipment such as the CCS heat exchangers, room coolers and chillers and subsequently the containment spray heat exchangers.

2.4 TEMPERATURE & RIVER LEVEL DISCUSSION FOR ACCIDENT MITIGATION The first part of the limiting accident Large Break LOCA (LBLOCA) is the injection phase which uses stored water (Refueling Water Storage Tank, Condensate Storage Tank) and the Ice Condenser for the initial accident mitigation. This injection phase lasts up to one hour. The minimum river level throughout the accident is assumed to be 665.9 ft (assuming loss of Chickamauga Dam) and the ERCW intake temperature isassumed to be at its maximum. After the Refueling Water Storage Tank has been depleted (- 10 minutes for 2 train operation or -1 hour for single train operation) "swap over" occurs. At this time, the reactor building sump is recirculated through the Containment Spray System (CSS) heat exchangers which are cooled by the ERCW system. ERCW is required for continued cooling for the long term via the Component Cooling System (CCS) Heat Exchangers (which cool the Residual Heat Removal (RHR) Heat Exchangers) and CSS Heat Exchangers thereby maintaining the containment building pressure below design limits. The ERCW flow rate to the CSS and CCS Heat Exchangers is dependent on the water surface elevation at the intake pumping station (higher river elevation gives higher flow) and pump performance capabilities.

Ensuring acceptable flow is provided to the equipment served by the ERCW was accomplished by flow balancing the ERCW system during preoperational testing such that the ERCW flow rates assumed in the accident analyses are exceeded in practice. The ERCW system was flow balanced based on the assumed loss of Chickamauga Dam condition with the ERCW pumps' discharge throttled to simulate both operation at river EL 665.9 and minimum expected pump head/flow performance.

New recession curves at the WBN river location with initial Chickamauga headwater elevations of 670 feet up to 682.5 feet were evaluated. In considering a LBLOCA with LODD as the limiting accident, the initial (short term) mitigation phase of the accident (prior to switchover to recirculation of the containment sump) is implemented by the ECGS and the containment Ice Condenser ice bed. Within approximately 10 minutes (2 train operation) or one hour (single train operation), the reactor building sump is recirculated through the Containment Spray System (CSS) heat exchangers which are cooled by the ERCW system for long term accident mitigation. The recession curves show that within about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, a LODD drawdown results in the WBN intake channel dropping only about 2 feet. Since the near term mitigation after ice bed meltout involves the CSS and C0S heat exchangers, additional river head (elevation) above the level simulated in the system flow balance performed in PTI-067-02 isadvantageous. An ERCW performance margin of at least 7 feet elevation head exists above the minimum design pool elevation of 665.9 feet assumed in the WBN ERCW flow balance for both short term and long term accident mitigation.

2.5 SYSTEM DESCRIPTION - WBN ERCW SYSTEM The ERCW System provides cooling water to various equipment in both safety-related and nonsafety-related portions of the plant during all modes of normal and Design Basis Event (DBE) conditions. The ERCW System is common to Units 1 and 2. The ERCW system is isolated from those Unit 2 components that are not required for Unit 1 operation and safe shutdown. During normal operating conditions, the ERCW provides cooling water from the Tennessee River to normally operating safety-related and nonsafety-related equipment and normally discharges to the cooling tower basin. During accident conditions, the ERCW System continues to provide cooling water to safety-related equipment (and some non-safety-related equipment) and can discharge either to the cooling tower basin or to the holding pond, or out of the ERCW Overflow Structure if the non-seismic discharge headers become blocked. It also provides water to other safety-related systems and non-safety-related systems when their normal water supply is unavailable.

During floods above plant grade, additional safety-related and non-safety-related equipment normally cooled by the CCS is connected to the ERCW System since the CCS pumps are made inoperable due to flooding.

A simplified flow diagram is shown on the following page:

Page 9

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation FIGURE 3 SIMPLIFIED DIAGRAM - PRIMARY ERCW USERS I Return To Cooling Tower 88 OF Basin or If Yard Holding Pond 1

The ERCW System consists of traveling screens, pumps, and strainers, which are contained in the Intake Pumping Station structure, hydraulic gradient structures in the discharge piping, a discharge overflow structure, valves and piping arranged in two separate and independent trains (A and B), each of which has two supply headers (1A and 2A, and 1B and 2B). Electric power is provided from the four shutdown power trains (1A and 2A, and 1B and 2B). ERCW System components and heat loads are arranged so that during a DBE, with a concurrent LOOP, the system can tolerate either a single active failure of a system component in one train or the loss of a shutdown power train without compromising the ERCW System's safety function or exceeding its overall design temperature of 130OF.

The ERCW System provides a continuous and uninterrupted flow of cooling water to safety and nonsafety-related equipment during normal and DBE modes of plant operation. The ERCW system has been designed to perform its essential functions without offsite power or dependence on nonsafety-related systems.

The ERCW system can meet the minimum flow rates required using only two pumps on either Train A or Train B. WBN has 8 ERCW pumps; 4 pumps on each train (A and B). Two ERCW pumps per train are normally aligned to the emergency shutdown boards. This results in 4 pumps being aligned to emergency electrical power. ERCW flow requirements vary based on plant operational needs, specific plant operational Mode, and/or the presence of an accident condition. Specific flow requirements have been determined by analysis and system testing has confirmed the ability to meet the required flows. Additional pumps are usually available when off-site power is available, although only four pumps can be operated when power is supplied by the onsite diesel generator electric power system (one pump per shutdown board).

This evaluation covers Unit 1 only operation since Unit 2 is currently deferred. Some Unit 2 components do support Unit 1 operation and receive cooling water. Therefore they are included in the Unit 1 operational boundaries.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 2.6 SAFETY FUNCTIONS The ERCW system is required to mitigate the consequences of plant design basis events (DBE's) described in the UFSAR. The original design basis condition of ERCW system performance was based on the following events occurring simultaneously: 1) LOOP, 2) LODD, 3) loss of a shutdown power train, concurrent with one Unit in LOCA and the other unit in Hot Standby. (Note that Unit 2 is currently in the deferred status.) The ERCW system performs a primary safety function by providing cooling and makeup for essential safety-related plant equipment and components in response to adverse plant operating conditions which impose safety-related performance requirements on the systems being served. The safety functions of the ERCW system consist of the following basic modes of operation:

2.6.1 Phase A Containment Isolation This mode responds to minor accidents during which desirable non-safety-related heat loads located in containment are not isolated from the ERCW System. Although the cooling function of the equipment in the Reactor Building which is cooled by the ERCW System is non-essential for accident mitigation, continued operation of the cooling function of the equipment significantly improves the ability to cope with a small steamline break, small LOCA, or steam generator tube rupture.

2.6.2 Phase B Containment Isolation This mode responds to major accidents during which equipment inside containment is isolated from the ERCW System by the containment isolation valves. In addition, ERCW flow through the Containment Spray System Heat Exchangers is established by opening valves that are normally closed and adjusting ERCW flow to the Component Cooling Heat Exchangers, as necessary, to compensate for the flow diverted to the Containment Spray System Heat Exchangers.

2.6.3 LOOP With Loss of One Shutdown Power Train In this mode, the plant's essential cooling requirements are met by reliance on the ERCW train aligned with the operating Shutdown Power Train. Each ERCW train serves safety-related equipment that is either redundant to the other train or that can be realigned to the functioning train by valve manipulation. Each train is powered by independent emergency power sources. In either case, the ERCW system retains full capability to respond to postulated accident events.

2.6.4 Flood Mode During this mode, the equipment required to maintain the plant in a safe shutdown condition is cooled as required. Some of these loads are normally cooled by the CCS. Because the Component Cooling pumps will be flooded and inoperable during floods above plant grade, these loads will be connected to the ERCW System with spool pieces which inter-tie the EROW system to the CCS.

2.6.5 Other Modes of ODeration The ERCW System also supplies emergency flow to the following equipment when the normal water supply is unavailable:

CCS Surge Tanks (upon loss of demineralized makeup water).

Auxiliary Feedwater Pumps (upon loss of Condensate Storage Tank inventory).

2.7 NORMAL FUNCTIONS The ERCW System provides cooling to various safety and nonsafety-related equipment in various parts of the plant. It is a common system to Units 1 and 2, and operates during all normal and DBE modes of plant operation. The ERCW pumps are sized such that the operation of two pumps on each plant train will supply all cooling water requirements for the two unit plant during all modes of operation. Also, two pumps on one Page 11

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 881F Maximum Operating Temperature Evaluation plant train shall be sufficient to supply all cooling water required for the two-unit plant for unit cooldown, refueling, or post accident operations.

2.8 ONE UNIT ONLY OPERATION Heat exchangers in Unit 2 not required for Unit 1 operation or that do not require maintenance of flow for lay-up, are isolated from the ERCW system administratively. Other equipment that serves either unit issecured to serve Unit 1 if required to support unit operation. Full service capability and safety function for Unit 1 operation is retained.

2.9 LOSS OF UPSTREAM AND/OR DOWNSTREAM DAM The Intake Pumping Station has been designed to retain functional capability for floods up to and including the Design Basis Flood, and a Loss of Downstream Dam (LODD). The ERCW system has full capability down to a water surface elevation of 665.9 feet which corresponds to the surface level with a LODD. The resulting minimum surface water surface elevation following a LODD is primarily influenced by dam break assumptions (size and location of break) and WBH releases following the break. Initial lake levels have minimum impact on long term lake elevations at WBN.

2.10 INSTRUMENTATION AND CONTROL Plant personnel can monitor ERCW temperatures using instrument readings that are displayed in the MCR.

This is required to ensure compliance with the Technical Specification requirement (3.7.9) on UHS. ERCW temperatures are input and displayed in the ICS computer system which provides alarms and screen output associated with high temperatures.

2.11 COMPONENT COOLING SYSTEM The original design of the CCS was based on a maximum ultimate heat sink temperature of 850F. River water temperatures exceeding 850F would affect the time required for plant cooldown. During normal operating conditions, the maximum temperature of the component cooling water exiting the component cooling heat exchangers is approximately 950F when the ERCW supply temperature is 850F, during the initiation of RHR cooldown at the fourth hour of unit shutdown, the temperature could approach 110F. Since the CCS is required for post-accident removal of heat from the reactor, the CCS is designed such that no single active or passive failure will interrupt cooling water to both Train A and Train B Engineered Safety Feature (ESF) trains. One ESF train is capable of providing sufficient heat removal capability for maintaining safe reactor shutdown. The CCS pumps and required motor-operated valves are automatically transferred to auxiliary onsite power upon LOOP.

Page 12

WAITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88 0 F Maximum Operating Temperature Evaluation 3.0 METHODOLOGY AND RESULTS The evaluation of the acceptability of 880F UHS temperature was based primarily on available plant margins in the following three areas:

1) The Loss of Downstream Dam (LODD) analysis and water release capability indicates additional capacity exists for the flow established in preoperational testing for the loss of downstream dam simulations due to a higher river elevation following a loss of downstream dam than that simulated during the preoperational test.

2) Establishment of existence of margin in containment design. There are margins to the allowed maximum peak pressure in the containment design based on containment reanalysis using 880 F UHS/ERCW temperature.

3) Establishment that there are sufficient margins in the ERCW system flow rates established during the preoperational testing program to each affected component.

The following subsections provide a brief discussion regarding specific areas of review in determining the acceptability of an 880F UHS design temperature that may exist for limited periods of time during plant operation. Detailed and specific results from the review and evaluation are also provided in the individual sections below.

3.1 SAFETY ANALYSIS There are several major design and operating areas to be considered for power operation and also for accidents with the UHS at 880F. These include:

Normal Plant Operation ERCW Design Basis Events Accident Analysis ECCS and attendant equipment performance UHS Thermal Transport - ERCW system Spent Fuel storage impacts Environmental impacts Operating Experience review Tritium Production/Power Uprating The detailed summaries are listed below and the results documented throughout Section 3.0.

3.2 PLANT OPERATION Plant operation and routine start ups and shut downs were considered with river water temperature at 880F.

Calculations which might be affected by an increase in ERCW temperature were identified and reviewed for impact by the proposed increase in temperature. Tube plugging assumptions or criteria were not altered for any safety related heat exchangers. Most safety related equipment needed before, during, or after an accident was determined to be acceptably cooled by 880F ERCW based on available flow margins above flows required in any accident scenario for a temperature of 850F. Safety related coolers not initially shown acceptable by their flow margins were otherwise found acceptable by a more detailed investigation, details of which are presented in the equipment specific section of this evaluation.

Steam cycle efficiency and MWe output are increasingly affected as ambient air and UHS temperatures increase. Any environmental thermal compliance limit that is challenged can be dealt with successfully by reducing plant power and/or manipulating water releases from Watts Bar Dam. Continued compliance with procedures ECM-3.0 "NATIONAL POLLUTANT DISCHARGE ELIMINATION SYSTEM (NPDES)

PROGRAM", EITP-100 "ENVIRONMENTAL COMPLIANCE", and SPP-5.5, "ENVIRONMENTAL CONTROL" is required.

Page 13

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 3.3 ERCW DESIGN BASIS EVENTS 3.3.1 Appendix R compliance strategy for Unit 1 operation required realigning the ERCW supply headers to the CCS Heat Exchanger A by opening valves 1-FCV-67-223-A (CCS Heat Exchanger C Header 1B Supply Isolation Valve), 2-FCV-67-223-A (CCS Heat Exchanger Supply ERCW Header 2A/1 B Crosstie Valve), closing valve 1-FCV-67-458-A (CCS Heat Exchanger A Header 1B ERCW Supply Valve) and opening their associated breakers in the Motor Control Center. This was done to minimize operator action for an Appendix R scenario by aligning the ERCW Train A to CCS Heat Exchanger A for the scenario in which only ERCW pump A-A or B-A is available for service. This Appendix R scenario assumes a LODD, LOOP, and the loss of the 1B-B, 2A-A, and 28-B power trains thereby resulting in the lowest available flows to components cooled by the ERCW System of any Design Basis Event postulated to occur. Subsequent to the Appendix R event, the Unit will be placed in hot standby. Flow margin exists but in order to ensure adequate longer term cooling, nonessential loads can be isolated, i.e., CCS Heat Exchanger B (if not providing cooling to SFPCS Heat Exchanger B), and DG 2A-A (since it is not operating during this scenario). The flow to CCS Heat Exchanger A can be reduced initially to Hot Standby flow requirements. When the other ERCW pump aligned with EDG 1A-A is loaded on to DG 1A-A, there will be enough ERCW flow available to shut down the unit. This Appendix R scenario was flow balanced/flow tested during preoperational test PTI-067-02 at WBN. Measured test flows to each component and acceptance criteria are listed in Section 3.21, Tables 1A, 1B, and 1C.

3.3.2 Loss of downstream dam (LODD) does not initiate any fault or accident but is assumed to occur concurrently with other design basis accidents. As a result of the lower river elevation, delivered ERCW pressures and flows decrease due to pump suction side static head losses.

3.3.3 Loss of off-site power (LOOP) is considered concurrently with design basis accidents and creates a limiting safeguards condition to which the ERCW is required to perform.

3.3.4 Station Black-out (SBO) can occur during a LOOP but is not coupled with other accident scenarios.

The SBO event does not challenge the ERCW beyond a limiting safeguards condition.

3.3.5 Loss of a Diesel Power Train would remove either the 'A" train or 'B' train ERCW pumps. Thus, only two ERCW pumps on a single train would be available for cooling. This is the limiting safeguards condition. All cooling requirements are met under the minimum safeguards condition with a single train of ERCW cooling water due to equipment redundancy.

3.3.6 A Critical Crack in Category 1 piping is not limited to any specific component under any condition and does not typically present a significant loss of cooling capability to any single component since the ERCW flow rates are relatively large in comparison to the flow from the critical crack.

3.3.7 An ERCW Pipe Break in the Turbine Building results in automatic isolation of non seismic piping based on decreasing pressure and increasing flow in the piping needing isolation.

The design basis events discussed above have been considered as applicable as the basis for all analyses and evaluations performed in support of the proposed change in maximum UHS temperature.

3.4 WESTINGHOUSE CONTAINMENT ACCIDENT REANALYSIS UFSAR Chapter 6, Containment Analysis (containment integrity analysis, (WCAP-15699 Revision 1) describes containment temperature and pressure response following a design basis accident (LOCA). The Westinghouse analysis utilizes the NRC approved LOTIC-1 computer code to concurrently determine containment pressure and temperature response following a design basis LOCA event. This analysis utilizes the UHS temperature as an input for both the CSS Heat Exchanger cooling water and the CCS Heat Exchanger cooling water. The CCS heat exchangers provide cooling water to the RHR heat exchangers and other essential safeguards feature equipment. The containment integrity analysis is based on a double-ended pump suction guillotine break of the reactor coolant system piping as the large break loss of coolant accident (LBLOCA) assuming the minimum ice condenser weight of 2.0294 million pounds of ice and minimal ECCS safeguards for the increased UHS temperature. (The evaluation was for the steam Page 14

WAITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation generators currently installed at WBN. A subsequent evaluation was performed for the replacement steam generators which used the increased UHS temperature of 880F.) The bulk of the heat energy in containment is eventually transferred outside of containment by the UHS via the CSS heat exchanger(s) to the ERCW, and from the RHR heat exchanger(s) via the CCS to the ERCW.

The containment analysis was not significantly affected as a result of increasing the UHS input parameter for the temperature to 880F. Specifically, the new peak pressure of 10.9 psig at 6,664.62 seconds and an ice melt time of 3,624.50 seconds represent a small increase of 0.2599 psi and a small decrease of 1.75 seconds compared to previous peak pressure and ice melt values, respectively. The containment pressure resulting from the increase in the UHS temperature to 880F is below the containment design pressure of 13.5 psig. The 1.75 second reduction in ice melt is not significant with regard to spray system switchover to recirculation.

The containment sub-compartment pressure analysis, the peak containment temperature analysis and the long term containment cooling analysis are also discussed in UFSAR Chapter 6. The containment sub-compartment analysis is for the immediate (initial) response to the double-ended break and it does not utilize the UHS as a heat removal source since the ERCW system is not providing containment cooling at this time.

The peak containment temperature results from a Main Steam Line Break (MSLB) and this occurs very early in the transient during the initial blowdown from the faulted steam generator piping. For a MSLB, the containment pressure/temperature response is not governed by the UHS/ERCW temperature at the time when swap-over from ECCS injection to the containment sump is initiated. ERCW is utilized later on in the cooldown in the containment spray heat exchangers and in the lower compartment coolers (if the cooling coils are operable) for a MSLB to maintain the containment temperature within the EQ limits. The heat release at this time from a MSLB is much less than the initial release when it occurs at full power operation and it continues to decrease as the RCS is depressurized below 370 psig and placed on RHR cooldown. The long term containment cooling capability is based on RHR cooldown input parameters, such as CCS temperature and RCS temperature.

The containment sub-compartment pressure analysis is not affected by the increase in UHS temperature, since it does not utilize the UHS as a heat removal source. Likewise, the peak containment temperature analysis is unaffected by the UHS increase since the peak containment temperature, which results from a main steam line break (MSLB), occurs very early in the transient during the initial blowdown from the faulted steam generator piping before ERCW is utilized for containment cooling after the accident. ERCW is utilized later on in the cooldown by the containment spray heat exchangers and in the lower compartment coolers (if coils are operable) for a MSLB to maintain the containment temperature within the environmental qualification (EQ) limits. Only the LCC's forced air flow iscredited in the accident analysis, not the ERCW supplied cooling water, since the cooling coils are not safety-related.

The results of the Westinghouse analyses show that the containment integrity is maintained and is within the established safety limits for both the LBLOCA and the MSLB. The resultant pressure increases were very small. This is due to the WBN ice condenser design and as such, the ERCW cooling to the containment spray heat exchangers does not come into effect until ice melt-out, RWST depletion, and swap over to the containment sump for recirculation has occurred.

The revised analysis performed by Westinghouse in support of the proposed change to a higher UHS temperature did not involve a change in analysis methodology. The only revised input associated with the analysis was that of UHS temperature being 880F instead of the previous 850F. The changes associated with the revised analysis, including revised peak containment pressure and time to ice bed melt out will be reflected in the design basis and the UFSAR.

Page 15

WATrS BAR NUCLEAR PLANT -UNIT I Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 3.5 OTHER ACCIDENTS OR EVENTS The following accidents are indirectly affected by an increase in UHS temperature to 880 F since the UHS may be required for the shutdown of the unit after the accident or event.

Major or minor secondary system ruptures Complete loss of forced RCS flow or single reactor coolant pump locked rotor Rod cluster withdrawal at full power Rod cluster control assembly ejection Fuel handling accident Waste gas decay tank rupture Inadvertent loading of a fuel assembly into improper location Steam generator tube rupture consequences 3.6 ENVIRONMENTAL QUALIFICATION IMPACTS Regulatory Guide 1.27 requires that the UHS and ERCW be available for a minimum of 30 days following the event with procedural guidelines for beyond 30 days. TVA has chosen a 100 day post-accident timeframe for equipment qualification purposes, so the UHS and ERCW system is also required to be available for that same duration to ensure that EQ limits are met. It is not necessary to assume that the UHS will actually remain at 88 0F for the entire 100 days or beyond, since actual operating experience indicates that any temperature excursion would be for a limited duration. Extensive environmental heat input is required to obtain and maintain 880 F ERCW along with low uncontrolled river flows (usually less than 3,500 cubic feet per second [CFS]). The design duration was considered for both 30 days for accident mitigation and 100 days for long term EQ impact. The decay heat loads at 30 days and 100 days are a small fraction of the initial value respectively.

Even though the net post-accident impact of this temperature change on qualified life of electrical and mechanical equipment would not be significant, a comparison was performed. Outside the reactor containment building, it was found that bounding temperatures were used in evaluating qualification which would not be impacted by this ERCW temperature change. Inside the containment building, it was found that the 88 degree post-accident temperature profile curves for the upper compartment, active sump, and dead ended compartments had slight exceedances using 88 degrees. However, these curves were also bounded by the curves generated to reflect replacement of the steam generators in RFO7. The impacts of the higher replacement steam generator curves have been assessed by that project and determined to be acceptable.

Those curves were also generated using an ERCW temperature of 88 degrees.

Based on this review, no impacts to qualified life of electrical or mechanical equipment will result from this change.

3.7 FIRE PROTECTION The Intake Pumping Station provides a source for the high pressure fire protection water for local fire fighting. A review of HPFP system calculations determined that no hydraulic analysis utilized the maximum UHS temperature as an input, therefore the analyses and system are unaffected by the proposed change to allow WBN Unit 1 operation at 880F.

3.8 FLOOD MODE For flood mode operation, ERCW is connected through temporary spool pieces to provide long term cooling for various systems that may be inundated. In particular, the CCS system pumps are inoperable due to flooding so the ERCW piping system is connected to the CCS piping by the use of pipe spool pieces to ensure cooling water is supplied to the equipment required to be in service. However, peak river temperatures and extreme flood conditions are not postulated to occur simultaneously. Weather conditions which would bring about Tennessee River flooding would be in direct contrast to the low flows and drought conditions needed to result in high river temperatures. Therefore, the impact of increased UHS temperatures on flood mode operation was not required to be evaluated.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 3.9 SPENT FUEL POOL The Spent Fuel Pool Cooling and Cleanup System (SFPC&CS) is influenced by the ERCW System through the interfacing system CCS. Increases in ERCW system temperature have the potential to increase CCS System and ultimately SFPC&CS temperatures. In addition, raw water from the HPFP system, which is also supplied from the river, is used as emergency Spent Fuel Pool (SFP) makeup in the event all normal SFP cooling is lost. The SFPC&CS and its operation were evaluated to ensure that there were no consequences associated or anticipated with spent fuel handling, cooling, or storage. Spent Fuel Pool Cooling was evaluated analytically using an existing calculation which models CCS performance, given RHR, SFP and miscellaneous heat loads. The results of the analysis indicated that operation at 880 F was acceptable due to margin that exists in SFP heat load considerations. Refueling outages, which place the highest heat load in the SFP, are not typically conducted during the time frame associated with maximum UHS temperatures.

For this reason, the heat load demand on ERCW is lower with respect to CCS/SFP cooling. Should a refueling outage coincide with maximum UHIS temperatures, procedures are in place to limit the amount of heat placed into the SFP to assure maximum SFP design heat loads are not exceeded. Spent Fuel Pool makeup during emergency conditions was previously evaluated with makeup from a source at a higher temperature than 880 F UHS, therefore existing boil-off and time to boil analyses are bounding. Based on the discussion above, long term cooling and operation of the Spent Fuel Pool is not challenged by the proposed change to 880 F maximum UHS temperature.

3.10 AUXILIARY BUILDING SECONDARY CONTAINMENT ENCLOSURE AND AUXILIARY BUILDING GAS TREATMENT SYSTEM The Auxiliary Building Secondary Containment Enclosure is not affected by a river water temperature increase. The Auxiliary Building Gas Treatment System interface could be indirectly affected by the increased temperature since the equipment is located in areas with coolers supplied by ERCW cooling water.

However, this evaluation determined the coolers received sufficient flow to perform their design functions.

3.11 RAW COOLING WATER Raw Cooling Water (RCW), which is a non-safety grade system, was evaluated for impacts since it supports balance of plant equipment, nonsafety related coolers, and the ice making equipment. However, since the RCW system also receives its water source from the UHS, any increase in the allowable UHS temperature to 880 F will also affect RCW supply temperatures. High ambient temperatures which accompany high river water temperatures will likely result in some RCW served non-safety equipment temperature limits being closely approached. Certain components, i.e., CRDM Control Logic Cabinet, Stator Bus Cooling, have operated at their upper temperature ranges during periods of peak UHS temperature. While these cooler functions are not safety related, further evaluation and monitoring of temperature sensitive components is performed by system engineers as part of their normal responsibilities to ensure reliable and efficient plant operation.

3.12 ENVIRONMENTAL ANALYSIS A review of UFSAR and Technical Specification environmental requirements determined there were no impacts as a result of the increased UHS temperature. WBN site considerations and interactions were also looked at even though not directly related to the accident mitigation of WBN. WBN site establishes normal operating restrictions on hot water discharges to the Tennessee River. At WBN, waste heat from the main condenser is continuously transferred to the air via natural draft cooling towers. (The ERCW and RCW systems normally discharge to the cooling tower basin as make-up to the condenser circulating water system.

The flow then passes through the condenser before going through the cooling tower.) The natural draft cooling towers are not safety related. The cooling towers are not required to be in service by any Technical Specification requirement and they provide no safety function.

3.13 PLANT/SYSTEM CONDITION REVIEW Page 17

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation Also considered in the UHS temperature evaluation was a review of recent design changes and licensing changes. These included Tritium production activities (approved TS change TVA-WBN-TS-00-015),

1.4% power uprate, increased Reactor Coolant System leakage on Reactor Coolant Pump Seals, and the Replacement Steam Generators (refer to section 6.0). No impacts on these activities will result from WBN Unit 1 operation at 880F ERCW temperature. The System Health Report, Problem Evaluation Reports, NRC or generic communications, industry operating experience, and Standard Technical Specification and travelers were reviewed for impacts resulting from the temperature increase. The review did not uncover any new issues not already addressed in the evaluation.

3.14 TECHNICAL SPECIFICATION The primary controlling UHS parameter at WBN is Technical Specification (TS) 3.7.9. The LCO currently requires when the plant is in modes 1, 2, 3, or 4 that the average water temperature of the UHS must be verified once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be at or below 850F, or the plant must be in mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

NRC Inspection Manual (Part 9900 Technical Guidance, STS 375.TG) Position

1. Time averaging is not permitted outside the conditions established in the STS.

2. No allowable outage time is permitted for the UHS and this situation is described as the basis for the action statement response times.

TVA, unlike most other nuclear facilities, can exert some control over river system temperatures and flows.

Furthermore, WBN was designed as a hot stand-by plant with ERCW as the main cooling system. The ERCW system was originally designed for two-unit worst case heat loads with the faulted (accident) unit in a LBLOCA and the other unit in Hot Standby. The original design considered that if the non-accident unit is in hot shutdown or cold shutdown, it would be allowed to return to Hot Standby. All evaluations and analyses performed in support of the UHS temperature increase considered Unit 1 operating and Unit 2 in deferral, assuming worst case conditions including LOOP, LODD and failure of one electrical power train.

The Westinghouse STS typically require actions to obtain cold shutdown. However, WBN was not designed for a simultaneous dual unit shutdown from 100% power to cold shutdown. One unit is placed in Hot Standby while the other unit is brought to cold shutdown. Subsequently, the other unit is then brought to cold shutdown. The ERCW, CCS, and RHR systems can perform this function (for future two unit operation)

(both trains operable) meeting the intent (action statements) of the Technical Specification as written.

Should one train of cooling be lost, cold shutdown can still be achieved but additional time is required.

3.15 PIPING AND SUPPORTS ERCW supply lines that have continuous flow or stagnant branch lines that have the temperature from the continuous flow line extending up the stagnant line were identified for this evaluation. The affected piping analysis problems were identified and reviewed along with the impacted pipe support loads. All other ERCW supply line piping that does not have continuous flow will have a maximum operating temperature reflecting the maximum abnormal ambient environmental temperature and will have been analyzed at a temperature greater than the proposed UHS of 880F. There were no continuously flowing lines into any interfacing systems.

Any component in which the existing design margin in the ERCW mass flow rate was insufficient to suppress the ERCW exiting temperature to a value less than existing design value was reviewed by piping/support analysts for potential impact. It was determined that the impacted piping analyses and pipe supports have enough design margin to accommodate the UHS increase to 881F.

The ERCW discharge lines are designed for abnormal ambient environmental conditions except for a portion of the 30 inch diameter A & B discharge headers. The temperature of the ERCW discharge lines with a temperature increase due to a UHS of 880 F is below the abnormal ambient environmental conditions. The portion of the 30 inch diameter A and B discharge headers not designed for the abnormal ambient environmental conditions is limited to buried piping which was previously analyzed at a higher temperature value than the ERCW system design temperature. Analyses performed in support of the UHS evaluations Page 18

WATIS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation have shown that the ERCW system design temperature is not exceeded under any circumstance while operating at the elevated 88F UHS temperature. Therefore, piping analyses and pipe supports of the ERCW return lines are bounded by either the abnormal ambient environmental condition or other bounding temperatures.

Civil engineering evaluations and analyses performed in support of this evaluation have been documented in calculation N3-PA-92 Rev. 0. The results of calculation N3-PA-92 concluded that there are no adverse impacts on piping or supports from the proposed change to maximum ERCW temperature from 850 F to 880F.

3.16 Residual Heat Removal System IMPACT Since the RHR System is cooled by the Component Cooling System during normal plant operation, it is affected indirectly by ERCW temperatures. The RHR system was reviewed from the CCS interaction standpoint. The Westinghouse plant cooldown analysis also evaluated the RHR interface since containment and plant cooling is accomplished in part by using the RHR System. In addition, ERCW cooling water is supplied directly to the RHR heat exchangers during Flood Mode operation, after the unit has been shutdown. At that time, the 880F temperature would have a negligible impact on plant safety.

The results of the Westinghouse cooldown analysis indicated that with Unit 1 operational and Unit 2 deferred, cold shutdown could be achieved within 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> on 2 trains of RHR and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> on 1 train of RHR assuming an ERCW supply temperature of 880F. The Westinghouse analysis concluded that acceptable system performance during normal plant cooldowns at 880F would be obtained and compliance with TS shutdown requirement of cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was achievable under normal operation. For the case of single train cooldown, in which TS compliance with the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to cold shutdown is dependent on single train availability of RHR, the cooldown time period could be achieved by ensuring SFP cooling is isolated for up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, and the remaining reactor coolant pump is secured no later than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after shutdown. Securing the last RCP at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> is consistent with and bounded by operational practices that would occur during a LOOP, where all power isprovided by the Diesel Generators, since the Reactor Coolant Pumps are not loaded on the emergency shut down electric boards.

As for long-term containment cooling capability, the analyses have shown, as discussed above, that an increase to the UHS temperature slightly decreases the rate of cooldown thus extending the duration of the event. However, the capability to achieve TS mandated cooldown times has not been lost; therefore, there is no long-term impact on EQ limits. There are no changes affecting long term on-site or off-site dose rates or consequences. The air flow capacity of the lower compartment coolers (LCC's) for a MSLB is acceptable and not influenced by ERCW at 880F, therefore the LCC's performance in mitigating a MSLB remains acceptable. The LCC's long term cooling performance at 880 F ERCW following a MSLB is bounded by the condition where no ERCW is available or the coils are inoperable, since ERCW flow has not been credited in the analysis. Thus, the long-term EQ impact for equipment located in containment remains unchanged.

3.17 TRITIUM PRODUCTION There is a small heat load increase to the ERCW due to tritium process interactions. A review was performed of applicable analyses to ensure the heat load increase had been adequately incorporated into the SFPCCS, CCS, and ERCW system design heat load calculations and would be acceptable with the 880 F ERCW operating temperature. The heat load analysis impacts from tritium operation are conservatively based on maximum heat loads, independent of actual number of TPBARS loaded in the core. The results of the review of the analyses and the program associated with tritium production indicated no adverse impacts, as the only interface with ERCW is associated with fuel decay heat loads, which were analyzed and documented in supporting CCS calculations.

3.18 RIVER DATA A review and evaluation was performed to show the current operating margins available in the TVA river system related to failure of Chickamauga Dam. This information was developed by RSO&E. Specifically, Page 19

WATYS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation Chickamauga Dam breach size, time to breach, breach side slopes, and tail water flow parameters were modeled and evaluated for sensitivity on reservoir drawdown resulting in a time dependent reservoir drawdown curve for WBN and the remainder of Chickamauga Reservoir. The minimum river water pool elevation calculated at WBN (Tennessee River mile 528.0) is 672.9 feet which is -7 feet greater than the level (665.9 feet) used when flow balancing the ERCW system at WBN. The higher level provides somewhat higher flow capacity to ERCW cooled components due to the increased pump suction pressure.

The TVA River model developed in November 2002 was reviewed. Performance graphs for several scenarios are presented in the report. The river system was found to be capable of performing the UHS function with additional margin over that originally analyzed based upon refinements in the downstream dam loss analytical model. Also included in the data are the operating guides for Chickamauga and Watts Bar Reservoirs. Actual yearly operating average levels have been plotted. Additional Chickamauga Dam and Reservoir information is also presented.

Water released (at 14,000 CFS) from the upstream dam (Watts Bar) is assumed to be delayed up to twelve hours after a LODD. This establishes the minimum draw down time line. RSO&E has committed a minimum release from Watts Bar Dam of 14,000 CFS within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the Chickamauga Dam breach. This will provide a minimum pool elevation of 672.9 feet at the WBN intake located at Tennessee River mile 528.0.

However, the additional head available for the ERCW and HPFP pumps was not quantitatively used in any analysis developed for this task. The increased margin therefore represents additional margin available above design, but not used in the analyses. Current design analyses (prior to Feb-March 2003) are based and future design analyses will be based on a minimum Chickamauga Reservoir elevation at WBN intake of 665.9 feet. No existing Design Basis Analyses are based on the 14,000 CFS WBH outflow. No analyses in support of the UHS project utilized the higher reservoir level 672.9) associated with the 14,000 CFS WBH outfall. The seven feet of additional head represents additional margin above design.

The UHS and ERCW design basis establishes the LOCA-recirculation condition as the most limiting with a concurrent LODD (Loss of Downstream Dam). While the LOCA-recirculation that requires ERCW is not initiated immediately, the UHS / ERCW does provide cooling water to other engineered safety feature (ESF) equipment such as the diesel generators, the CCS heat exchangers, and various room coolers and chillers during this time. Long term, accident mitigation heat removal is transferred to the UHS via the ERCW system. The margin of safety as currently defined by the existing Technical Specifications and safety analysis has not been reduced by this change as has been demonstrated by new evaluations and analysis for various scenarios.

The first part of the limiting accident (LBLOCA) is the injection phase which uses stored water (from the Refueling Water Storage Tank and Condensate Tank) and the Ice Condenser for the initial mitigation. The injection phase lasts about 10 minutes when both ESF trains are operational, or one hour when one ESF train is operational. The minimum river level for design basis analysis and surveillance testing conditions is 665.9 feet assuming LODD.

The ERCW system was flow balanced during pre-operational testing at a simulated UHS level consistent with LODD and a simulated degraded ERCW pump performance level for the design basis LBLOCA mode.

The test data resulted in ERCW mass flow rate margin to all components. The flow rate values established during the preoperational testing program have been used, in part, as the basis for acceptable ERCW system operation at the proposed 880F UHS temperature. The minimum test flow value obtained in any of the test simulations, have been used in all evaluations utilizing actual ERCW flow rates. The lowest flow was normally associated with an Appendix R event alignment (see Section 3.3.1).

Without LODD, the pool elevation remains full and the delivered head from the ERCW pumps to the system is at its highest value even with minimum safeguards (single train of on-site power). If considered with LODD and loss of off-site power, the local pool is quickly turned-over and replaced by tributary reservoir water releases with an outflow of at least 14,000 CFS from the Watts Bar Dam. This minimum sustained flow isabove the low flow regime and this can provide a cooling mechanism along with reduced surface area for solar heating in upstream reservoirs, however, no credit is taken for the resulting drop in UHS temperature to below 880F.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 3.19 DELETED 3.20 OTHER OPERATING EXPERIENCE The following operating experience items were reviewed for applicability to the proposed UHS temperature change for WBN. Nothing was found in the reviews which indicated that there would be problems involved in making the proposed 850F to 880F change:

• 11/8/02 Diablo Canyon manually tripped due to indications of high dP across traveling water screens at the intake structure (ocean debris)

• 8/9/02 NOUE at North Anna due to UHS reservoir level falling due to lack of rain

• 7/1/02 Two zebra shells found in 28 SDB room chiller (PER 02-008062-000)

• NRC Information Notices 81-96, 81-91,81-92, 86-96

• Open Functional Evaluations and GL 91-18 issues associated with the UHS or ERCW The review of applicable operating experience indicated that prevailing industry issues associated with raw water cooling systems were primarily related to bio-fouling, silting, and clam shell induced plugging of heat exchangers. WBN experience in this area issimilar, in that some silting, clam shell remnants and general system bio-fouling have occurred in the raw water systems. However, WBN has implemented programs for raw water system treatment and control of bio-fouling mechanisms. In addition, procedures are in place to implement the raw water cooled heat exchanger program to ensure safety related heat exchangers that receive cooling water from the ERCW System can perform their design safety functions as required by Generic Letter 89-13.

3.21 ERCW FLOW MARGINS The flow data utilized in this evaluation for the evaluation of the equipment served by the ERCW system was obtained in 1995 during performance of pre-operational test PTI-67-02. Heat exchanger flow passages have been maintained in compliance with the requirements of NRC Generic Letter 89-13. The pre-operational and design flow rate data has been tabulated in the Tables 1A, 18, and 1C on the following pages. The higher flow rates established during the preop tests help compensate for the increased cooling water temperature in maintaining the process temperatures of the systems served by the ERCW system within their design values.

Not all component evaluations were performed using the preop test data. Specifically, the CSS heat exchanger and EDG jacket water heat exchangers utilized their design flow rates in their performance analyses, which was more conservative than using the preoperational test flow rates. Insome cases, the pre-operational test data was only used as input to the piping analysis / support analysis temperature determinations.

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WATITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88'F Maximum Operating Temperature Evaluation TABLE 1A - ERCW Flow Rates, Minimum Test Data and Acceptance Criteria Train A minimum flows, Preoperational test Minimum of Acceptance Ratio, PTI-067-02 Nominal criteria in (Minimum) /

flows, gpm gpm (Acceptance Component cooled by ERCW from Criteria)

Data Sheets I I_

Diesel Gen Heat Exchanger IAl 730.00 650 1.123 Diesel Gen Heat Exchanger 1A2 720.00 650 1.108 Diesel Gen Heat Exchanger 2A1 720.00 650 1.108 Diesel Gen Heat Exchanger 2A2 740.00 650 1.138 Shutdown Bd RM A/C A 580.95 560 1.037 Electric Bd Rm A/C A 383.58 370 1.037 Main Control Rm A/C A 313.05 293 1.068 Lower Containment Cooler 1A 320.32 306 1.047 Lower Containment Cooler 1C 320.32 306 1.047 CRDM Cooler 1A 130.36 124 1.051 CRDM Cooler 1C 133.58 124 1.077 RCP Motor Cooler 1 124.60 110 1.133 RCP Motor Cooler 3 120.00 110 1.091 CCS and AFW Pump Space Clr 1A 112.66 102 1.105 Boric Acid Trnf and AFW Sp Clr 2A 64.17 60 1.070 Instrument Room Water Clr 1A 32.16 30 1.072 SFP & TB Booster Pump CIr 1A 32.21 29 1.111 Containment Spray Pump Room Cir 1A-A 30.65 28 1.095 Centrifugal Charging Pump Cooler 1A-A 27.02 25 1.081 Upper Containment Cooler 1A 24.25 23 1.054 Upper Containment Cooler 1C 24.00 23 1.043 Safety Injection Pump Room Cooler 1A-A 24.50 22 1.114 Residual Heat Removal Pump Room Clr 1A-A 21.36 19 1.124 Pipe Chase Cooler 1A 29.70 15 1.980 Pipe Chase Cooler 2A 28.86 15 1.924 Penetration Room Cooler 1A-A e1692 21.36 12 1.780 Penetration Room Cooler I A-A e1713 22.40 11 2.036 Penetration Room Cooler IA-A e1737 22.40 12 1.867 Penetration Room Cooler 2A-A e1692 18.29 12 1.524 Penetration Room Cooler 2A-A e1713 20.82 11 1.893 Penetration Room Cooler 2A-A e1737 21.71 12 1.809 Emergency Gas Trt Rm Cir 2A 25.57 10 2.557 Station Air Compressor A Intercooler 17.90 16.5 1.085 Station Air Compressor A Aftercooler 45.24 12.4 3.648 Station Air Compressor B Intercooler 17.80 16.5 1.079 Station Air Compressor B Aftercooler 44.84 12.4 3.616 Station Air Compressor C Intercooler 18.30 16.5 1.109 Station Air Compressor C Aftercooler 44.05 12.4 3.552 Station Air Compressor D Intercooler 105.00 85 1.235 Aux Control Air Compressor A 6.85 3.5 1.957 Page 22

WAITS BAR NUCLEAR PLANT -UNIT I Ultimate Heat Sink - 88 0 F Maximum Operating Temperature Evaluation TABLE 1B - ERCW Flow Rates, Minimum Test Data and Acceptance Criteria Train B minimum flows, Preoperational Test Minimum of Acceptance Ratio PTI-067-02 Nominal Criteria in (Minimum) /

flows, gpm gpm (Acceptance from Criteria Data Sheets 8.20, 8.21, 8.22 COMPONENT Containment Spray HTX I B (gpm) 5800.00 5200 1.115 Diesel Gen Heat Exchanger 1B1 (gpm) 720.00 650 1.108 Diesel Gen Heat Exchanger 1B2 (gpm) 710.00 650 1.092 Diesel Gen Heat Exchanger 2B1 (gpm) 710.00 650 1.092 Diesel Gen Heat Exchanger 2B2 (gpm) 720.00 650 1.108 Shutdown Bd RM A/C B 580.95 560 1.037 Electric Bd Rm A/C B 396.00 370 1.070 Main Control Rm A/C B 373.70 350 1.068 Lower Containment Cooler 1B 323.51 306 1.057 Lower Containment Cooler 1D 350.89 306 1.147 CRDM Cooler 1B 133.58 124 1.077 CRDM Cooler 1D 135.68 124 1.094 RCP Motor Cooler 2 119.06 110 1.082 RCP Motor Cooler 4 124.60 110 1.133 CCS and AFW Pump Space Cir 1B 128.21 102 1.257 Boric Acid Trnf and AFW Sp CIr 2B 72.12 60 1.202 Instrument Room Water CIr 1B 32.63 30 1.088 SFP & TB Booster Pump CIr 18 33.56 29 1.157 Containment Spray Pump Room Cir 1B-B 37.53 28 1.341 Centrifugal Charging Pump Cooler 1B-B 31.62 25 1.265 Upper Containment Cooler 1B 24.00 23 1.043 Upper Containment Cooler 1D 24.50 23 1.065 Safety Injection Pump Room Cooler 1B-B 30.46 22 1.385 Residual Heat Removal Pump Room Cir 1B-B 31.68 19 1.667 Pipe Chase Cooler 1B 35.00 15 2.333 Pipe Chase Cooler 2B 35.00 15 2.333 Penetration Room Cooler 1B-B e1692 24.97 12 2.081 Penetration Room Cooler 1B-B e1713 23.40 11 2.127 Penetration Room Cooler 1B-B e1737 23.72 12 1.977 Penetration Room Cooler 2B-B e1692 27.30 12 2.275 Penetration Room Cooler 2B-B e1713 25.27 11 2.297 Penetration Room Cooler 2B-B e1737 26.45 12 2.204 Emergency Gas Trt Rm CIr 2B 27.85 10 2.785 Recip Charging Pump Rm CIr IC 23.72 12 1.977 Station Air Compressor Supply Header 234.00 171.7 1.363 Auxiliary Control Air Compressor B 10.23 3.5 2.923 Page 23

WATIS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 881F Maximum Operating Temperature Evaluation TABLE IC ERCW Flow Rates, Minimum Test Data and Acceptance Criteria CCS Heat Exchanger Flows, Nominal Acceptance Preoperational test PTI-067-02 Observed Criteria flow in gpm gpm CCS Heat Exchanger A Data Sheet 8.15.1, Appendix R Fire 4627 4425 Data Sheet 8.16, Cold Shutdown 11900 10480 Data Sheet 8.17, LOCA/LOOP 7229 6750 Data Sheet 8.18 Hot Shutdown/LOOP 10400 8325 CCS Heat Exchanger B (note 1) Note 1 Note 1 CCS Heat Exchanger C Data Sheet 8.20 Cold Shutdown 9563 5000 Data Sheet 8.21 LOCA/LOOP/LODD 7596 7000 Data Sheet 8.22 Hot Shutdown/LOOP 8172 7275 Note 1 CCS Heat Exchanger flow set at 1000 gpm in normal operation per drawing 1-47W845-5 note 32 The components cooled by the ERCW system were evaluated using the higher proposed UHS temperature of 880F. The thermal analyses included utilization of existing ERCW mass flow margins and/or existing heat load margins to evaluate acceptable heat exchanger performance. Where mass flow rate margins were insufficient to fully remove the required heat load without an increase in outlet ERCW temperature, the piping and support design associated with these specific components were further evaluated to ensure sufficient margin existed in the piping/support analyses.

The results of the quantitative analyses and qualitative assessments indicated most components would perform acceptably at the higher UHS temperature of 88'F using the nominal ERCW flows. Two components, however, were determined to have minimal or unacceptable performance.

The Shutdown Board Room (SDBR) chiller was found to have a potential performance deficiency based on discussions with the component vendor. The chiller compressors thermodynamic design point indicated 851F was the maximum recommended operational value. For this reason, WBN has implemented a vendor recommended change to re-gear the compressor to assure acceptable thermodynamic operation of the chiller unit at the higher UHS temperature of 880F.

Past performance issues associated with the Emergency Diesel Generator's (EDG) jacket water heat exchangers were resolved in a manner that also addressed the increase in UHS temperature to 880F by ensuring the heat exchangers are cleaned in the spring time of each year to ensure optimal performance during the summer months when the UHS temperatures are highest. The evaluations concluded that acceptable EDG heat exchanger performance was achievable based on design ERCW flow rates, actual EDG heat loads, and credit for actual fouling rates that would be expected in the time period in which maximum UHS temperatures occur. While no modifications to the EDG heat exchangers were determined to be warranted, a change to the EDG heat exchanger cleaning frequency and timing, now specified in the System Description and plant procedures was required to prepare the EDGs for routine UHS temperatures that may occur in the summer including the potential of reaching 880F, thus assuring acceptable EDG operation. See Note 9 to Table 2 for additional detail regarding the EDG review.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation A comparison of the required flows to the demonstrated pre-operational test data flow rates ensures that the new temperature design limit is acceptable. Also considered in the UHS temperature change evaluation was a review of recent and pending design changes and licensing changes. This included 1.4% power uprate, tritium production, proposed Shutdown Board Room and 480 Volt Board Room Air Conditioning Technical Specification Change TVA-WBN-TS-01 -08, and other current modifications through Amendment 5 of the UFSAR, including the Living UFSAR.

Component specific evaluations were conducted and documented in calculations. Tabulated in Table 2 and in the Notes to Table 2 are the results of the component specific evaluations.

TABLE 2 - COMPONENT SPECIFIC EVALUATIONS COMPONENT Notes:

PIPE CHASE COOLERS 1A, 1B 2,3,11 PIPE CHASE COOLERS 2A, 2B 2,11 PENETRATION RM COOLERS 1A-A el 692 2,3,11 PENETRATION RM COOLERS 2A-A el 692 11 PENETRATION RM COOLERS 2A-A el 713 2,11 PENETRATION RM COOLERS 2A-A el 737 2,3,11 PENETRATION RM COOLERS 1B-B el 692 2,3,11 PENETRATION RM COOLERS 2B-B el 692 and 713 2,11 PENETRATION RM COOLERS 2B-B el 737 2, 3,11 PENETRATION RM COOLER 1A-A el 713 and 737 2, 3, 4 PENETRATION RM COOLER 1B-B el 713 and 737 2, 3, 4 RHR PUMP RM COOLERS 1A-A 3, 4 RHR PUMP RM COOLERS 1B-B 3,11 CONTAINMENT SPRAY RM COOLERS 1A-A, 1B-B 3,11 SI PUMP RM COOLERS 1A-A 3, 4 SI PUMP RM COOLERS 1B-B 3,11 CENT CHARGING PUMP RM COOLERS 1A-A, 1B-B 3,4 CCS & AUX FW PUMPS SPACE COOLERS 1A-A 3, 4 CCS & AUX FW PUMPS SPACE COOLERS 1B-B 3,11 BAT& AUX FW PUMPS SP CLRS A-A, B-B 3,4 CVCS RCP CHGR PUMP ROOM CLR 1C 2, 11 CONTROL ROD DRIVE VENT COOLERS 1A, 1B, 1C, 1D 4, 6,12 RCP MOTOR CLRS 1A, 1B, 1c, 1D 4,6,12 SHUTDOWN BOARD RM A/C COOLER 1A, 1B 7,10,11 LOWER CONTAINMENT VENT COOLERS 1A, 1B, 1C, 1D 4,6,12 SFPP/TB BOOSTER PUMPS/CCS COOLERS 1A, 1B 3, 4 ELECTRICAL BOARD RM COOLERS 1A, 1B 7, 11 MAIN CONTROL ROOM A/C COOLERS 1A, 1B 7,11 INSTRUMENT RM COOLERS 1A, 1B 6, 4 DIESEL GENERATOR COOLERS 1A1, 1A2, 1B1, 1B2 9 2A1, 2A2, 2B1, 2B2 UPPER CONTAINMENT VENT COOLERS 1A, 1B, 1C, 1D 4,6,13 CCS HEAT EXCHANGERS HX A, HX B, HX C 8 CONTAINMENT SPRAY HEAT EXCH. 1A, 1B 1,11 STATION AIR COMPRESSORS COOLERS A, B, C, D 5 EGT ROOM COOLERS A-A, B-B 2,11 AUX CONTROL AIR COMPRESSOR A, B-B 5, 11 Page 25

I1KX WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation NOTES to TABLE 2 The following notes contain discussions concerning the ERCW cooled components identified during the detailed design review 1.The CSS HTX ERCW flows were evaluated by calculation. The results of the Westinghouse Containment Pressurization Analysis indicated that adequate containment cooling occurs when using the higher ERCW maximum temperature of 880F. The containment response analysis utilized the more conservative design (not demonstrated) ERCW flow rate of 5200 gpm and design CSS Hx fouling factors. Affects on piping and support analyses utilized nominal existing ERCW flow rates in determining exiting temperatures from the CSS Hx. Containment and interfacing cooling system integrity is assured with ERCW flow /temperature at these values. No credit was taken for reduced fouling, reduced tube plugging, or increased ERCW flow rates to the CSS heat exchangers.

2. Current flows to this cooler greatly exceed the minimum required, and were determined to be acceptable in calculation MDQ00006720030079. The actual ERCW flow rates exceeded the required values by at least 75%. These high flow rates ensure that the exiting ERCW temperatures are less than design, and that ample margin exists in the cooling capability.

3. These coolers were analyzed in TMG Calculation WBNOSG4-136. Room temperatures were found to be acceptable based on the revised thermal model of the Auxiliary Building, in which actual ERCW flow rates were used at the higher UHS temperatures of 880F. For some coolers, modified heat loads were used where sufficient data exists to quantify actual heat loads vs. the excessively high margin design heat loads. No credit was taken for reduced fouling, reduced tube plugging allowances, or increased (actual) air flow rates.

4. Exiting ERCW temperatures, in excess of existing Op mode analyses, did not invalidate any civil analyses, as documented in Civil Calc N3-PA-92, RO. This conclusion was based on a review of existing piping and support analyses and assessing the minor change of temperature on margins contained within the analyses.

5. Evaluated in Calculation MDQ00006720030078. This calculation utilized actual ERCW flow rates for the analysis of major components served by ERCW . This analysis primarily focused on ERCW exiting temperatures, but also evaluated acceptable heat exchanger performance in cooling the process stream.

6. Evaluated in Calculation MDQ00006720030079. This calculation utilized actual ERCW flow rates for the analysis of minor loads such as HVAC room coolers and ESF equipment. Actual heat loads were used if available that were based on actual plant data.

7. The chiller capacity is greatly oversized relative to design heat loads, as documented in calculation MDQ00006720030079. Design heat loads were typically based on highly conservative assumptions regarding electrical cable and equipment rejected heat. Documented plant experience indicated that actual heat loads were significantly less. Excess capacity ensures adequate cooling of areas served by the chiller units. No credit was taken for actual air flow rates, decreased fouling, or increased chill water cooling loop flow rates. Due to the higher ERCW mass flow rates and lower actual heat loads, there was no increase in exiting ERCW temperatures and therefore, no impact on piping or support analyses.

8. CCS Heat Exchanger analyses for the 880F ERCW evaluations were performed in Appendix B of calculation EPMJN010890 RIO. Results of the evaluation indicated that the performance and capability of the CCS Heat Exchangers and CCS System is not diminished at 881F ERCW temperature. No credit for reduced fouling or reduced tube plugging was taken. Exiting ERCW temperatures did not exceed the acceptance criteria established in EPMJN01 0890 Ri0.

9. The Emergency Diesel Generator heat exchanger performance was evaluated in calculation MDQ00008220030077. The results of the calculation indicated that EDG cooling using elevated ERCW temperatures up to and including 880F would be acceptable. The analysis utilized reduced heat loads (-5%

reduction) based on current vendor input which was validated by actual performance test data, maintaining 190OF jacket water temperature or less which is consistent with existing design, coupled with 110% loading (4840 Kw) of the EDG. While loading of the EDG up to 110% has been included in the design basis of the EDG, actions are taken to reduce the electrical loads by shedding components until the 100% loading is reached. Excessive fouling of the EDG heat exchangers, which reduces available tube plugging margin, has occurred in the past and has been documented in PER 02-013555-000. The evaluations concluded that acceptable EDG Cooler performance was achievable based on design ERCW flow rates, actual EDG heat Page 26

WATrS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation loads, jacket water shell temperature of 1900F, and credit for actual fouling rates that would be expected in the time period in which maximum UHS temperatures occur. The evaluation concluded that rate of fouling increase in the EDG heat exchanger provided acceptable EDG cooling provided tube cleaning was performed during the spring time frame of each year, assuring relatively clean EDG coolers during the late summer time period of maximum UHS temperatures. While no modifications to the DG were determined to be warranted, a change to the DG heat exchanger cleaning frequency and timing was made to ensure acceptable EDG operation. EDG cooling was deemed acceptable at 880 F ERCW based on a combination of quantitative results and engineering judgment. Exiting ERCW temperatures increased 1.20 F above existing design values based on actual ERCW flow rates, however, piping and support analyses were not adversely affected due to existing margin captured in the piping/support analyses from analyzing the piping and supports at a higher bounding temperature.

10. The SDBR Chiller was found to be non-qualified at 880 F ERCW. Discussions with the chiller vendor indicated that compressor surging/stalling may occur at the higher condenser temperatures and recommended re-gearing the compressor to modify the impeller RPM. The SDBR Chiller has been modified to provide acceptable performance at the elevated ERCW temperature of 88 0F.

11. Due to equal or higher ERCW mass flow rates and/or lower actual heat loads than design, exiting ERCW temperatures are less than or equal to design values, therefore piping and support analyses were not impacted. The analyses are shown in calculations MDQ00006720030078 and MDQ00006720030079.

12. Lower compartment temperatures are maintained by the Lower Compartment Coolers (LCC), CRDM, and RCP Motor coolers. Calculation MDQ0006720030079 evaluated the effect of the higher ERCW temperature on lower compartment cooling capability by utilizing actual plant data and extrapolating to 880F.

The results of the evaluation indicated that maximum lower compartment temperature of 120OF would not be exceeded at the higher 88 0 F UHS temperature.

13. Upper compartment temperatures are maintained by the Upper Compartment Coolers (UCC).

Calculation MDQ0006720030079 qualitatively evaluated the effect of higher ERCW temperature on upper compartment cooling capability. Actual plant experience has shown that typically, only one UCC cooler is required even during maximum UHS temperature periods of operation. The UCC coolers have significant excess heat removal capacity relative to actual heat loads in upper containment. The results of the evaluation concluded that acceptable upper compartment temperatures can be maintained during plant operation at the higher 880 F UHS temperature.

3.22 TUBE PLUGGING Heat exchanger analyses on all heat exchangers were evaluated at 88 0F assumed the current maximum permitted tube plugging, therefore, no reduction in allowable tube plugging margins was taken.

3.23 NPSH ERCW and HPFP Pump NPSH are not challenged by the incremental increase in temperature. The pumps remain functional long term even with low river level following the LODD. Pump performance has been acceptable. Routinely 1 or 2 ERCW pumps per train are run during normal operation. Any 1 pump may be out of service for overhaul. 2 pumps on 1 train provide the minimum safeguards for LOCA. Revised LODD drawdown curves increase NPSHa above design assumptions on all ERCW and HPFP pumps, thus providing additional performance margin.

3.24 REGULATORY IMPACTS/REVIEW 10CFR50 Appendix A, Criterion 44 -- Cooling water, requires a system to be provided to transfer heat from structures, systems, and components important to safety, to an UHS. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming Page 27

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation onsite power is not available) the system safety function can be accomplished, assuming a single failure.

The analyses performed to determine the acceptability of operating WBN Unit 1 at 880F UHS were performed considering a LODD, loss of power train, and the most demanding heat load and flow rate requirement placed on ERCW.

Regulatory Guide 1.27 requires that the UHS and ERCW be available for a minimum of 30 days following the event with procedural guidelines for beyond 30 days. TVA has performed analyses that show that EQ temperatures will not be exceeded for a 100 day duration even when considering operation at 880F for long term post accident period. Furthermore, it is unreasonable to assume that the UHS will actually remain at 880F for the entire 100 days or beyond with LODD. Extensive environmental heat input is required to obtain and maintain 880F ERCW along with low uncontrolled river flows (usually less than 3,500 cubic feet per second [CFS]). Reasonable margin is also created given the fact that the decay heat loads are a fraction of the initial event at 30 days and 100 days respectively. Based on these considerations, the conclusion was made that EQ temperatures are not adversely affected, and margin remains prior to reaching the EQ temperature limits.

Generic Letter 89-13 requires nuclear utilities to establish a program to assure that the heat removal requirements of their safety-related service water system (ERCW) heat exchangers are maintained.

Recommended Action II of GL 89-13 states "Conduct a test program to verify the heat transfer capability of all safety-related heat exchangers cooled by service water". WBN's program to comply with the Recommended Action II is captured in the procedure TI-79.000, "Generic Letter 89-13 Heat Exchanger Test Program". Within the site program for GL 89-13, all ERCW served safety-related heat exchangers are listed and their performance monitoring method and frequency of test is specified. The safety-related heat exchangers are monitored via the clean/inspect test method or the flow measurement test method, with frequencies ranging from 2 years down to quarterly, based on an evaluation of previous testing data. The CCS heat exchangers have heat transfer performance tests performed in conjunction with the unit's refueling outages.

The GL 89-13 program test frequency for the Emergency Diesel Generators is currently specified in TI-79.000 as each fuel cycle. This test frequency has been revised to a yearly test frequency. Data from the GL 89-13 testing program on the EDG was used to validate vendor acknowledged decreases in jacket water heat loads (See Note 9 to Table 2 in Section 3.21). The quarterly flow measurement utilized on the ESF space and room coolers allows early recognition of equipment flow deficiencies, assuring that flow rates utilized in the analysis supporting the proposed UHS temperature change are maintained, and input assumptions are not invalidated. No other impacts are foreseen and no further use of GL 89-13 program requirements or data were used in the UHS temperature increase effort.

3.25 MCR HABITABILITY The MCR air conditioners (chillers) which control the Operator's environment are cooled via the ERCW. For Equipment Qualification purposes, Reg. Guide 1.52 requires a minimum of 30 days for ESF equipment following a DBA. The TVA Tennessee River system is capable of providing water beyond the 30 days (up to one year without any rainfall) as stated in the UFSAR. As stated previously, TVA has chosen the time period to be 100 days for EQ requirements. The 100 day period remains bounding. The UHS can meet the 100 day duration without consequence. There are no effects on MCR habitability long term (30 days or more) since the ERCW provides adequate flow to the MCR A/C units under these conditions even with a LODD.

3.26 RADIOLOGICAL IMPACTS There are no changes required to existing baseline calculations and there is no impact on radiological effluents.

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WAITS BAR NUCLEAR PLANT -UNrr 1 Ultimate Heat Sink - 880F Maximum Operating Temperature Evaluation 3.27 MOTOR OPERATED VALVES The ERCW motor operated flow control valves were reviewed against the design basis valve calculations for possible temperature impact. The system design temperature (130 0F) was used in the calculations, so the calculations remain bounding and unaffected.

3.28 ASME SECTION Xl The ASME Section Xl calculations were reviewed for program impacts. The Section Xl boundary drawings and performance criteria were not impacted. Non-safety related equipment (non-accident related) and their conditions were evaluated and deemed inconsequential to operation at 880 F assuming that there is no pressure boundary leakage or system flow loss other than that analyzed by design.

3.29 INSTRUMENTATION AND CONTROL I&C components and documents were reviewed to determine associated impacts resulting from increasing the UHS operable temperature from 850 F to 880 F. Monitoring of ERCW temperatures in the MCR is required to ensure compliance with the Technical Specifications requirement on UHS. The results of the review indicated that SSD's for Instrument Loops 1-T-67-455, 1-T-67-456, 2-T-67-455, and 2-T-67-456 (ERCW Supply Header Temperature Loops) currently show Technical Specifications values at 850 F. The ICS computer alarms "High" and "High High" setpoints for ERCW SUP HEADER 1A, 1B, 2A, and

28. Affected setpoints include: T2612A, T2613A, T2614A, and T2615A. Drawing 1-47W605-243, Electrical Technical Specification Compliance Tables, ERCW Header Temperature Table, displays the Technical Specification operability limit for the UHS. When the SR 3.7.9.1 temperature readings are taken, the logs take into account the instrumentation accuracy (0.2 to 0.3F depending on the number of channels that are available). (See 1-SI-0-2-00 page 11 & 47). The appropriate documents will require revision or clarification for operation with the UHS temperature limit of 88 0F.

3.30 AUXILIARY FEEDWATER SYSTEM IMPACT The ERCW system serves as an emergency supply to the AFW system during emergency cooling operations when loss or depletion of condensate supply from the condensate storage tank(s) occurs. Use of AFW is required for mitigating Appendix R events, ATWS events, as well as normal plant shutdowns. The connection for the supply to both the motor driven and the steam turbine driven AFW pump suction is on the ERCW discharge headers. The ERCW is not loaded with significant decay heat levels until later in the cool down after RHR is placed in service during Hot Shutdown or LOCA-Recirculation. Since RHR is cooled by CCS/ERCW, increases in RHR heat load increases the CCS loop temperature and ultimately the ERCW temperature exiting the CCS heat exchangers. A review of existing Calculation EPMJNO1 0890 revealed actual temperatures in the ERCW system are well below the 120OF AFW design value during the time frame of expected AFW operation. The review and analysis of AFW System impact, documented in Calculation MDQ00006720030078, indicated that there would be no adverse impact to AFW operation due to emergency supply of ERCW at the proposed 880 F maximum supply temperature. The AFW system will continue to perform its normal safety-related design functions, including mitigating ATWS and Appendix R events, at the proposed UHS temperature of 880 F.

3.31 REGULATORY PROGRAM REVIEW Regulatory programs imposed on the nuclear industry from previous experience or industry issues were reviewed by existing processes for applicability to WBN. These programs, once they have been determined to be applicable, are addressed in the design basis through analyses, procedures, or programs specific to individual system application. While the UHS temperature increase effort did not review each regulatory program individually for impact on EROW temperatures, a review was performed of design basis calculations, drawings, and related design documentation, which did include the applicable regulatory prescribed program requirements. Examples of regulatory programs found in the design basis include the MOV program Generic Letter 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance", (See section 3.27), Generic Letter 96-06 "Assurance of Equipment operability and Containment Integrity during Page 29

WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation Design-Basis Accident Conditions", and Generic Letter 89-13 "Service Water System Problems Affecting Safety-Related Equipment' (see Section 3.24).

From the reviews performed on design basis documents, previous regulatory program initiatives were captured in the overall UHS temperature increase task. No adverse impacts to previously considered regulatory programs were noted.

3.32 DESIGN BASIS CONTROL The overall design of the UHS temperature increase effort has been to utilize existing margins in the ERCW and plant design basis in proving acceptable plant operation and safety performance for a short duration and an infrequent event in which the UHS temperature limit could be as high as 880F. The approach taken examined applicable analyses and demonstrated that safe operation of WBN can be achieved if the UHS temperature reached 880F. As a result of this effort, the ERCW and UHS design bases will be revised. A new abnormal operating temperature of 880F will be defined to reflect a short duration event related to seasonal variations in UHS temperature above the normal operating design limit of 850F. In recognition that certain key analyses are critical in establishing the design basis of components and system performance important to safety, the Containment Pressurization and Temperature response analysis, as well as the plant cool down analysis, were revised to incorporate a new design basis of 880F (at all times) for ERCW temperature.

Since ERCW flow margins above the existing design flow requirements were utilized in validating acceptable performance at the higher ERCW temperature, specific evaluations will be taken prior to unit operation above 850F. The performance of these specific evaluations will validate any margin based inputs utilized in the original analyses that determined acceptable performance could be achieved at the higher ERCW temperature. The evaluations will also assure that no degraded or non-conforming conditions exist that would invalidate the conclusions of the original analyses. To ensure consistency between RCW / ERCW /

UHS / HPFP temperatures, the design, specification, and procurement of new components shall be based on an 880F design basis ERCW / UHS temperature to ensure acceptable operation at the higher allowable UHS temperature of 880F. Performance of future design analyses shall consider the impact of 3TF temperature increase above the previous 850F ERCW System design basis to ensure the conclusions of the 880 F UHS temperature evaluation are not invalidated by subsequent design changes.

The results indicate that operation of WBN Unit 1 isacceptable at the higher ERCW/UHS temperature of 880F. Documentation for the acceptability of the ERCW supplied components and piping is provided in Calculations No. MDQ0000672003 0078, MDQ00006720030079, and N3-PA-092. These calculations in turn reference the calculations and documents which were reviewed and evaluated in the process of determining acceptable performance at the higher 880F temperature, including several new calculations issued for this change. In addition to the calculations and documents referenced in calculations MDQ0000672003 0078, MDQ00006720030079, and N3-PA-092, there are numerous other input and cross-reference documents and calculations that are contained the references of these calculations. Examples include individual pipe support calculation sheets, design drawings, purchase specifications, bills of material, controlled vendor manuals, environmental qualification specifications, etc. These lower tier documents may continue to reflect the original 850F ERCW/UHS design basis normal operating temperature limit rather than the new short duration design basis temperature of 880F unless revised in the future when the document is being revised to support other design changes. As mentioned, several key calculations have been revised to support this project as well as other design basis documents and the WBN UFSAR which will be revised to accurately reflect the completed evaluation and future design controls.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 88 0 F Maximum Operating Temperature Evaluation 4.0

SUMMARY

OF RESULTS During the hottest period of the summer, the impact on the region is directly felt in increasing electrical demand with a corresponding increase in the demand on the Tennessee River watershed to provide cooling water to various industries. Hotter than normal summers and below normal river flows provide additional challenges to river operations. Evaluations described in this document provide assurance that margins exist in the operation of WBN with an UHS temperature up to and including 880F. It is concluded there is low risk to the health and safety of the public and to the personnel and equipment at WBN if the plant is operated with an UHS temperature of 880F. This evaluation summarizes the various reviews of the limiting components and finds that WBN plant operation up to an ERCW supply temperature of 880F is acceptable for all plant conditions and accident scenarios. A design change has been initiated that will incorporate changes to the affected design basis documents (Design Criteria, Calculations, System Descriptions, drawings, etc.) and the licensing basis documents (UFSAR and Technical Specifications), subsequent to NRC approval of the requested Technical Specification Change.

The effect of the proposed 30F UHS temperature increase to 880F on equipment, components, systems, and safety analysis has been examined in detail and described in Section 3.0. The proposed change has been demonstrated to be an acceptable increase to the UHS temperature limit based on available margins in three areas, specifically, 1)there are margins in the river system water level for LODD scenarios; 2) there are margins to the allowed maximum peak pressure in the containment design based on containment reanalysis, and 3) there are sufficient margins in the ERCW system flow rates to each affected component.

There is no increase in risk to the public health, to equipment, or the plant site. Operating equipment performance weaknesses, degraded conditions, and system health have been considered in this evaluation.

There is no increase in risk to shutdown, inspections, system alignments, or operator burden for this plant condition. No component is affected by the proposed change in a way that would result in transient of a different kind than that previously evaluated. No new analysis methodologies were utilized in any evaluation performed in support of this task, and methodologies utilized for certain critical analyses are consistent with methodologies previously reviewed and approved by the NRC. Existing procedures are adequate to control plant activities.

5.0 RECOMMENDATIONS 5.1 Technical Specification 3.7.9 and the maximum temperature can be raised to 880F.

5.2 Issue Engineering Design Change (EDC) to revise Setpoint Scaling Documents, System Descriptions, Design Criteria, Drawings, Procedures, Instructions, to ensure that 880 F is considered in the future designs used in procurements, specifications, analyses, and evaluations. The UFSAR will be revised accordingly.

5.3 Continuined monitoring of RCW cooled components is warranted to assess any limiting components on continued plant operation.

5.4 Revise the affected documents to address single-train RHR cooldown restrictions for an ERCW temperature of 880F which consist of a minimum five hour isolation time for spent fuel pool cooling and a requirement to secure the remaining reactor coolant pump within 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after shutdown.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation 6.0 Revalidation Effort The previous version of this evaluation was issued in April 2004. In order to support this license amendment, a revalidation was performed to ensure that the original conclusions remain accurate.

This revalidation focused on identifying and assessing any potential changes to the facility between 2003 and 2006 which might affect function of the ERCW system under peak river temperature conditions. Where applicable, changes since the initial preoperational tests were considered. A summary of the revalidation results is provided below:

ISSUE CONCLUSIONS Silt Accumulation An adequate program is in place to ensure that silt accumulation does not occur in such a manner as to result in degradation of system performance. This program includes periodic high velocity online flushes, additional flushing during refueling outages, ultrasonic testing to detect buildup in key susceptible areas, and periodic inspection/cleaning of the IPS pits.

Macrofouling Problems in this area early in the plant life led to development of an aggressive chemical treatment program which has been successful at eliminating any recurrences.

No degradation is predicted due to this phenomenon.

Microbiologically Even though chemical treatment is routinely performed, Influenced Corrosion some MIC presence in low flow areas is unavoidable.

(MIC) However, design of the system is such that the presence of MIC nodules is very unlikely to affect system hydraulic performance.

Heat Exchangers Performance monitoring is performed on safety related heat exchangers as part of the GL 89-13 program. This ensures that heat transfer performance will continue to meet system requirements. All affected heat exchangers are within pre-established tube plugging margins. Review of this program did not identify any aspects which might invalidate the conclusions of this evaluation.

Pumps Performance test data for the ERCW pumps was compared with data obtained during pre-operational testing. No appreciable reductions in output were noted except for a 3% decline in output head for the A-A pump.

However, this pump still retains greater than 10% margin over the design basis minimum head, therefore sufficient margin remains to support operation at 88 degrees.

Component Margins Calculations MDQ00006720030078 and MDQ00006720030079 were issued to support revision 0 of this evaluation. They demonstrate adequacy of the ERCW system to provide the required cooling with an 88 degree UHS temperature. All revisions to referenced input documents for these calculations were reviewed. No impacts were identified which would affect the conclusions Page 32

WATYS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation ISSUE CONCLUSIONS reached.

Throttle Valve Changes Throttle valves control ERCW flow to many components.

Current valve positions documented in TI-31.08 were compared with the positions of the same valves documented during pre-operational testing. A few changes were identified. These were evaluated and shown to provide sufficient flow margin to support operation at 88 degrees.

Replacement Steam The replacement steam generators to be installed during Generator Impacts RFO7 will have various minor impacts on post-accident conditions inside containment. These impacts have been evaluated and shown to be acceptable considering the new 88 degree peak temperature for ERCW.

Problem Evaluation Open PERs were reviewed which were identified against Reports systems 030, 031, 032, 067, 070, 078, or 003 or which contained any of the following words:

silt, clams, debris, Raw, heat sink, ERCW, 89-13, chiller, and heat exchanger.

No open PER conditions were found which might impact the conclusions of this evaluation.

GL 91-18 Conditions Open GL 91-18 conditions were reviewed with no impacts identified.

The significant GL 91-18 condition involving the release of air entrained in the ERCW cooling water resulted in voids in the ERCW discharge headers. An evaluation determined that the temperature increase did not adversely affect the condition.

Design Changes Design change packages issued since revision 0 of this evaluation were reviewed. No changes were identified which might impact the conclusions of this evaluation.

Surveillance Instructions Several surveillance instructions will be impacted by the proposed Technical Specification change. These will be revised during the implementation process.

Equipment Qualification Conclusions reached in revision 0 of this evaluation were reviewed and verified including potential impact on mechanical equipment qualification.

UFSAR Amendments Amendments 4 and 5 were reviewed. No additional relevant issues were identified.

NRC Generic Letters Generic letters issued 2003 - 2006 were reviewed. No additional relevant issues were identified.

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WATTS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation Based on this revalidation it was concluded that the ERCW system remains capable of performing all required design functions with an ultimate heat sink maximum temperature of 880F.

7.0 REFERENCES

WBN Technical Specification, Plant Systems, 3.7.9 Ultimate Heat Sink UFSAR Section 9 AOI-13 Loss of ERCW AOI-14 Loss of RHR Shutdown Cooling AOl-15 Loss of Component Cooling Water AOI-22 Break of Downstream Dam SO1-67.01 ERCW System S01-70.01 Component Cooling Water System SO-74.01 Residual Heat Removal System SOI-3B.01 Auxiliary Feedwater System GO-6 Unit Shutdown from Hot Standby to Cold Shutdown 1-47W845 Series MECH Flow Diagrams ERCW 1-47W865-3 Flow Diagram AC Chilled Water 1-47W865-7 Flow Diagram AC Chilled Water TI-79.000 GL 89-13 Heat Exchanger Test Program WM-28-1-85-100 Effect of WBNP and WBSteamPlant Discharges on Chickamauga Reservoir Water Temperature PTI-67-02 ERCW Preoperational Test Instruction/Record N3-78-4001 SFPCCS System Description N3-70-4002 CCS System Description N3-67-4002 ERCW System Description N3-74-4001 RHR System Description N3-72-4001 CSS System Description N3-84-4001 Flood Mode Boration System Description N3-03B-4002 Auxiliary Feed Water System Description N3-30AB-4001 Aux. Bldg HVAC System Description N3-30CB-4002 Cont. Bldg HVAC System Description N3-30ADB-4002 Addtl D/Generator Bldg HVAC System Description N3-30DB-4002 Diesel Generator Bldg HVAC System Description N3-30PS-4002 Pumping Station HVAC System Description N3-30RB-4002 Reactor Bldg HVAC System Description WB-DC-20-20 Traveling Water Screens & Trashracks Design Criteria WB-DC-40-29 Flood Protection Provisions Design Criteria WB-DC-40-37 Heat Rejection System Design Criteria WB-DC-40-63 Raw Cooling Water System Design Criteria WB-DC-40-28.2 Additional Diesel Generator Building Environmental Control Design Criteria EPMRCT121490 Calculation: ERCW Maximum Rejected Heat Load Requirement EPMJFL120285 Calculation: ERCW System Flow Requirements EPMSC1 01987 Calculation: ERCW Temperature at the Outlets of various Space Coolers, Air Conditioning Condensers and Heat Exchangers EPMVA043092 Calculation: Analysis of Extreme Intake Water Temperature MDQ00107020010069 Calculation: Determination of CCS Heat Exchanger C Performance under Fouled Conditions MDQ1070000061 Calculation: CCS Heat Exchanger STER Validation EPMJN01 0890 Calculation: Performance of CCS Heat Exchangers MDQ00008220030077 Calculation: Emergency Diesel Generator Jacket Water HX Evaluation MDQ00006720030078 Calculation: 880F UHS Impact on ERCW Cooled Components MDQ00006720030079 Calculation: 881F UHS Impact on ERCW Cooled ESF and HVAC Equipment WBNOSG4-136 Calculation: Transient AB TMG Model N3-PA-092 Calculation: Evaluation of Impact of 880F ERCW (UHS) on Safety Related Piping and Pipe Supports WBN Technical Requirements Manual (Not affected)

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WAITS BAR NUCLEAR PLANT -UNIT 1 Ultimate Heat Sink - 880 F Maximum Operating Temperature Evaluation Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants, R1, Mar. 1974; and R2, Jan. 1976 UFSAR 2.4.4 Potential Dam Failure, Seismically Induced UFSAR 2.4.11.1 Low Flow in Rivers and Streams UFSAR 2.3 Meteorology UFSAR 2.4 Hydrologic Engineering UFSAR Appendix 2.4A Flood Protection Plan UFSAR 2.5 Geology, Seismology, and Geotechnical Engineering Summary of Foundation Conditions UFSAR 6.0 Engineered Safety Features UFSAR 6.2 Containment Systems UFSAR 9.2.1 Essential Raw Cooling Water (ERCW)

UFSAR 9.2.2 Component Cooling System UFSAR Fig 9.2.2-Series, Essential Raw Cooling Water Flow Diagrams UFSAR 9.2.8 Raw Cooling Water System UFSAR 9.3 Process Auxiliaries UFSAR 9.4 Air-Conditioning, Heating, Cooling and Ventilation Systems WBN System 67 Health Reports IE Circular 78-13 Inoperability of Multiple Service Water Pumps IN 81-21 Potential Loss of Direct Access to Ultimate Heat Sink IN 86-96 Heat Exchanger Fouling Causing Inadequate Operability of Service Water Systems IN 88-37 Flow Blockage of Cooling Water to Safety System Components (Asiatic clams)

IN 88-65 Plant Operation Beyond Analyzed Conditions GL 89-13 Service Water System Problems Affecting Safety-Related Equipment GL 89-13 Supplement 1,Service Water System Problems Affecting Safety-Related Equipment NRC Inspection Manual, Inspection Procedure 71111.07, Heat Sink Performance NRC Inspection Manual, Part 9900, Technical Guidance, Section 3.7.5, Ultimate Heat Sink Technical Specification TF-330 R3, Industry/TSTF Standard Technical Specification Change Traveler, Allowed Outage Time -Ultimate Heat Sink Information Notice No. 81-21: Potential Loss of Direct Access to Ultimate Heat Sink Description of Circumstances: IE Bulletin 81-03, issued April 10, 1981, requested licensees to take certain actions to prevent and detect flow blockage caused by Asiatic clams and mussels.

I-SI-OPS-000-002.0, shift log, Revision 64 1- 0-PI-OPS-000-666.0'R0, River Temperature Limits Specified by WBN site specific NPDES Permit 0-TI-SXX-000-146.ORO Program for Implementing NRC Generic Letter 89-13 Updated Predictions of Chickamauga Reservoir Recession Resulting From Postulated Failure of the South Embankment at Chickamauga Dam, November 2002, River Systems Operations & Engineering Monitoring & Moderating WATTS BAR Ultimate Heat Sink, River Systems Operations & Engineering, April 2006 Page 35

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 Proposed Technical Specification Changes (mark-up)

I. AFFECTED PAGE LIST 3.7-21 II. MARKED PAGES See attached.

UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verf 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> U ~*iP880F.

Watts Bar-Unit 1 3.7-21

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 Proposed Technical Specification Bases Changes (mark-up)

I. AFFECTED PAGE LIST B 3.6-28 B 3.6-37 B 3.7-39 B 3.7-44 B 3.7-48 B 3.7-49 B 3.7-50 II. MARKED PAGES See attached.

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential

(-2.0 psid) with respect to the Shield Building annulus atmosphere in the event of inadvertent actuation of the Containment Spray System or Air Return Fans.

Containment pressure is a process variable that is monitored and controlled.

The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.

APPLICABLE Containment internal pressure is an initial condition used in the DBA SAFETY ANALYSES analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients. The worst case LOCA generates larger mass and energy release than the worst case SLB.

Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. 1).

The initial pressure condition used in the containment 5.0 psia.

This resulted in a maximum peak pressure from a LO f 40 64 10.9 ps The containment analysis (Ref. 1) shows that the maxi allowable inter 2 containment pressure, Pa (15.0 psig), bounds the calcu limiting LOCA. The maximum containment pressure resulting roWthl worst case LOCA, does not exceed the containment design pressure, 13.5 psig.

(continued)

Watts Bar-Unit 1 B 3.6-28 Revision 44, 55 Amendment 33

Containment Spray System B 3.6.6 BASES BACKGROUND and water from a DBA. During the post blowdown period, the (continued) Air Return System (ARS) isautomatically started. The ARS returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam through the ice condenser, where heat is removed by the remaining ice and by the Containment Spray System after the ice has melted.

The Containment Spray System limits the temperature and pressure that could be expected following a DBA. Protection of containment integrity limits leakage of fission product radioactivity from containment to the environment.

APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES OPERABILITY are the loss of coolant accident (LOCA) and the steam line break (SLB). The DBA LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of the Containment Spray System, TI~h and the ARS being rendered inoperable (Ref. 2).

The DBA analyseshow that the maximum peak containment pressure of I a 464 10.9 ps s 1ts from the LOCA analysis, and is calculated to be less th n the q 'nnntdesign pressure. The maximum peak containment tenperature results from the SLB analysis. The calculated transient containment atmosphere temperatures are acceptable for the DBA SLB.

(continued)

Watts Bar-Unit 1 B 3.6-37 Revision 44, 55 Amendment 33

COS B 3.7.7 BASES (continued)

APPLICABLE The design basis of the CCS is for one CCS train to remove the post loss of SAFETY ANALYSES coolant accident (LOCA) heat load from the containment sump during the recirculation phase, with a maximum COS temperature of 11 00 F (Ref. 2). The Emergency Core Cooling System (ECCS) LOCA and containment OPERABILITY LOCA each model the maximum and minimum performance of the CCS, respectively. The normal temperature of the CCS is 951F, and, during unit cooldown to MODE 5 (Tcold < 2000 F), a maximum temperature of 1 100 F is assumed. The CCS design based on these values, bounds the post-accident conditions such that the sump fluid will not increase in temperature after alignment of the RHR heat exchangers during the recirculation phase following a LOCA, and provides a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System (RCS) by the ECCS pumps.

The COS is designed to perform its function with a single failure of any active component, assuming a loss of offsite power.

The COS also functions to cool the unit from RHR entry conditions (Tcold < 350 0F),

to MODE 5 (Tc:d 0 < 200'F), during normal and post accident operations. The time required to cool from 3500 F to 2001F is a function of the number of COS and RHR trains op ~n is sufficient to remove decay heat during subseg t operations with < 200'F. This assumes a maximum EROW temp~ure of 88- 880 F osccg simultaneously with the maximum heat loads on the sy The CCS satisfies LCO The C0S trains are independent of each other to the degree that each has separate controls and power supplies and the operation of one does not depend on the other. In the event of a DBA, one COS train is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. To ensure this requirement is met, two trains of COS must be OPERABLE. At least one CCS train will operate assuming the worst case single active failure occurs coincident with a loss of offsite power.

(continued)

Watts Bar-Unit 1 8 3.7-39

4 ERCW B 3.7.8 BASES BACKGROUND .(Ref. 1). The principal safety related function of the ERCW System is the (continued) removal of decay heat from the reactor via the CCS.

APPLICABLE The design basis of the ERCW System is for one ERCW train, in conjunction SAFETY ANALYSES with the CCS and a 100% capacity Containment Spray System and Residual Heat Removal (RH R), to remove core decay heat following a design basis LOCA as discussed in the FSAR, Section 9.2.1 (Ref. 1). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The ERCW System is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.

The ERCW System, in conjunction with the CCS, also cools the unit from RHR, as discussed in the FSAR, Section 5.5.7, (Ref. 2) entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCS and RHR System trains that are operatinn is sufficient to remove decay heat during subsequetp~ nsinlMOD~ 5 and 6. This assumes a maximum ERCW temperature of 88- 880 F oc rrinssimultaneously with maximum heat loads on the st_>_f-y The ERCW S iyset.t~fi- s Criterion 3 of the NRC Policy Statement.

LCO Two ERCW trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.

An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

(continued)

Watts Bar-Unit 1 B 3.7-44

UHS B 3.7.9 B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for processing and operating heat from safety related components during a transient or accident, as well as during normal operation. This is done by utilizing the Essential Raw Cooling Water (ERCW)

System and the Component Cooling System (CCS).

The UHS is defined as the Tennessee River, including the TVA controlled dams upstream of the intake structure, Chickamauga Dam (the nearest downstream Pant intake channel, not including the intake structure, as tdlussedi FS Section 9.2.5 (Ref. 1). The maximum UHS temperature of 88F ensures equate heat load removal capacity for a minimum of 30 days aftreactor s s ~wn or a shutdown following an accident, including a Loss of la (OCA).

Additional information on the design and operation of the system, along with a list of components served, can be found in Reference 1.

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational occurrences in which the unit is cooled down and placed on residual heat removal (RHR) operation. Its maximum post accident heat load occurs approximately 20 minutes after a design basis loss of coolant accident (LOCA). Near this time, the unit switches from injection to recirculation and the containment cooling systems and RHR are required to remove the core decay heat.

The operating limits are based on conservative heat transfer analyses for the worst case LOCA. Reference 1 provides the details of the assumptions used in the analysis, which include worst expected meteorological conditions, conservative uncertainties when calculating decay heat, and worst case single active failure.

The UHS is designed in accordance with Regulatory Guide 1.27 (Ref. 2), which Watts Bar-Unit 1 B 3.7-48 Revision 29

0 UHS B 3.7.9 BASES APPLICABLE requires a 30 day supply of cooling water in the UHS.

SAFETY ANALYSES (continued) The UHS satisfies Criterion 3 of the NRC Policy Statement.

LCO The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and witho imum design temperature of the equipment served by t 44CW Sy tem. Toalet this condition, the UHS temperature should ntxceed 88 88'F.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

ACTIONS A.1 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies that the ERCW System is available to cool the CCS to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency Watts Bar-Unit 1 B 3.7-49 Revision 29

a UHS B 3.7.9 BASES SURVEILLANCE SR 3.7.9.1 (continued)

REQUIREMENTS is based on operating experi eeg of the parameter variations during the applic IaMODES. This SR'rifies that the average water temperature of the t is < _88* not account for instrument error, Ref. 3).

I REFERENCES 1. Watts Bar FSAR, Section 9.2.5, "Ultimate Heat Sink."

2. Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants,"

Revision 1, March 1974.

3. Watts Bar Drawing 1-47W605-243, "Electrical Tech Spec Compliance Tables."

Watts Bar-Unit 1 B 3.7-50 Revision 29

ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 List of Comuitments

1) The UFSAR will be revised to require the Emergency Diesel Generator jacket water heat exchangers to be cleaned once each spring.

2) The UFSAR will be revised to address single-train RHR cooldown restrictions for ERCW temperatures above 85 degrees F which consist of a minimum five hour isolation time for spent fuel pool cooling and a requirement to secure the remaining reactor coolant pump within 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after shutdown.

3) Revise System Descriptions, Design Criteria, and other key design documents and the UFSAR to implement operation above 85 degrees F and the new UHS design basis of 88 degrees F.