ML18038B958

Application for Amends to Licenses DPR-52 & DPR-68, Consisting of TS Change 384,allowing Bfn,Units 2 & 3,to Operate at Uprated Power Level of 3458 Mwt.Proprietary Rept Re Power Uprate Safety Analysis,Also Encl.Rept Withheld

ML18038B958
Person / Time
Site:	Browns Ferry
Issue date:	10/01/1997
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18038B959	List: ML18038B958 ML18038B960 ML18038B961
References
TVA-BFN-TS-384, NUDOCS 9710070314
Download: ML18038B958 (41)

Text

CATEGORY 1 R GULA, Y INFORMATION DISTRIBUTI SYSTEM (RIDS)

ACCESSION NBR:9710070314 DOC.DATE: 97/10/Ol NOTARIZED: YES DOCKET ¹ FACIL 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION ABNEY,T.E. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFXLIATION gang Document Control Branch (Document Control Desk)~EE,~~ 7~

SUBJECT:

Application for amends to licenses DPR-52 a DPR-68, consisting of TS Change 384,allowing BFN,Units 2 6 3,to operate at uprated power level of 3458MWt.Proprietary rept re power uprate safety analysis, also encl. Rept withheld. A DISTRIBUTION CODE: D030D COPIES RECEIVED:LTR 3 ENCL t SIZE: + 8I T TITLE: TVA Facilities Routine Correspondence E

NOTES:

RECIPIENT COPIES RECIPIENT COPXES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3 1 1 PD2-3-PD 1 1 WILLIAMS,J. 1 1 INTERNAL: ACRS 1 1 LE CENTER 1 1 OGC/HDS3 1 0 RE B SES 1 1 EXTERNAL: NOAC g NRC PDR 1 0 D

E NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 41S-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL

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Tennessee Valley Authority, Post Office Box 2000, Decatur, Afabama 35609-2000 October 1, 1997 TVA-BFN-TS-384 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of Docket Nos. 50-260 Tennessee Valley Authority 50-296, BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 and 3 TECHNICAL SPECIFICATION (TS) CHANGE TS-384 REQUEST FOR LICENSE AMENDMENT FOR POWER UPRATE OPERATION.

In accordance with the provisions of 10 CFR 50.54 and 50.90, TVA is submitting a request for, an amendment (TS-384) to Operating Licenses DPR-52 and DPR-68 and Appendices A thereto, for the Browns Ferry Nuclear Plant, Units 2 and 3, respectively. The proposed changes will allow these two units to operate at an uprated power level of 3458 MWt. This represents a power level increase of five percent.

This proposal follows the generic guidelines for BWR power r uprate described in References 1 and 2. Enclosure 1 contains a detailed description of the specific proposed changes necessary for power uprate operation and the technical bases for these changes. Enclosure 2 is an Environmental g gp Assessment (EA) and Finding of No Significant Impact (FONSI).

The Assessment was prepared in accordance with the National Environmental Policy Act (NEPA) and TVA's implementing procedures. It addresses specific issues and potential environmental impacts associated with power uprate at BFN.

E closg 3 contains the affected page list and copies of the L~M.

lllilllllllllliiiJlllllllillllll/IilIiil, PDR ADGCK 05000260 p PDR

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U.S. Nuclear Regulatory Commission Page 2 October 1, 1997 appropriate Operating License and TS pages for Units 2 and 3 marked-up to show the proposed changes. Enclosure 4 contains the affected page list and copies of the appropriate Operating License and TS pages for Units 2 and 3 revised to show the proposed changes. The TS markups and revised pages in Enclosures 3 and 4 are based on the proposed BFN Improved Standard Technical Specifications (Reference 3) including the changes proposed in References 4, 5 and 6. Enclosure 5 contains the detailed plant-specific submittal information required by the generic guidelines (References 1 and 2).

Portions of the report are proprietary and these portions should be withheld from public disclosure in accordance with 10 CFR 2.770(a)(4). The affidavit supporting the request is in accordance with 10 CFR 2.790(b)(1) is provided in . Section 11.4 of Enclosure 5 also contains a non-proprietary no significant hazards evaluation for the proposed changes. Enclosure 7 provides the commitments made in this letter.

It should be noted that TVA has not completed the Environmental Qualification (EQ) of electrical equipment portion of its power uprate analysis. The EQ software program establishing environmental conditions currently in the BFN Licensing Basis approved by the staff is the MONSTER computer model. The MONSTER software for EQ evaluations is no longer available. Therefore, an updated code, GOTHIC, will be used as the basis to evaluate changes in EQ due to uprate. Due to uprated conditions, it is expected that the use of this updated model will require revised TS and possible modifications. TVA expects to issue a TS change to NRC detailing these changes by first quarter of 1998.

Any necessary power uprate modifications are scheduled to be implemented during the Cycle 8 outage in October of 1998 for Unit 3 and the Cycle 10 outage in March of 1999 for Unit 2.

Therefore, TVA requests approval of the proposed changes for Unit 3 by July 1, 1998 and Unit 2 by November 1, 1998. NRC approval three months before the Unit 3,Cycle 8 outage is necessary to support pre-outage planning and TVA management approval for required power uprate modifications. Examples include:

a. Core Reload Desi n Power uprate operation requires additional fuel and the reload design and core loading pattern are established prior to the outage.

U.S. Nuclear Regulatory Commission Page 3 Oct:ober 1, 1997

b. Desi n Chan es/Maintenance Work Orders Power uprate implementation involves resetting the main steam relief valve setpoints as well as recalibration and resetting of plant instrumentation.

TVA requests that the proposed license amendments become effective before startup in Cycle 9 in November of 1998 for Unit 3 and before startup in Cycle 11 in April of 1999 for Unit 2.

In accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.

If you have any questions about this change, please telephone me at (205) 729-2636.

Si cerely T. . ne Manager of Lice n on and Industry,Affai Subscribed and this Notary Public

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U.S. Nuclear Regulatory Commission Page 4 October 1, 1997

References:

1. NEDC-31897P-A, "Generic Guidelines For General Electric Boiling Water Reactor Power Uprate," May 1992.

2. NEDC-31894P, "Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," July 1991; and Supplements.

3. TVA letter to NRC dated, September 6, 1996, in regards to TVA-BFN-TS-362, Browns Ferry Nuclear Plant (BFN)

Units 1, 2, and 3 Technical Specification (TS) Change TS-362 Request to Convert Current TSs to Improved Standard TS (ISTS) Consistent With NUREG-1433, Revision 1.

4. TVA Letter to NRC dated April 11, 1997, in regards to TVA-BFN-TS-353S1, Browns Ferry Nuclear Plant (BFN)

Units 1, 2, and 3 Technical Specifications (TS)

Change TS-353S1 Power Range Neutron Monitor (PRNM)

Upgrade With Implementation of Average Power Range Monitor (APRM) And Rod Block Monitor (RBM) TS (ARTS)

Improvements And Maximum Extended Load Line Limit (MELLL) Analyses Supplement 1 Improved Standard Technical Specifications (ISTS) Format.

5. TVA Letter to NRC dated May 1, 1997, In regards to Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Technical Specification (TS) 362 Improved Technical Specifications (ITS) Supplement 1, Section 3.8.

6. TVA letter to NRC dated December 11, 1996, Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Technical Specification (TS) Set Point Requirements for Reactor Coolant system Integrity, TS 2.2.A.

U.S. Nuclear Regulatory Commission Page 5 October 1, 1997 (Enclosures):

Chairman Limestone County Commission 310 West Washington Street Athens, Alabama 35611 Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.

Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Dr. Donald E. Williamson State Health Officer Alabama State Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-3017

ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384 GE AFFIDAVIT FOR NEDC-32751P (See attached.)

General Electric Company AFFIDAVIT I, George B. Stramback, being duly sworn, depose and state as follows:

(1) I am Project Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report NEDC-32751P, Power Uprate Safety Analysis for the Browns Ferry Nuclear Plant Units 2 ck 3, Class III (GE Proprietary Information), dated September 1997. This document, taken as a whole, constitutes a proprietary compilation of information, some of it also independently proprietary, prepared by the General Electric Company. The independently proprietary elements are delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, C 97992d8777DCC7.79927, d

~v~, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies; GBS-97-6-afTVApu l.doc AffidavitPage 1

b. Information which,'if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Both the compilation as a whole and the marked independently proprietary elements incorporated in that compilation are considered proprietary for the reason described in items (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

That information (both the entire body of information in the form compiled in this document, and the marked individual proprietary elements) is of a sort customarily held in'confidence by GE, and has, to the best of my knowledge, consistently been held in confidence by GE, has not been publicly disclosed, and is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager,project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

GBS-97-6-afTVApu 1.doc AffidavitPage 2

(8) The information identified by bars in the margin is classified as proprietary because it contains detailed results and conclusions from these evaluations, utilizing analytical models and methods, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of transient and accident events in the GE Boiling Water Reactor ("BWR"). The development and approval of these system, component, and thermal hydraulic models and computer codes was achieved at a significant cost to GE, on the order of several million dollars.

The remainder of the information identified in paragraph (2), above, is classified as proprietary because it constitutes a confidential compilation of information, including detailed results of analytical models, methods, and processes, including computer codes, and conclusions from these applications, which represent, as a whole, an integrated process or approach which GE has developed, obtained NRC approval of, and applied to perform evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of a given increase in licensed power output for a GE BWR. The development and approval of this overall approach was achieved at a significant additional cost to GE, in excess of a million dollars, over and above the very large cost of developing the underlying individual proprietary analyses.'o effect a change to the licensing basis of a plant requires a thorough evaluation of the impact of the change on all postulated accident and transient events, and all other regulatory requirements and commitments included in the plant's FSAR. The analytical process to perform and document these evaluations for a proposed power uprate was developed at a substantial investment in GE resources and expertise. The results from these evaluations identify those BWR systems and components, and those postulated events, which are impacted by the changes required to accommodate operation at increased power levels, and, just as importantly, those which are ~ so impacted, and the technical justification for not considering the latter in changing the licensing basis. The scope thus determined forms the basis for GE's offerings to support utilities in both performing analyses and providing licensing consulting services. Clearly, the scope and magnitude of effort of any attempt by a competitor to effect a similar licensing change can be narrowed considerably based upon these results. Having invested in the initial evaluations and developed the solution strategy and process described in the subject document GE derives an important competitive advantage in selling and performing these services.

However, the mere knowledge of the impact on each system and component reveals the process, and provides a guide to the solution strategy.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive GBS-97-6-atTVApu t.doc AffidavitPage 3

physical database and "analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods, including justifications for not including certain analyses in applications to change the licensing basis.

if GE's competitive advantage will be lost its competitors are able to use the results of the GE experience to avoid fruitless avenues, or to normalize or verify their own process, or to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. In particular, the specific areas addressed by any document and submittal to support a change in the safety or licensing bases of the plant will clearly reveal those areas where detailed evaluations must be performed and specific analyses revised, and also, by omission, reveal those areas

~ so affected.

While some of the. underlying analyses, and some of the gross structure of the process, may at various times have been publicly revealed, enough of both the analyses and the detailed structural framework of the process have been held in confidence that this information, in this compiled form, continues to have great competitive value to GE. This value would be lost ifthe information as a whole, in the context and level of detail provided in the subject GE document, were to be disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources, including that required to determine the areas that are ~ affected by a power uprate and are therefore blind alleys, would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing its analytical process.

GBS-97-6-afTVApu l.doc AffidavitPage 4

STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

George B. Stramback, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this ~g, day of 1997.

G rge B. Stramback General Electric Company Subscribed and sworn before me this day of U'd~Z~ 1997. ~

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O ' NOTARY PUBLIGCALIFORNIA Q

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ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE Introduction This proposed amendment consists of a number of changes which will permit uprated power operation for Browns Ferry Units 2 and 3. The Browns Ferry units are GE, BWR/4's with Mark I Containments. Other plants similar in design to Browns Ferry have already received NRC approval to operate at uprated conditions and this evaluation and submittal follows these guidelines.

Both units were originally designed and the safety analyses performed for a maximum power level of approximately 3431 MWt, which corresponds to 105 percent of the rated steam flow. This power level is often referred to as "stretch power." This power level corresponds to approximately 104.2 percent of the current licensed rated power level (3293 MWt). Because of the significant economic advantages of operating at a higher power level, TVA is proposing a permanent amendment to the operating license for Browns Ferry Units 2 and 3, which will enable them to be operated at power 3.evels up to 105 percent of the current rated power, level (i.e., approximately 0.8 percent above the "stretch power" level).

The analyses and evaluations supporting these changes were completed using the guidelines in Topical Report NEDC-31897P-A, "Generic Guidelines For General Electric Boiling Water Reactor Power Uprate," (Reference 1). The NRC approved this Topical Report by letter dated September 30, 1991. Resolution of generic issues associated with power uprate was addressed in Topical Report NEDC-31984P, "Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," (Reference 2). The NRC approved this Topical Report by letter to dated July 31, 1992.

An increase in electrical output is accomplished primarily by generation and supply of higher steam flow to the turbine 0 . 97100703 ~ 4

generator. Continuing improvements in the analytical techniques (i.e., computer codes and data) based on several decades of BWR safety technology, plant performance feedback, and improved fuel and core design have resulted in a significant increase in the margin between calculated safety analysis results and the licensing limits. This available safety analysis margin, combined with the excess capability of as-designed equipment, systems, and components, provide the potential for an increase of five percent in the full power rating of a plant without the need to perform major Nuclear Steam Supply System or Balance-of-Plant hardware modifications. The full power level can be increased safely, and the installed systems and equipment are capable of performing their required functions at the uprated conditions.

The method for achieving higher power is to extend the power/flow map by increasing the core flow along the pre-uprate limiting flow control line. However, maximum recirculation flow will not exceed the pre-uprate limit.

The plant-specific safety analyses, to support this change are documented in Enclosure 5. This report demonstrates that Browns Ferry Units 2 and 3 can operate safely with a five percent increase in maximum reactor thermal power and an associated 30 psi increase in the operating reactor vessel pressure. This includes the corresponding increase in main turbine inlet steam flow and the corresponding increases in flow, temperature, pressure, and capacity in supporting systems and components.

Table El-1 summarizes the TS and Bases changes needed to support the power uprate. Each proposed change is discussed in this enclosure.

El-2

k TABLE E1-1 TECHNICAL SPECIFICATION AND BASES CHANGES FOR POWER UPRATE Proposed Location Change (page, item) Change Number

p. 1.1-5, RTP Revise value of Rated Thermal Power Definition (RTP) (from 3293 to 3458 MWt) to reflect increased power level.

2. p ~ 3.1-23'R Revise value of Standby Liquid 3.1.7.6 Control Syst: em (SLCS) pump discharge

p. B 3.1-44, pressure (from 1275 to 1325 psig) to SR 3.1.7.6 account for the effects of increased main steam relief valve setpoints.

3. p. 3. 3-6, Revise allowable value for Average Table 3.3.1.1- Power Range Monitor (APRM) Flow 1, Function Biased Simulated Thermal Power High 2.b Scram setpoint intercept (from 71 percent to 66 percent RTP) to maintain the pre-uprate power/flow relationship in terms of absolute power.

4. p. 3 3 7I ~ Revise allowable value for Reactor Table 3.3.1.1- Vessel Steam Dome Pressure High 1, Function 3 scram setpoint (from 1055 to 1090 psig) to account for increased '(30 psi) reactor operating pressure.

5. p. 3.3-34, Revise allowable value for Reactor SR 3.3.4.2.3b Steam Dome Pressure High Anticipated Transient Without Scram-Recirculation Pump Trip (ATWS-RPT) setpoint (from 1146.5 to 1175.0 psig) to account for increased (30 psi) reactor operating pressure.

6. p. 3.4-4, Revise lower bounds (rod lines) for Figure 3.4.1-1 Thermal Power vs. Core Flow Stability

p. B 3.4-3, Regions (from 108%, 100% and 80% to Applicable 102.9%, 95.2$ and 76.2% respectively)

Safety to maintain the pre-uprate power/flow Analyses ratios in terms of absolute power.

7. p. 3.4-7, Revise Main Steam Relief Valve (MSRV)

SR 3.4.3.1 setpoints (from 1105, 1115 and 1125 to 1135, 1145 and 1155 respectively) to account for increased (30 psi) reactor operating pressure.

E1-3

TABLE E1-1 (continued)

TECHNICAL SPECIFICATION AND BASES CHANGES FOR POWER UPRATE Proposed Location Change (page, item) Change Number

8. p. 3. 5-5, Revise upper and lower bounds of High SR 3.5.1.7 Pressure Coolant Injection (HPCI) and

p. B 3.5-12, Reactor Core Isolation Cooling (RCIC)

SR 3.5.1.7 pump test pressure (from 1010 and 920 psig to 1040 and 950 psig respectively) to account for

p. 3.5-13, increased (30 psi) reactor operating SR 3.5.3.3 pressure.

p. B 3.5-28, SR 3.5.3.3

9. p. 3 '-1g Add Limiting Condition for Operation LCO 3.7.1 and Action addressing Ultimate Heat Sink (UHS) for Residual Heat Removal

p. 3.7-3g Service Water (RHRSW). Add upper Condition D limit for the UHS average water SR 3.7.1.2 temperature for RHRSW to be in accordance with new Figure 3.7.1-1.

p. 3.7-3ag The value is used for calculating Figure 3.7.1-1 long-term suppression pool

p. 3.7-5, temperature response in the new SR 3.7.2.1 containment analysis. Add a note to

p. B 3.7-1, for the UHS average water temperature Background for Emergency Equipment Cooling Water

p. B 3.7-2, (EECW) to refer to additional UHS Applicable requirements from RHRSW.

Safety Analyses,

p. B 3.7-3, LCO, Applicability

p. B 3.7-5, Actions, SR

p. B 3.7-8, LCO

10. p. B 3.5-4, Revise value of upper design pressure Background for HPCI and RCIC operating range 54 psi (from 1120 to 1174 psig) to account for the effects of increased

p. B 3.5-24, main steam relief valve setpoints.

Background

E1-4

0 TABLE E1-1 (continued)

TECHNICAL SPECIFICATION AND BASES CHANGES FOR POWER UPRATE Proposed Location Change (page, item) Change Number

p. B 3.6-2, Revise value of P (from 49.6 to 50.3 Applicable psig) to account for results of the Safety new containment analysis.

Analyses

p. B 3. 6-7, Applicable Safety Analyses

12. p. B 3.7-2 Revise flow rate for RHRSW (from 4500 B 3.7.1 to 4000 GPM) used in peak containment 3.7-6, temperature analysis. Revise peak B containment pressure (from 49.6 to References 50.3 psig) to reflect new analysis results. Revise reference for peak suppression pool temperature.

El-5

Pro osed Chan es Table E1-1 summarized the TS and Bases changes needed to support the power uprate. These changes are also identified in Table ll-1 of Enclosure 5. Enclosure 3 contains the affected page list and copies of the appropriate Operating License and TS pages for Units 2 and 3 marked-up to show the proposed changes. Enclosure 4 contains the affected page list and copies of the appropriate Operating License and TS pages for Units 2 and 3 revised to show the proposed changes. Each of the operating license and TS change is evaluated below:

Technical Specifications

1. Rated Thermal Power is increased from 3293 Mwt to 3458 MWt on

~

page 3 of the Unit 2 and Unit 3 Operating Licenses, and in Section 1.1 (Definitions) of the Unit 2 and 3 TSs.

Evaluation This increase and redefinition of rated thermal power for Browns Ferry follows the generic guidelines of NEDC-31897P-A (Reference 1) for GE BWR power uprates. NEDC-31897P-A provides generic licensing criteria, clarified methodology, and a defined scope of analytical evaluations and equipment review to be performed to. demonstrate the ability to operate safely at the uprated power level. Technical Specification parameter values, which are expressed as a percentage of rated reactor thermal power or steam flow, were not changed because the uprated values were used in the bounding analyses and evaluations required by Reference 1 unless otherwise specified in this submittal. Enclosure 5 provides the results of the evaluations supporting the proposed uprated power operation consistent with the methodology presented in Reference 1. The report concludes that an uprated power rating of 3458 MWt can be achieved without a significant impact on equipment or safety analyses.

2. The surveillance test discharge pressure for the standby liquid control pump is increased from 1275 psig to 1325 psig.

This value appears in Surveillance Requirement (SR) 3.1.7.6 and the corresponding Bases Section B 3.1.7 in both the Unit 2 and Unit 3 TSs.

Evaluation As discussed in Section 6.5 of Enclosure 5, the surveillance test pressure is based on the maximum SLCS injection pressure and allowances for system test inaccuracies. Therefore, this pressure is increased from the pre-power uprate value of 1275 to 1325 psig to account for the increase in system injection pressure at power uprate conditions. Increasing the test

pressure by 50 psi assures the continued capability of these positive displacement pumps to deliver design rated flow at operating pressures expected at the uprated conditions. This change, therefore, maintains the original intent of SR 3.1.7.6.

The allowable value for the APRM Flow Biased Simulated Thermal Power High scram setpoint intercept is changed from 71 percent to 66 percent RTP, consistent with the revised analytical limit for power uprate. This value appears in Table 3.3.1.1-1, Function 2.b in both the Unit 2 and Unit 3 TSs.

Evaluation The flow-biased APRM simulated thermal power monitor analytical limit and scram setpoints are lowered proportionally to the increase in rated power, such that they remain substantially unchanged, in terms of absolute power and core flow. As discussed in Reference 1 and Section 5.1 of Enclosure 5, this proposed change ensures that the pre-uprate design margins and licensing basis are preserved for operation at the uprated power.

The allowable value for the Reactor Vessel Steam Dome Pressure High Scram setpoint is increased from 1055 psig to 1090 psig. The allowable value was increased 35 psi (as opposed to 30 psi) in order to place the instrument in the middle of its calibration range. The chosen allowable value is acceptable based on the analytical limit for the parameter. The Allowable Value appears in Section 3.3.1.1, Table 3.3.1.1-1, Function 3, in both the Unit 2 and Unit 3 TSs.

Evaluation The reactor vessel steam dome high, pressure scram limit is increased because the steam dome operating pressure is increased. Operating pressure for uprated power is increased to assure that satisfactory reactor pressure control is maintained. The operating pressure was chosen on the basis of steam line pressure drop characteristics and the steam flow capability of the turbine. Satisfactory reactor pressure control requires an adequate flow margin between the uprated operating condition and the steam flow capability of the turbine control valves at their maximum stroke. An operating dome pressure of 1035 psig, which is 30 psi higher than the current operating dome pressure, is expected.

Therefore, the high pressure scram is increased approximately by 35 psi to preserve existing margins to reactor scram.

The high pressure scram terminates a pressurization transient

not terminated by direct scram or high neutron flux scram.

The setting is maintained above the nominal reactor vessel operating pressure and below the specified analytical scram limit used in the safety analyses. The revised high pressure scram setpoint will preserve the hierarchy of pressure setpoints. This means that the high pressure scram setpoint will remain below the opening setpoint of the MSRVs. The MSRV nominal setpoints are also increased 30 psi, as discussed in proposed change 7. This hierarchy of setpoints provides assurance that there is a low probability of opening more than one MSRV without scram'ntervention.

The allowable value for the ATWS-RPT Reactor Steam Dome Pressure High setpoint is increased from 1146.5 psig to 1175.0 psig, a 28.5 psi increase. An increase of 30 psi would have yielded an allowable value of 1176.5 psig. Due to human factors consideration, the value of 1175 psig was chosen. The Allowable Value appears in Section 3.3.4.2, SR 3.3.4.2.3b, in both the Unit 2 and Unit 3 TSs.

Evaluation The ATWS-RPT high pressure setpoint initiates a trip of the recirculation pumps, thereby adding negative reactivity following events in which a scram does not (but should) occur. As discussed in Section 5.1.3.2 of Enclosure 5, the analytical limit for the ATWS-RPT high pressure setpoint was increased 30 psi in the power uprate ATWS safety evaluations to account for the 30 psi increase in vessel operating pressure, MSRV setpoints, etc. The analyses demonstrate that the ATWS criteria are met with the higher analytical limits.

Therefore, the allowable value is increased consistent with the analytical limit used in the safety analysis. Raising the ATWS-RPT high pressure setpoint to correlate with the increased operating pressure and analytical limit will tend to prevent unnecessary recirculation pump trips following pressurization transients with reactor scram (e.g., turbine trip or load rejection with bypass). Recirculation pump operation following a scram allows for better mixing of the reactor coolant and reduces thermal stratification in the vessel.

The rod lines defining lower boundaries for Thermal Power versus Core Flow Stability Regions I and II are reduced by the ratio of 1/1.05 (i.e., from 108%, 100%, and 80% rod lines for pre-uprate operation to the 102.9%, 95.2%, and 76.2% rod lines respectively for uprated power operation). This TS limit appears in Section 3.4.1 (Figure 3.4.1-1) and the corresponding Bases (B 3.4.1) of both the Unit 2 and 3 TSs.

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4 Evaluation During development of the generic power uprate program, GE and the NRC agreed to maintain the pre-uprate exclusion regions in the power-to-flow map related to thermal-hydraulic stability. The pre-uprate regions are defined by the 108%,

100%, and 80% rod lines. As noted previously, power uprate will redefine 100 percent rated power, and therefore rated rod or flow control lines. At uprated conditions, the 102.9%, 95.2%, and 76.2% rod lines are equivalent, on an absolute rather than percentage basis, to the pre-uprate rod lines.

The Main Steam Relief Valve lift setpoints will be increased 30 psi (from 1105, 1115 and 1125 psig to 1135, 1145 and 1155 psig, respectively). These values appear in SR 3.4.3.1 in both the Unit 2 and Unit 3 TSs.

Evaluation The MSRVs are designed to prevent overpressurization of the reactor pressure vessel during abnormal operational transients. The MSRV lift setpoints are increased to accommodate the increase in operating pressure that accompanies power uprate. The increase in MSRV setpoints ensures that adequate margins are maintained so that the increase in dome pressure during normal operation does not result in an increase in the number of unnecessary MSRV actuation's. The setpoint increase also maintains the hierarchy of pressure setpoints described in these proposed changes. Transient evaluations include a positive 3 percent tolerance to the nominal setpoints. As described in Section 3.2 of Enclosure 5, transient peak vessel pressure increases at uprated conditions, but remains below the 1375 psig American Society of Mechanical Engineers Code limit.

The adequacy of BWR MSRVs to operate at uprated temperatures and pressures has been evaluated generically in Section 4.6 of Reference 2. The reactor operating pressure and temperature increases of less than 40 psi and 5'F, respectively, used in that evaluation bound the uprated operating conditions.

The impact of power uprate on the containment dynamic loads due to MSRV discharge has also been evaluated. As discussed in Section 4.1.2 of Enclosure 5, the vent thrust loads with power uprate were calculated to be less than the loads used in the containment analysis. The effect of power uprate on MSRV air-clearing, the discharge line, the pool pressure boundary, and submerged structure drag loads is also discussed in Section 4.1.2 of Enclosure 5. That discussion concludes that the small increase in the setpoint pressure is

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well within the margin in the MSRV loads defined in the Mark I Torus Integrity Long-Term Program. Therefore, power uprate does not impact the MSRV load definitions used in the containment analysis.

The upper and lower bounds on reactor pressure, for purposes of performing HPCI and RCIC pump flow rate surveillance tests at high pressure, are increased 30 psi (i.e., from the pre-uprate values of 1010 and 920 psig to 1040 and 950 psig respectively). These values appear in SRs 3.5.1.7 for HPCI and 3.5.3.3 for RCIC in both the Unit 2 and Unit 3 TSs.

Evaluation The reactor operating pressure range for HPCI and RCIC surveillance tests at high pressure is increased to correspond with the increase in normal reactor operating pressure that accompanies power uprate. The change is needed to provide a more appropriate test range for the higher uprate reactor operating pressure. The requested changes will allow the quarterly demonstration of HPCI and RCIC capability to be performed at normal reactor operating pressures, which meets the original intent of the TSs. As discussed in Enclosure 5, Sections 3.8 and 4.2.1, the pre-uprate flow rates remain valid for uprated power conditions.

An upper limit for the UHS average water temperature for RHRSW is established. The value is used for calculating long-term suppression pool temperature response in the new containment analysis. This addition has been included in Section 3.7.1 and its associated bases of both the Unit 2 and 3 TSs. Also, added a Limiting Condition for Operation and Action addressing UHS for RHRSW and added a note to for the UHS average water temperature for EECW to refer to additional UHS requirements from RHRSW.

A note has been added to EECW SR 3.7.2.1 and the corresponding Bases (B 3.7.2) to xefer to additional UHS requirements based on RHRSW in both the Unit 2 and 3 TSs.

Evaluation The post-LOCA long term containment temperature response was evaluated for uprated power conditions as discussed in Section 4.1.1 of Enclosure 5. The evaluation was based on an RHRSW temperature of 92 to 95 degrees F for purposes of calculating long-term suppression pool temperature response.

Figure 3.7.1-1 provides the maximum allowable reactor thermal power as a function of UHS temperature. As indicated in the Bases for TSs 3.7.1 and 3.7.2, the RHRSW pumps and the EECW pumps share a common water source (i.e., the UHS).

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Therefore, the temperature of the RHRSW is dependent upon the UHS- temperature. Consequently, this change is necessary to maintain operation consistent: with the assumptions of the safety analysis for power uprate.

Bases Changes Several changes to the Unit 2 and Unit 3 Technical Specification Bases are proposed for consistency with the power uprate safety analyses. These proposed changes are in addition to the Bases changes corresponding to proposed changes 1 through 9.

10. The maximum operating pressure for the HPCI and RCIC systems in Bases Sections B 3.5.1 and B 3.5.3, respectively, is increased 54 psi for both units.

The Bases changes support the design of these high pressure systems to pump rated flow from approximately 150 psig up to a pressure associated with the analytical setpoint of the first group of MSRVs. This pressure conservatively considers the increase in the nominal setpoints, a 3 percent setpoint tolerance and a 5 psi margin. The capability of the HPCI and RCIC systems to deliver design flows at these pressures is discussed in Sections 3.8 and 4.2.1 of Enclosure 5.

Note that the upper design pressure for HPCI and RCIC is different from t: he maximum surveillance test pressure for HPCI and RCIC discussed in proposed change 8. The maximum surveillance test pressure corresponds to reactor operating pressure because the surveillance test: is performed when the unit is operating. The HPCI and RCIC upper operating pressure reflect:s the capability to inject water to the vessel following a reactor isolation.

11. The peak post accident containment pressure (P ) is changed to 50.3 psig. This value appears in Bases Sections B 3.6.1.1 and B 3.6.1.2 in both the Unit 2 and Unit 3 TSs.

Section 4.1.1.3 of Enclosure 5 discusses the peak short-term containment pressure response which was recalculated for power uprate conditions. As shown in Enclosure 5, Table 4-1, the calculated peak pressure for power uprate conditions is 50.3 psig. Therefore, the Technical Specification Bases Sect'ions 3.6-2 and 3.6-7 were changed to reflect the revised evaluation for power uprate.

12. The RHRSW flow assumed in the analysis is changed to 4000 GPM. This value appears in Basis section 3.7.1. Section 4.1.1.(a) of Enclosure 5 discusses the long-term

suppression pool response. The power uprate evaluation resulted in a peak bulk suppression pool temperature of less than 177 degrees F.

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References

1. NEDC-31897P-A, "Generic Guidelines For General Electric Boiling Water Reactor Power Uprate," May 1992.

2. NEDC-31984P, "Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," July 1991; and Supplements.

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4 ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384 ENVIRONMENTAL ASSESSMENT BROWNS FERY NUCLEAR PLANT POWER UPRATE PRO JECT (See attached)