ML051170244

(BFN) - Unit 1 - Response to Nrc'S Request for Additional Information Related to Technical Specifications (TS) Change No. TS-431 - Request for Extended Power Uprate Operation

ML051170244
Person / Time
Site:	Browns Ferry
Issue date:	04/25/2005
From:	Abney T Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NUREG-0800, TAC MC3812, TVA-BFN-TS-431
Download: ML051170244 (88)

Text

April 25, 2005 TVA-BFN-TS-431 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN, P1-35 Washington, D.C. 20555-0001 Gentlemen:

In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority )

BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - RESPONSE TO NRC's REQUEST FOR ADDITIONAL INFORMATION RELATED TO TECHNICAL SPECIFICATIONS (TS) CHANGE NO. TS-431- REQUEST FOR EXTENDED POWER UPRATE OPERATION (TAC NO. MC3812)

This letter contains the additional information requested by the NRC Staff concerning testing at EPU conditions. TS-431 which was submitted on June 28, 2004 (Reference 1), requested a license amendment and TS changes that support a requested increase in the reactor thermal power level to 3952 MWt, an approximate 20 percent increase in thermal power.

By letter dated February 23, 2005 (Reference 2), TVA supplemented the application, providing additional information requested by the NRC on Large Transient Testing.

In Enclosure 8 of Reference 1, and in Reply 4 of Reference 2, TVA provided justification for elimination of Large Transient Testing. However, based on February 10, 2005, and February 17, 2005, teleconferences between TVA management and NRC Staff, it was determined that the justification for the elimination of Large Transient Testing would be in accordance with NUREG-0800, Standard Review Plan (SRP), Section 14.2.1, Draft, Revision 0, Generic Guidelines for Extended Power Uprate Testing Programs.

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U.S. Nuclear Regulatory Commission Page 2 April 25, 2005 The Enclosure to this letter provides TVA's BFN Extended Power Uprate Testing Program and further justification for elimination of large transient testing. The enclosure replaces Enclosure 8 of the initial BFN Unit 1 application and the information in Reply 4 of the February 23, 2005, letter in their entirety. The information has been revised, expanded and in some cases reordered to fully address the review criteria contained in SRP Section 14.2.1. The basic content contained in Table 1 of this letter was provided in Enclosure 8 of Reference 1, but has been revised substantially, and renumbered for clarification. In the initial submittal, if a test was not planned specifically to address EPU implementation, the "Testing Planned for EPU" column was marked as "No." In Table 1 of this enclosure, where testing such as normal startup testing required by the BFN Technical Specifications would perform either a full or partial startup test, Table 1 contains a "Yes" in the column with an associated explanation. Therefore, some of the responses for the column entitled "Testing Planned for EPU", have been changed since our initial submittal.

Table 2 provides a comparison of the steady state and transient tests from the initial BFN startup tests to those described in SRP 14.2.1, Attachments 1 and 2. Table 3 tabulates the modifications, setpoint adjustments, and parameter changes required to implement EPU, and identifies the EPU tests associated with those changes, addressing the review criteria contained in SRP 14.2,Section III.B.

TVA is providing similar information regarding the Units 2 and 3 EPU application in a separate submittal. There are no new regulatory commitments associated with this submittal. If you have any questions concerning this letter, please telephone me at (256) 729-2636.

Pursuant to 28 U.S.G. § 1796 (1994), I declare under penalty of perjury that the forgoing is true and correct. Executed on this 25th day of April, 2005.

Sincerely, Original signed by:

T. E. Abney Manager of Licensing and Industry Affairs E-2

U.S. Nuclear Regulatory Commission Page 3 April 25, 2005

References:

1. TVA letter, T. E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Unit 1 -

Proposed Technical Specifications (TS) Change TS - 431- Request for License Amendment - Extended Power Uprate (EPU) Operation," dated June 28, 2004.

2. TVA letter, T. E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Unit 1 Response to NRCs Acceptance Review Letter and Request for Additional Information Related to Technical Specifications (TS) Change No. TS-418, Request for Extended Power Uprate Operation, dated February 23, 2005.

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U.S. Nuclear Regulatory Commission Page 4 April 25, 2005 Enclosure cc (Enclosure):

(Via NRC Electronic Distribution)

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Margaret Chernoff, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 E-4

U.S. Nuclear Regulatory Commission Page 5 April 25, 2005 MJB:BAB Enclosure cc (w/o Enclosure):

A. S. Bhatnagar, LP 6A-C J. C. Fornicola, LP 6A-C R. G. Jones, NAB 1A-BFN K. L. Krueger, POB 2C-BFN R. F. Marks, PAB 1C-BFN F. C. Mashburn, BR 4X-C N. M. Moon, LP 6A-C J. R. Rupert, NAB 1A-BFN K. W. Singer, LP 6A-C M. D. Skaggs, PAB 1E-BFN E. J. Vigluicci, ET 11A-K NSRB Support, LP 5M-C EDMS WT CA - K (with Enclosure) s:lic/submit/Techspec/BFN Units 1 EPU Testing Program TS 431.doc E-5

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 1 DOCKET NO. 50-259 RESPONSE TO NRC's REQUEST FOR ADDITIONAL INFORMATION RELATED TO TECHNICAL SPECIFICATIONS (TS) CHANGE NO. TS-431- REQUEST FOR EXTENDED POWER UPRATE OPERATION I. INTRODUCTION By letter dated June 28, 2004 (Reference 1), TVA submitted to the NRC a license amendment application requesting authorization for Extended Power Uprate (EPU) operation for Browns Ferry Nuclear Plant (BFN) Unit 1. By letter dated February 23, 2005 (Reference 2), TVA supplemented that application, providing additional information requested by the NRC as part of their acceptance review. In Enclosure 8 of the initial application, and in Reply 4 of the February 23, 2005 supplemental submittal, TVA provided justification for elimination of the requirement for large transient testing upon implementation of the EPU. The initial amendment application provided an EPU safety analysis report for EPU herein referred to as the PUSAR.

This submittal replaces Enclosure 8 of Reference 1 in its entirety. The information has been revised, expanded and in some cases reordered to fully address the review criteria contained in the NUREG-0800 (the Standard Review Plan or SRP), Section 14.2.1, Draft, Revision 0. The basic content contained in Tables 1 and 2 of this enclosure was provided in Enclosure 8 of Reference 1, but has been revised substantially, and renumbered for clarification. In our initial submittal, if a startup test was not planned specifically to address EPU implementation, the Testing Planned for EPU column was marked as No. In Table 1 of this enclosure, where testing such as normal startup testing required by the BFN Technical Specifications would perform either the full or partial initial startup test, Table 1 contains a Yes in the column with an associated explanation. Therefore, some of the responses for the column entitled Testing Planned for EPU have been changed since our initial submittal.

In accordance with SRP Section 14.2.1, this information demonstrates that structures, systems, and components (SSCs) will perform satisfactorily at the requested power level and thus the plant can be operated safely at the uprated power level.

II. BACKGROUND BFN Unit 1 was voluntarily shutdown in 1985 and is in the process of being recovered to an operating status. Unit 1 original and current licensed reactor thermal power is 3293 MWt with a steam dome pressure of 1005 psig. As part of the recovery process, TVA will uprate the unit to 3952 MWt with an associated 30 psi pressure increase. The scope of modifications required to recover the unit will also include the modifications E-1

required to achieve the equivalent of the 105% uprate completed on Units 2 and 3 as well as those for 120% uprate. Thus, Unit 1 will be recovered in a manner that will allow it to operate within the same design/licensing basis as BFN Units 2 and 3 following implementation of 120% EPU on those units. As part of the BFN Unit 1 restart project, and in addition to the acceptance testing planned for the modifications needed to implement EPU, BFN Unit 1 will undergo extensive testing prior to and during power ascension as part of its Restart Test Program (RTP). The RTP is a comprehensive program developed to confirm that the plant is capable of meeting its safe shutdown requirements. Further information concerning the BFN Unit 1 RTP is provided in Section III.A below.

BFN Units 2 and 3 have been licensed to operate at 105% of the original licensed thermal power (OLTP), 3458 MWt. Unit 3 has been operating at these uprated conditions since the fall of 1998 and Unit 2 since the spring of 1999. The power uprate to 105% for Units 2 and 3 included a 30 psig vessel steam dome pressure increase.

Because the steam dome pressure increase has already been accomplished, Units 2 and 3 will operate at the present reactor vessel pressure following implementation of EPU.

III. EPU TESTING PROGRAM REVIEW The following sections address the review criteria of Standard Review Plan (SRP),

Section 14.2.1, Subsection III, entitled Review Procedures. The below sections III.A through III.D correspond to Sections in the SRP.

A. Comparison of Proposed EPU Test Program to the Initial Plant Test Program

1. General Discussion A comparison of the proposed EPU testing program to the original startup test program performed during initial plant licensing is provided in Table 1 for all three units. The original BFN startup tests are included regardless of the power level at which they were performed.

The planned power ascension testing at BFN following uprate will provide a controlled and systematic testing program for the power levels above Current Licensed Thermal Power (CLTP) to EPU. Testing will be performed in accordance with the Technical Specifications (TS) and/or applicable procedures on instrumentation re-calibrated to EPU conditions.

Steady-state data will be taken during power ascension and continuing at each EPU power increase increment. This data will allow system performance parameters to be projected through the EPU power ascension.

EPU power increases above 100% CLTP are planned along an established flow control/rod line in increments of equal to or less than 5% power. Steady-state and transient data, including fuel thermal margin, will be taken at each step. Routine measurements of reactor and system parameters will be E-2

evaluated, prior to the next power increment. Plant procedures will identify specifically planned EPU tests, the associated acceptance criteria and the appropriate test conditions. All testing will be done in accordance with written procedures as required by 10 CFR 50, Appendix B, Criterion XI.

Table 1 also provides the resulting EPU Power Ascension Test Plan matrix.

The matrix specifies expected steady state and transient tests at different power levels. Similar matrices for the initial startup testing are contained in UFSAR Tables 13.5-4, 13.5-5, and 13.5-6.

In addition to the testing planned under the EPU Testing Program, BFN Unit 1 will perform a comprehensive Restart Test Program (RTP) as part of its restart program. The RTP is a component of the BFN Nuclear Performance Plan (NPP), which was submitted to NRC in 1986 to address TVA actions needed to restart all three BFN units. The NPP formed the basis for restart of Units 2 and 3 in 1991 and 1995 respectively. Likewise, the Unit 1 RTP forms part of the basis for the restart of BFN Unit 1. The BFN Unit 1 RTP is being developed based on the RTP developed, reviewed by the NRC, and approved for BFN Unit 3 as described and documented in NRC Safety Evaluation transmitted to TVA by letter dated August 30, 1994 (Reference 3).

Lessons learned during implementation of both the BFN Units 2 and 3 RTPs are being incorporated into the development of the BFN Unit 1 RTP. In a December 13, 2002 letter (Reference 4), TVA transmitted its proposed regulatory framework for the restart of BFN Unit 1, which reiterated the intent to follow the Unit 3 precedent with regard to the restart test program. The NRC accepted the approach in its August 14, 2003, letter to TVA (Reference 5).

The RTP for Units 1 and 3 was submitted to NRC on September 27, 1991, (Reference 6), and supplemented on February 18, 1992 (Reference 7),

December 28, 1992, (Reference 8) and July 19, 1993 (Reference 9). As a framework for its evaluation, the NRC reviewed the BFN RTP against the guidance contained in NRC Regulatory Guide 1.68, "Initial Test Programs for Water-Cooled Nuclear Power Plants", Revision 2. As part of that review, TVA described differences between the BFN RTP and the preoperational test guidance contained in NRC Regulatory Guide 1.68. As a result of that review, the NRC concluded that implementation of the Browns Ferry RTP would ensure proper verification of the functional integrity of the safety systems at BFN.

NRC approved the RTP for Unit 3 by SER and letter dated August 30, 1994 (Reference 3). TVA is following the RTP program for Unit 1 as previously submitted and as reiterated in its December 13, 2002, letter to NRC transmitting TVAs proposed regulatory framework for the restart of Unit 1 (Reference 4), and accepted by the NRC Staff by letter dated August 14, 2003 (Reference 5).

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A significant portion of the BFN Unit 1 RTP was derived from a comprehensive Safe Shutdown Analysis (SSA), which addresses all three units. The SSA is a systematic analysis of the requirements for ensuring the safe shutdown of BFN Units 1, 2, and 3 for transients, accidents, and special events. It documents the safety system actions for which credit has been taken in the UFSAR, reload analyses, and other communications concerning transients, accidents and special events. The events considered in the SSA included:

• Generator Trip

• Recirculation Flow Controller Failure

• Turbine-Generator Trip with Bypass Increasing Flow Failure

• Startup of Idle Reactor Recirculation

• Pressure Regulator Failure - Closed Pump

• Turbine Trip

• Loss of Shutdown Cooling

• Isolation of all Main Steam Lines

• Feedwater Controller Failure -

• Closure of One Main Steam Isolation Maximum Demand Valve

• Control Rod Drop Accident

• Loss of Condenser Vacuum

• Pipe Break Inside Containment -

• Loss of Feedwater Heater Large Break

• Shutdown Cooling (RHRS)

• Intermediate Pipe Break Inside Malfunction (Temperature Decrease) Containment

• Inadvertent Pump Start

• Small Pipe Break Inside Containment

• Control Rod Withdrawal Error

• Pipe Break Inside - Containment

• Fuel Assembly Insertion and Radiological Effect

• Control Rod Removal Error

• Fuel Handling Accident

• Pressure Regulator Failure - Open

• Pipe Break Outside Containment

• Inadvertent Opening of All Bypass

• Shutdown From Outside of Control Valves Room

• Inadvertent Opening of A

• Shutdown Without Control Rods Safety/Relief Valve

• Overpressure Protection (MSIV

• Loss of Feedwater Flow Closure - Backup Scram)

• Loss of Offsite AC Power

• Rotated or Mis-located Bundle

• Recirculation Control Failure-

• Flood Decrease

• Low Reservoir Downstream Dam

• Recirculation Pump Trip (One Pump Failure Trip)

• Tornado

• Recirculation Pump Trip (Two Pump Trip)

• Earthquake

• Recirculation Pump Seizure

• Fire

• Loss of Fuel Pool Cooling/Makeup E-4

From evaluation of the above events, the systems and associated system functions required to ensure safe shutdown were identified. These system functions or modes identified in the SSA were then systematically evaluated, and tests required to ensure capability of these functions identified. The required tests to support the SSA for each applicable system are identified in BFN Unit 1 Baseline Test Requirements Documents (BTRDs). Testing the modes identified in the SSA under the RTP at EPU conditions, as applicable, will verify the capacity of the systems to support their required safety functions.

2. Specific Acceptance Criteria BFN UFSAR Section 13.5.2.2 presents a general description of the initial startup testing that was performed for Unit 1 and Section 13.5.2.3 for Units 2 and 3. These UFSAR sections provide the objectives and acceptance criteria for the initial startup tests. Objectives and acceptance criteria as modified to reflect operation at 120% reactor thermal power will be used for planned EPU tests.

Table 1 consolidates the initial startup tests for all three units. The table includes the tests performed at a power level of equal to or greater than 80 percent of the original licensed thermal power level as well as the tests performed at power levels lower than 80 percent. The table provides a comparison of the initial tests to the planned EPU tests; in many cases, there is a direct correlation. Initial tests which are not planned to be repeated at EPU conditions are denoted by a No designation in the column entitled Testing Planned for EPU. As denoted in the table, some of the No answers are for tests which are not affected by operation at EPU conditions and thus no additional justification is provided beyond that in the table.

Others such as Full Main Steam Isolation Valve (MSIV) Closure and Turbine Stop/Control Valve Trips (referred to Generator Load Rejection in the ELTR) are high power tests for which the justification for not performing these tests are provided in Section III.C below.

Table 2 provides a comparison of the steady state and transient tests from the initial BFN startup tests to those described in SRP 14.2.1, Attachments 1 and 2. This table demonstrates that the applicable tests in Attachments 1 and 2 are addressed by the testing planned for BFN EPU implementation.

B. Post Modification Testing Requirements for Functions Important to Safety Impacted by EPU-Related Plant Modifications

1. General Discussion EPU for BFN will include the implementation of several plant modifications, mostly in the balance-of-plant systems, in addition to setpoint adjustments and operating parameter changes. Individual system or component E-5

performance characteristics affected by modifications are normally demonstrated by testing required by Technical Specifications and existing 10 CFR Part 50, Appendix B, quality assurance programs. Also, routine and EPU specific planned startup testing will demonstrate system and plant acceptability.

BFN has reviewed the aggregate impact of EPU plant modifications, setpoint adjustments, and parameter changes that could adversely impact the dynamic response of the plant to anticipated initiating events. The details of this review are provided below and verify the testing program adequately demonstrates that EPU related modifications will be adequately implemented.

The testing will demonstrate that functions important to safety that rely on the integrated operation of multiple systems, structures and components following an anticipated operational occurrence are capable of performing their design function prior to extended operation at the EPU power level.

2. Specific Acceptance Criteria BFN has identified: a) plant modifications, b) setpoint adjustments necessary to support operation at EPU conditions and c) changes in plant operating parameters resulting from operation at EPU conditions. These modifications, setpoint adjustments and parameters have been reviewed as described below to ensure that adequate testing will be provided.

Plant modifications and setpoint adjustments necessary to support EPU for BFN Unit 1 were listed in Enclosure 7 to the original license amendment submittal. These modifications enhance, upgrade, and/or replace existing plant components to allow for operation at EPU conditions. With the exception of allowing operation at EPU conditions, these modifications do not change the design functions of the equipment or the method of performing or controlling the function. These modifications do not involve first-of-a-kind plant modifications, the introduction of new system dependencies, or changes in types of system response to initiating events.

In accordance with SRP 14.2.1,Section III.B, the EPU plant modifications, setpoint adjustments and parameter changes have been evaluated to ensure that the testing program is adequate for those that meet all of the following criteria:

a. Impacts a function important to safety

b. Is required to mitigate a plant transient listed in Attachment 2 of SRP Section 14.2.1

c. Transient/accident response involves the integrated action of multiple systems, structure and components E-6

Table 3 tabulates the above information for the modifications, setpoint adjustments and parameter changes. The modifications have been re-ordered from that previously provided in Enclosure 7 of the original license amendment submittal to facilitate the discussion below. The modification to install safety related pressure actuation logic for the main steam relief valves has been removed from the table as it is not classified as an EPU required modification, and is not credited in the transient analyses. The modification to install five Local Power Range Monitors (LPRMs) is also removed, since this modification is not classified as an EPU required modification.

For the items evaluated in Table 3 which have a Yes determination for all three of the above criteria, the SRP 14.2.1,Section III.B requires an evaluation of the adequacy of the proposed EPU test program.

The items found to have a Yes answer for all three criteria are the feedwater/condensate pumps, the MSIVs and the Electro-Hydraulic Control (EHC) modification. An evaluation of the testing for these modifications is provided in Section III.C under the MSIV closure and turbine stop/control valve trip discussions.

The remaining items in Table 3 were evaluated and found to not meet the criteria of Section III.B.2. Further discussion of the testing associated with the items in Table 3 is provided below.

a. Modifications The high pressure main turbine steam path and turbine sealing steam system will be modified in order to accept the higher steam flows generated at the higher power level. In addition, the low pressure turbine steam path is being replaced. The steam path modifications will not change the normal turbine control operation and will not modify or impact the turbine stop and control valves or turbine bypass valves. Both the design function of the turbines and the system response to event initiators will not change as a result of these modifications. The turbine steam flow path is not required for response to plant transients and therefore no specific testing associated with the turbines is required to demonstrate plant transient response. Thus, these modifications have no impact on the integrated plant response during transient conditions.

The condensate, condensate booster, and reactor feedwater pump modifications are performed to upgrade the components to provide the higher flows for EPU operating conditions. These pump modifications do not change the design function of the condensate/feedwater system nor will a new system interaction be created.

The moisture separators, selected feedwater heaters, main condenser extraction steam bellows, condensate demineralizers, and steam packing E-7

exhauster bypass are being upgraded to accept the EPU operating conditions. These modifications have no impact on the integrated plant response during transient conditions.

The drywell building steel, torus attached piping supports and main steam supports are being upgraded to accept the EPU transient conditions. These changes maintain acceptable design margins for the piping and support configurations. These components are passive structural elements of the plant and will receive applicable structural installation testing. These modifications have no impact on the integrated plant response during transient conditions.

The steam dryer will be modified, as required, in order to accept the higher steam flows generated at the higher power level and accommodate the effect of flow induced vibrations. As discussed in the February 23, 2005, TVA letter, the steam dryer modifications will be based on the finalized and accepted loading methodology. The steam dryer is a passive component with no associated transient response testing.

The main steam relief valves will be reset based on the higher reactor pressure. The ability of these valves to perform their transient response function is demonstrated by bench testing coupled with the fact that the valves are individually remote manually opened at reactor pressure during unit startup.

Many of the Generic Letter 89-10 valves will be replaced while the remaining 89-10 valves will be refurbished. The transient response of these valve is demonstrated through their system testing coupled with MOVATS testing.

The recirculation pump motors have been re-rated and evaluated to ensure that the motors will accommodate the EPU conditions. Operation of the recirculation pumps is not relied upon for transient response. Re-rating of the pump motors does not involve physical changes to the motors or pumps and thus, does not change their transient response characteristics such as inertia or coast down rate. Therefore, there is no testing associated with transient response for this modification. Recirculation system jet pumps will have jet pump sensing line clamps installed as required to reduce vibration from the Recirculation Pump vane passing frequencies at EPU conditions. The jet pumps are passive components with no associated transient response testing. These changes do not affect the system functions and have no impact on the integrated plant response during transient conditions.

Main generator modifications include a stator and field rewind, installation of a redesigned core end assembly, additional monitoring instrumentation, correcting all known industry identified issues, and upgrading the temperature monitoring equipment. This will allow Unit 1 to re-rate the generator from E-8

1280 MVA to 1330 MVA. This has no effect on the generators transient performance. Modifications to the isolated phase bus duct cooling system are limited to addition of a fan to provide redundant cooling capacity. This has no effect on the isolated phase bus systems transient performance or its response to any plant transient. The BFN switchyard is being upgraded to a double breaker scheme. These modifications were considered in the latest Grid Stability study for BFN. These modifications do not change the design function of the affected equipment, nor will a new system interaction be created. These changes have no impact on the integrated plant response during transient conditions.

Integrated Computer System (ICS)/Safety Parameter Display System (SPDS) parameters are being revised by both Technical Specification and Balance of Plant (BOP) instrument upgrades as required. The ICS/SPDS does not perform any control function but is display only. Therefore, this change has no impact on the integrated plant response during transient conditions.

Reactor Recirculation vibration monitoring sensors have been installed on Unit 2 Reactor Recirculation piping and data has been collected at 105% of OLTP conditions. Additional vibration monitoring sensors will be installed as applicable to collect and analyze vibration data on selected Unit 1 systems at EPU conditions. These are monitoring only sensors and have no impact on the integrated plant response during transient conditions.

Modifications to the main steam isolation valves are being performed to reduce differential pressure across the MSIVs and increase stem size. MSIV closure times presently specified in the Technical Specifications are being maintained at EPU conditions and are utilized in transient analyses. Post modification testing and applicable Technical Specification testing will ensure that the MSIVs will continue to meet the current closure times used in transient analyses.

While not required or being implemented specifically to support EPU operations, several additional modifications involving control system upgrades are noteworthy. The following control systems are being modified to include digital control components: Reactor Recirculation, Reactor Feedwater and Turbine (EHC). These control systems upgrades to digital was not a requirement to implement EPU; however, these modifications will ensure, from a control standpoint, that BFN Unit 1 behaves and responds consistent with BFN Units 2 and 3. The control system settings will be installed with values that support operation at EPU conditions.

The design changes for Unit 1 control system upgrades were based on Unit 3 design changes that are currently in operation. The Unit 3 design is the same as Unit 2. The digital hardware utilized in the Unit 1 control systems will be essentially the same as the digital hardware installed in Unit 3. Minor E-9

differences in internal components are the result of the manufacturers providing the most current model of the hardware. These minor hardware differences will not affect the operation of the control systems since the software will be based on the Unit 3 software. Factory acceptance tests are performed to verify the operation of the hardware and software. TVA applies a standard process to each digital software program which includes software verification and validation testing. Post-modification testing ensures that the controllers properly communicate (both input and output) with their interfacing components. Therefore, since the Unit 1 digital control systems hardware and software were developed based on the Unit 3 equivalent, the Unit 1 control systems will respond to normal operation and operational transients in the same manner as the Unit 3 control systems. The scaling changes required for Units 2 and 3 EPU will be implemented at the corresponding Units 2 and 3 EPU implementation outage in order to maintain consistency with Unit 1.

The Recirculation System Motor-Generator Sets are being replaced with Variable Frequency Drives (VFDs) and the recirculation system speed controller upgraded to a digital control system. These upgrades were previously made for Units 2 and 3. The new recirculation system controls will be configured similarly to Units 2 and 3 for 105% reactor thermal power except the appropriate inputs will be rescaled to reflect EPU operation. The VFDs provide the electrical motive power input to the two recirculation pump motors which controls the speed of each of the recirculation pump motors.

Post modification testing of the speed controller and VFDs will verify the proper operation of the system to control the recirculation pump motor speeds. This post modification testing process has been used for the current operating configuration for Units 2 and 3 for both normal operation and event mitigation. The Recirculation System (including the VFDs) testing and tuning of the flow control system is performed each refueling startup during vessel hydro conditions and also at power conditions to analyze system response to speed demand of small and large changes. The recirculation system flow control capability has been proved by operation at current rated conditions for Units 2 and 3 to respond appropriately during normal operating conditions and large transient events such as turbine stop valve trips and turbine control valve trips. The analytical results for the limiting transients with the operating electrical characteristics of the upgraded recirculation flow control were provided in the PUSAR.

The Unit 1 Feedwater Control System and Reactor Feedpump Turbine Control System are being updated to a digital system; these upgrades were previously made for Units 2 and 3. The new Unit 1 digital systems will be configured similarly to Units 2 and 3 for 105% reactor thermal power except for the scaling changes required to implement EPU. As discussed in the section on operating experience for a turbine stop valve trip/turbine control valve trip, the control systems have been proven by operation at current rated E-10

conditions on Units 2 and 3 to respond appropriately to large transient events such as turbine stop valve trips and turbine control valve trips. Post modification testing of the control systems will verify the proper operation for EPU conditions. The post modification testing can be conducted by inserting simulated signals such as, for example, vessel level changes or flow demand changes. This process of using simulated signals versus conducting actual plant transients is an industry standard, and has been used for the current operating configuration. This method has proven to be effective in achieving appropriate system response for both normal operation and event mitigation.

These changes have no adverse impact on the integrated plant response during transient conditions since the control and trip circuitry (e.g., high vessel level, low pump suction pressure, etc.) will still function and prevent undesirable effects such as vessel overfill. The flow rates and response of the systems are modeled in the transient analyses. The EPU transient analyses have been performed utilizing the characteristics of the higher capacity system. Post modification testing will verify pump flow, head, and operating characteristics. EPU startup testing will perform testing of the condensate/feedwater control system, and calibration of the feedwater flow transmitters. This testing will verify the system parameters utilized in the transient analyses for the new pumps. System testing to verify the overall runout condition is not practical; however, planned post modification testing will confirm pump performance on an individual pump basis by a comparison of the design flow versus actual flow.

Feedwater/condensate pump performance will be confirmed on a system basis by comparison of data from startup testing to the calculated values.

The higher flow rates will assist in the recovery from certain transients such as trip of a single feedwater, condensate booster or condensate pump. The higher capacity pumps would not significantly affect the large transients (MSIV Closure, Turbine Stop Valve Trip, Turbine Control Valve Trip). Larger capacity pumps would potentially reduce the intensity of reactor vessel level decrease transients but generally have little effect in the short deviations in available flow and reactor water level.

The Unit 1 EHC system is being updated to a digital system; this upgrade was previously made for Units 2 and 3. The new Unit 1 digital system will be configured similarly to Units 2 and 3 for 105% reactor thermal power except for the scaling changes required to implement EPU. The scaling changes required for Units 2 and 3 EPU will be implemented at the corresponding Units 2 and 3 EPU implementation outage, which will ensure consistency with Unit 1. The EHC system controls the turbine and auxiliaries, including the turbine stop and control valves. During plant events requiring isolation of the steam lines to the main turbine, the EHC system provides the input to close the turbine stop valves. As discussed in the section on operating experience for a turbine stop valve trip/turbine control valve trip, the EHC software has been proven by operation at current rated conditions on Units 2 and 3 to E-11

respond appropriately to large transient events such as turbine stop valve trips and turbine control valve trips. Post modification testing of the EHC modifications will verify the proper operation of the EHC system, associated turbine controls and proper communication with the Reactor Protection System (RPS) for EPU conditions. The post modification testing can be conducted by inserting simulated signals such as low EHC pressure and stop valve position. This process of using simulated signals versus conducting actual plant transients has been used for the current operating configuration and has proven to be effective in achieving appropriate system response for both normal operation and event mitigation. The EHC modifications will have insignificant impact on transient system response.

The original Unit 1 Reactor Protection System (RPS) actuation instrumentation and Unit 1 Emergency Core Cooling System (ECCS) actuation instrumentation were replaced with analog trip systems as was previously performed for Units 2 and 3. The Units 2 and 3 analog trip systems design has been in operation for several years and has continued to provide reliable inputs to the actuation logic. The Unit 1 design provides physical and electrical separation as required to ensure high reliability and diversity while maintaining the original system actuation logic. Post modification testing validates the installation of the new instrument designs and ensures the systems will perform as required.

Based on the Unit 1 digital controls system software/hardware, and RPS and ECCS analog trip system design similarity with Units 2 and 3, the response of the Unit 1 instrumentation and controls to normal operation and transients should reflect the instrumentation and controls of Unit 3 response and thus the operational experience from Units 2 and 3 demonstrates that Unit 1 will respond properly to expected transients.

b. Setpoint Adjustments Technical Specification instruments have been reviewed for range/span and applicability at EPU conditions. Design changes have been prepared which will replace/re-span instruments as required to ensure proper instrument loop performance. The design modification process ensures appropriate testing for each loop when modified. The revised limits and setpoints are included in the transient analyses and have been evaluated as acceptable for EPU operation. Plant control systems for Feedwater, Recirculation, and the turbine will be upgraded to digital control systems as described in Section III.B.2.a above. The new systems will contain the required input for implementation of EPU so that the systems will be functionally identical to Units 2 and 3 after EPU implementation. Therefore, these changes will ensure proper transient system response.

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Balance of plant (BOP) instruments have been reviewed for range/span and applicability at EPU conditions. Design changes have been prepared which will replace/re-span instruments as required to ensure proper instrument loop performance. The design modification process ensures appropriate testing for each loop when required. This change has no impact on the integrated plant response during transient conditions.

c. Changes in Plant Operating Parameters For BFN EPU, the increase in electrical output is accomplished primarily by generation and supply of higher steam flow for the turbine generator. The high steam production is achieved by increasing the reactor power along slightly modified rod and core flow control lines. However, there is no increase in the maximum allowable recirculation flow value.

PUSAR Table 1-2 provides a summary of the reactor thermal-hydraulic parameters for the current rated and EPU conditions. Thermal power, reactor pressure, main steam flow, feedwater flow, and feedwater temperature are the only operating parameters with a significant change that have a potential impact on transient analyses.

As seen in the discussion of planned modifications above, many of the planned modifications are related to the change in the reactor pressure, main steam and feedwater operating parameters. Accordingly, these operating parameter changes are being appropriately incorporated into plant systems and are being evaluated and tested as part of those changes as described in Table 3. Additionally, the transient and accident analyses performed for EPU were evaluated utilizing the changed operating parameters for EPU conditions. The results of the limiting transients were provided in the PUSAR.

Parameters such as post accident containment pressure and torus temperature are not explicitly tested but are instead accounted for analytically in establishing the testing requirements.

Therefore, the changes in plant operating parameters resulting from operation at EPU conditions have been appropriately identified, evaluated, and incorporated in planned post modification testing and EPU startup testing.

C. Use of Evaluation to Justify Elimination of Power-Ascension Tests

1. General Discussion As discussed above in Sections III.A.1 and III.B.1, Tables 1, 2 and 3 define the testing that was accomplished for the respective unit startup and that planned for EPU related modifications. The tables also compare the BFN EPU testing to the testing proposed by SRP 14.2.1. In some cases, there are differences or exceptions in the BFN testing compared to the SRP proposed E-13

testing. The following issues require further evaluation and justification for not performing these tests in accordance with SRP Section III.C:

STP 25 Main Steam Isolation Valves STP 27 Turbine Stop and Control Valve Trips

2. Specific Acceptance Criteria The following provides an evaluation for each of the above listed tests against the specific acceptance criteria in SRP 14.2.1,Section III.C.2.

STP 25 Main Steam Isolation Valves This initial startup test required a simultaneous full closure of all MSIV's at about 100 percent of rated thermal power to demonstrate proper operation of the relief valves and the Reactor Core Isolation Cooling (RCIC). Since it is not intended to perform a similar test at EPU implementation, the following provides the evaluation of the acceptance criteria provided in SRP Section III.C.

The MSIV closure is an Anticipated Operational Transient as described in Chapter 14 of the BFN UFSAR. The MSIV closure is classified as an event that results in a sudden reduction of steam flow while the reactor is operating at power and, therefore, a significant nuclear system pressure increase. The closure of all MSIVs with direct scram failure (reactor scram on high neutron flux signal rather than the MSIV position switches) is the design basis event to analytically demonstrate compliance with the American Society of Mechanical Engineers (ASME) vessel overpressure protection criteria (upset condition).

The test proposed in the UFSAR by the initial startup test is the complete closure of all MSIVs with direct scram (scram on MSIV position switches).

The original MSIV closure tests for full isolation (closure of all eight valves simultaneously) are described in UFSAR Sections 13.5.2.2 as Test Number 25, and were intended to demonstrate the following:

1. The transient rise in simulated heat flux shall not exceed 10%,

2. The initial transient rise in vessel dome pressure occurring within 20 seconds of the main steam isolation valve trip initiation shall not exceed 150 psi,

3. MSIV closure time must be greater than 3 and less than 5 seconds,

4. Correct performance of the RCIC System,

5. Correct performance of the Main Steam Relief Valves.

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For Item 1, the intent was to monitor fuel performance. For this event, the closure of the MSIVs causes a vessel pressure increase and an associated increase in reactivity. The negative reactivity of the Reactor Protection System (RPS) scram from MSIV position switches offsets the positive reactivity of the pressure increase such that there is a minimal increase in heat flux. EPU will have minimal impact on the components important to achieving the desired thermal performance. RPS logic is unaffected and as discussed in PUSAR Section 2.5.1, overall control rod insertion times will not be significantly affected. MSIV closure speed is controlled by adjustments to the actuator and can be adjusted to occur within the required closure times.

The MSIV limit switches that provide the scram signal to the RPS are highly reliable devices that are suitable for all aspects of this application including environmental qualification requirements. There is no direct effect by any EPU changes on these switches or the RPS. Similarly there is no impact on the instrumentation actuated by reactor vessel pressure (other than rescaling). There may be an indirect impact caused by slightly higher ambient temperatures, but the increased temperatures will still be below the qualification temperature. These instruments are expected to be equally reliable before and after EPU. The RPS logic is unaffected by EPU, the MSIV closure time will not be affected by EPU, and the switches which initiate the scram and Recirculation Pump trip are highly reliable.

The increased post event pressure associated with the MSIV closure would not adversely impact the recirculation pump coastdown characteristics since the pressure acts on both sides of the pump. Consequently, the impact on transient heat flux results due to EPU is acceptable.

For Items 2 and 5, due to the minimal nature of the flux transient, the expected reactor pressure rise is largely dependent on Main Steam Relief Valve (MSRV) setpoints and initial reactor pressure. As discussed in the PUSAR, Section 3.1, the MSRV setpoints are increased consistent with the operating dome pressure increase. The uprated nominal MSRV setpoints ensure that adequate differences (simmer margin) between operating pressure and MSRV setpoints are maintained. Therefore, there is no effect on valve functionality (opening/closing).

As discussed below, in the section on operating experience for turbine stop valve trip/turbine control valve trip, the existing Units 2 and 3 MSRVs performance has proven acceptable during BFN transients. The most recent examples include:

• The Unit 2 May 15, 1999, Turbine Trip from 105% OLTP (LER 260/1999-003-01) where reactor pressure increased due to stop valve closure and resulted in five MSRVs opening initially. Pressure was then controlled by the turbine bypass system, and all systems responded as expected.

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• The Unit 2 July 25, 2001, Turbine Control Valve Fast Closure Scram from 105% OLTP (LER 260/2001-003-00) where eight MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

• The Unit 2 July 27, 2002, Generator Load Rejection from 105% OLTP (LER 260/2002-002-00) where four MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

• The July 8, 2004, Unit 2 main turbine trip from 105% OLTP (LER 260/2004-001-00) where seven MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

The MSRVs are tested each refueling outage in accordance with the Technical Specifications setpoint verification and through remote manual opening at reactor operating pressure. Performance of an actual MSIV closure test would provide no benefit for demonstrating the capability of the MSRVs for vessel overpressure protection that is not already accomplished by this component level testing.

For Item 3, the focus of the original testing was to verify that the steam flow from the reactor was not shut off faster than assumed (i.e., 3 seconds), since steam flow assists MSIV closure. During maintenance and surveillance, MSIV actuators are adjusted as necessary to control closure speed, and BFN test performance has been within the acceptance criteria. As discussed in Section III.B of this enclosure, the MSIVs are being modified to reduce the pressure drop across the valves. Procedurally required testing will ensure the valves continue to meet the Technical Specification required closure limits.

The BFN MSIVs were evaluated for EPU and determined acceptable for EPU operation. Industry experience, including BFN, has shown that there are no significant generic problems with this design.

For Item 4, a MSIV closure test is not required to demonstrate acceptable RCIC performance. During an MSIV closure event, the RCIC System would not initiate until the reactor vessel level decreased to the low low level (level

2) setpoint. This would not occur immediately since the level would decrease slowly due to relief valve actuations alone. This is not a rapid event in terms of RCIC response, the event does not produce any unique operating situations that have not been demonstrated previously by operation of the RCIC System, and there is no functional dependency between the closure of the MSIVs and the successful operation of the RCIC System. Thus, an MSIV closure test would not challenge the actuation logic beyond that exercised during periodic Technical Specification testing. Since there is a change to the E-16

normal reactor operating pressure and the MSRV setpoints change, the RCIC System performance parameters and associated setpoints will be rescaled accordingly as discussed in PUSAR Section 3.8. Operating history on Units 2 and 3 demonstrate that rescaling the system based on the increased pressure maintains full system functional capability over the expected pressure range. The rescaled system parameters will limit the maximum peak RCIC turbine speed. Additional confidence in this rescaling approach is provided by the operating experience on Units 2 and 3 for the rescaling on those units. As part of the power ascension testing for the 105% uprate on both Units 2 and 3, a RCIC simulated automatic cold quick start was performed when the reactor was within the uprated operating pressure window as defined by Technical Specifications. Since that time, RCIC has been routinely tested to assure that it can deliver rated flow and pressure and has initiated automatically during an operating event at the low low reactor vessel level as discussed below in the turbine stop valve trip/turbine control valve trip operating experience section. These items are adequate to show that RCIC can reliably deliver rated flow and perform its automatic function.

The following addresses the seven factors outlined in SRP 14.2.1,Section III.C.2.

a. Previous Operating Experience BFN Unplanned Power Uprate Related Transients Since implementation of the original 105% power uprate, BFN Units 2 and 3 have not experienced an MSIV closure event from near full power.

However, as discussed for the Turbine Stop and Control Valve Trip test (STP 27), several events have occurred which have successfully demonstrated operation of the RPS, control rod drives, MSRVs and RCIC.

Industry Unplanned Power Uprate Related Transients To date, thirteen other plants have implemented EPUs:

• Hatch Units 1 and 2 (from 105% to 113% of OLTP)

• Monticello (from 100% to 106.3% OLTP)

• Muehleberg (i.e., KKM) (from 105% to 116% OLTP)

• Leibstadt (i.e., KKL) (from 105% to 117% OLTP)

• Duane Arnold (from 105% to 120% OLTP)

• Brunswick Units 1 and 2 (from 105% to 120% OLTP)

• Quad Cities Units 1 and 2 (from 100% to 117% OLTP)

• Dresden Units 2 and 3 (from 100% to 117% OLTP)

• Clinton (from 100% to 120% OLTP)

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The following event involving an MSIV closure was experienced at the Hatch 2 plant. Hatch 2 experienced a reactor trip on high reactor pressure as a result of MSIV closure (from 113% OLTP (100% of uprated power)) in 2001. As noted in Hatch 2 LER 2001-003-00, systems functioned as expected and designed, given the conditions experienced during the event.

b. Introduction of New Thermal-Hydraulic Phenomena or Identified System Interactions As discussed earlier, no modifications are to be performed as part of EPU implementation that would cause BFN to behave differently from previous operating experience for an MSIV closure event. Related to the MSIV closure, EPU will have no impact on the components important to achieving the desired thermal performance. RPS logic is unaffected and as discussed in PUSAR Section 2.5.1, overall control rod insertion times will not be significantly affected. The event does not produce any unique operating situations that have not been demonstrated previously by operation of the RCIC system and there is no functional dependency between the closure of the MSIVs and the successful operation of the RCIC system. As discussed above, the MSIVs were evaluated for EPU and are acceptable for EPU operation. MSIV closure speed is controlled by adjustments to the actuator and is considered very reliable. As discussed in the PUSAR, Section 3.1 the MSRV setpoints are increased consistent with the operating dome pressure increase. The uprated nominal MSRV setpoints ensure that adequate differences (simmer margin) between operating pressure and MSRV setpoints are maintained. There is no effect on valve functionality (opening/closing).

Therefore, there are no new thermal-hydraulic or identified system interactions.

c. Facility Conformance to Limitations Associated With Analytical Analysis Methods The safety analyses performed for BFN used NRC-approved transient modeling codes. The NRC has accepted these for BWRs with a range of power levels and power densities that bound the requested power uprate for BFN. The codes have been benchmarked against BWR test data and have incorporated industry experience gained from previous transient modeling. Analyses use plant specific inputs and models all the essential physical phenomena for predicting integrated plant response to the analyzed transient. Thus, the codes will accurately and/or conservatively predict the integrated plant response to this transient at EPU power levels and no new information about transient modeling is expected to be gained from performing this large transient test.

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d. Plant Staff Familiarization With Facility Operation and Trial Use of Operating and Emergency Operating Procedures As discussed in Section 10.6 of the PUSAR, some additional training is required to enable plant operation at EPU conditions and power level.

For EPU conditions, the scope of operator responses to transient, accident and special events are not affected.

e. Margin Reduction in Safety Analysis Results for Anticipated Operational Occurrences The limiting pressurization transient events at EPU are the MSIV closure and turbine trip with turbine bypass failure. Both events are analyzed with failure of direct scram (i.e., scram based on neutron flux instead of valve position). BFN MSIV closure analyses assume that the events initiate at 102% of EPU reactor thermal power, the reactor dome pressure is 1055 psig (which is 20 psi higher than the nominal EPU dome pressure), one MSRV (with the lowest setpoint) is out-of-service (OOS) and uses a standard MSIV closure profile (percent area closure versus time) which has been shown by analysis results for the 105%

power level to increase the predicted peak pressure by approximately 14 psi compared to the BFN specific closure profile. The 20 psi increase in initial starting pressure coupled with the 14 psi from the closure profile results in the peak predicted pressure being conservatively increased approximately 34 psi between 105% and 120% power.

The results of the EPU overpressure protection analysis are given in Figure 3-1 of the PUSAR. The peak reactor vessel pressure is the only parameter for this transient that has a specific NRC acceptance limit.

The calculated peak reactor pressure vessel (RPV) pressure remains below the ASME limit, and the maximum calculated dome pressure remains below the Technical Specifications Safety Limit. However, the pressure increase does exceed the NRC definition of a minimum change in consequences of no more than a 10% reduction in margin.

Even though the less than 10% reduction in margin criterion would be exceeded, the performance of the actual test would not yield a substantial amount of confirmatory data. The actual test would be conducted in the following manner:

• Using a direct scram based on the MSIV position switches which reduces the resulting pressure rise by approximately 100 psi,

• Using the BFN specific MSIV closure profile which would reduce the reactor vessel pressure rise by 14 psi, E-19

• Using an initial reactor dome pressure of 1035 psig which would provide a 20 psi margin increase,

• All SRVs would be in service which would reduce the peak reactor vessel pressure by approximately 10 psi,

• The reactor thermal power would be limited to 100% of rated.

As discussed above, the performance of the RCIC system is not substantially challenged by this test. The system is tested by existing plant procedures that duplicate expected operational conditions during plant transients to demonstrate acceptable initiation and operation. No new information with regard to transient modeling or the analysis results are expected to be gained from performing this large transient test.

Therefore, TVA concludes that performance of a full MSIV closure transient test would not yield sufficient information to warrant conducting this test.

f. Guidance Contained in Vendor Topical Reports The EPU license application was prepared following the guidelines contained in the NRC approved General Electric (GE) Company Licensing Topical Reports NEDC 32424P-A (ELTR1), February 1999, and NEDC 32523P-A (ELTR2), February 2000, and its Supplement 1, Volumes I and II. Appendix L, Section L.2.4 of ELTR1 currently discusses the need to perform large transient testing, specifically a Main Steam Isolation Valve (MSIV) closure test and a Generator Load Rejection test. Therefore, as indicated for these tests in Table 1, a BFN plant-specific basis for exception to re-performing the original startup large transient tests is provided.

g. Risk Implications While this application is not "risk-informed," TVA believes, on a qualitative basis, that the benefits from performing these tests are negligible, when assessed against the risks of subjecting the plant to an otherwise unnecessary challenge. A scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. In addition, the risk posed by intentionally initiating a large transient event, although small, should not be incurred unnecessarily.

STP 27 Turbine Stop and Control Valve Trips During performance of STP 27, the turbine stop and control valves were tripped at selected reactor power levels (50% and 100%). Several reactor E-20

and turbine operating parameters were monitored to evaluate the response of the bypass valves, relief valves, and the RPS. The peak values and change rates of reactor steam pressure and heat flux were determined.

Since TVA does not intend to perform a similar test at EPU implementation, the following provides the evaluation of the acceptance criteria provided in SRP Section III.C.

The Turbine Stop Valve Trips and Turbine Control Valve Trip events are classified as an Anticipated Operational Transient as described in Chapter 14 of the BFN UFSAR. The events are classified as an event that results in a sudden reduction of steam flow while the reactor is operating at power and, therefore, a significant nuclear system pressure increase.

A turbine stop valve trip from high power conditions is the result of a turbine or reactor system malfunction which produces the following transient sequence:

a. Turbine Stop Valves fast closure (0.1 second closure time) which produces a fast steam flow shutoff,

b. Position switches on the stop valves sense the trip and initiate immediate reactor scram (for initial power levels above 30 percent),

c. The turbine bypass valves are opened simultaneously with turbine control valve closure, and reroute a portion of vessel steam flow to the condenser,

d. Reactor vessel pressure rises to the MSRV setpoints, causing them to open for a short period of time,

e. The steam passed by the MSRVs is discharged into the suppression pool, and

f. The turbine bypass valve (TBV) system controls nuclear system pressure after the MSRVs close.

A turbine control valve trip from high power conditions produces the following transient sequence:

a. Turbine-generator power/load unbalance circuitry and other generator trips initiate turbine control valve (TCV) fast closure (minimum response time of TCV fast closure: 0.15 seconds),

b. Turbine control valve fast closure is sensed by the reactor protection system, which initiates a scram and simultaneous recirculation pump trip (for initial power levels above 30 percent rated),

c. The turbine bypass valves are opened simultaneously with turbine control valve closure, and reroute a portion of the vessel steam flow to the condenser, E-21

d. Reactor vessel pressure rises to the MSRV setpoints, causing them to open for a short period of time,

e. The steam passed by the MSRVs is discharged into the suppression pool, and

f. The turbine bypass valve (TBV) system controls nuclear system pressure after the MSRVs close.

The original turbine stop/control valve closure tests are described in UFSAR Section 13.5.2.2 as Test Number 27, and were intended to demonstrate the following:

1. The transient rise in simulated heat flux shall not exceed 10 percent.

2. The initial transient rise in vessel dome pressure occurring within 10 seconds of the turbine/generator trip initiation shall not be greater than 150 psi.

3. Correct performance of the main steam relief valves.

4. Correct performance of the turbine bypass valves.

5. The turbine stop valves must begin to close before the control valves for the turbine trip. The turbine control valves must begin to close before the stop valves during the turbine control valve trip.

6. Following fast closure of the turbine stop and control valves, a reactor scram shall occur if the turbine first stage pressure is greater than the pressure corresponding to 30% power (26% following EPU implementation).

7. Feedwater systems must prevent flooding of the steamline following the transients.

8. The pressure regulator must prevent a low-pressure reactor isolation.

The feedwater controller must prevent a low-level initiation of the HPCI System and MSIV isolation as long as feedwater remains available.

For Item 1, the intent was to monitor fuel performance. For these events, the closure of the stop/control valves causes a vessel pressure increase and an increase in reactivity. The negative reactivity of the scram would offset the positive reactivity of the pressure increase such that there is a minimal increase in heat flux. EPU will have minimal impact on the components important to achieving the desired thermal performance.

The EHC system is being updated to a digital system; this upgrade was previously made for Units 2 and 3. The new Unit 1 digital system will be E-22

configured similarly to Units 2 and 3 for 105% reactor thermal power except for the scaling changes required to implement EPU. The scaling changes required for Units 2 and 3 EPU will be implemented at the corresponding Units 2 and 3 EPU implementation outage, which will ensure consistency with Unit 1. As discussed in the section on operating experience for a turbine stop valve trip/turbine control valve trip, the EHC software has been proven by operation at current rated conditions on Units 2 and 3 to respond appropriately to large transient events such as turbine stop valve trips and turbine control valve trips. Post modification testing of the EHC modifications will verify the proper operation of the EHC system, associated turbine controls and proper communication with the RPS for EPU conditions. The post modification testing can be conducted by inserting simulated signals such as low EHC pressure and stop valve position. This process of using simulated signals versus conducting actual plant transients has been used for the current operating configuration and has proven to be effective in achieving appropriate system response for both normal operation and event mitigation.

The EHC modifications will have insignificant impact on transient system response.

RPS logic is unaffected and as discussed in PUSAR Section 2.5.1, overall control rod insertion times will not be significantly affected. The switches that provide the scram signal are highly reliable devices that are suitable for all aspects of this application. There is no direct effect by any EPU changes on these switches. There may be an indirect impact caused by slightly higher ambient temperatures, but the increased temperatures will still be below the functional temperature range of the components. These switches are expected to be equally reliable before and after EPU.

The increased post event pressure associated with the turbine stop valve trip/turbine control valve trip would not adversely impact the recirculation pump coastdown characteristics since the pressure acts on both sides of the pump. Consequently, there is no effect on fuel thermal performance or heat flux results due to EPU.

For Items 2, 3 and 4, due to the minimal nature of the flux transient, the expected reactor pressure rise is largely dependent on MSRV and turbine bypass valve performance. As discussed in the PUSAR, Section 3.1 the MSRV setpoints are increased consistent with the operating dome pressure increase. The uprated nominal MSRV setpoints ensure that adequate differences (simmer margin) between operating pressure and MSRV setpoints are maintained. Therefore, there is no effect on valve functionality (opening/closing). The turbine bypass valves will be controlled by the digital EHC system with settings appropriate for EPU conditions.

As discussed in the below section on operating experience, the existing MSRVs performance as well as the performance of the turbine bypass valves E-23

has proven acceptable during BFN transients. The most recent examples include:

• The Unit 2 May 15, 1999 Turbine Trip from 105% OLTP (LER 260/1999-003-01) where reactor pressure increased due to stop valve closure and resulted in five MSRVs opening initially. Pressure was then controlled by the turbine bypass system, and all systems responded as expected.

• The Unit 2 July 25, 2001 Turbine Control Valve Fast Closure Scram from 105% OLTP (LER 260/2001-003-00) where eight MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

• The Unit 2 July 27, 2002 Generator Load Rejection from 105% OLTP (LER 260/2002-002-00) where four MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

• The July 8, 2004 Unit 2 main turbine trip from 105% OLTP (LER 260/2004-001-00) where seven MSRVs opened initially until pressure was controlled by the turbine bypass system, and all systems responded as expected.

The MSRVs are tested each refueling outage in accordance with the Technical Specifications setpoint verification and through remote manual opening at operating reactor pressure. The turbine bypass valves are exercised during startup to control reactor pressure which confirms their capability to perform under reactor operating pressure conditions.

Performance of an actual stop/control valve closure test would provide no benefit for demonstrating the capability of the MSRVs or turbine bypass valves for vessel overpressure protection that is not already accomplished by this component level testing.

For Item 5, during normal operation, the EHC emergency trip fluid system (ETS) is pressurized to open and control the main turbine steam valves.

When a turbine trip occurs, the ETS is depressurized removing EHC fluid from the turbine stop valves and relay trip valve. The relay trip valve then removes EHC fluid from the control valve allowing it to close. These physical attributes of the EHC system have not been modified and are not affected by pressure or flow changes associated with EPU.

The signal that initiates a turbine control valve trip is a power load imbalance.

This signal is used to anticipate a rapid acceleration of the turbine before a measurable increase in speed is detected. When this occurs, the control valve fast acting solenoids open removing EHC fluid from the control valves only. The control valves close to prevent turbine acceleration. This control E-24

function has not been modified since the original test and is not affected by pressure or flow changes associated with EPU.

For item 6, the intent was to demonstrate that a scram would occur based on stop/control valve position if the reactor power is above the low power setpoint. As demonstrated in the below section on Units 2 and 3 operating experience, a scram occurred in each event which confirms the functionality of the scram initiation on valve position. The conditions associated with EPU do not affect the capability of the electronic sensing system.

For items 7 and 8, the intent was to demonstrate that the feedwater control system and the pressure regulator were capable of controlling the post event pressure and reactor vessel water level such that no reactor vessel flooding, no low water level initiations of HPCI and no reactor vessel isolation occurred.

As demonstrated in the below section on operating experience, BFN Units 2 and 3 have experienced several events of this type with successful operation of the feedwater controller and pressure regulator. The controller adjustments to account for EPU conditions have been included in the transient analyses and demonstrated to provide proper control such that the acceptance criteria will be met. The modifications will have post-modification testing to confirm the proper operational characteristics. The conditions associated with EPU do not affect the capability of the systems to perform these functions.

The following addresses the seven factors outlined in SRP 14.2.1,Section III.C.2.

a. Previous Operating Experience BFN Unplanned Power Uprate Related Transients Since implementation of the original 105% power uprate, BFN Units 2 and 3 have previously experienced the following unplanned high power level large scale operating transients (100% power/automatic scram events):

• On May 15, 1999, Unit 2 received an automatic scram from 100 percent reactor power due to a turbine trip that occurred during routine turbine overspeed testing. The reactor scram caused reactor water level to go below the low level setpoint (level 3) which generated a redundant scram signal and initiated the PCIS as expected. The low reactor water level signal also initiated the SGTS and CREVS. The transient was mitigated by the automatic opening of nine Main Steam Bypass Valves and five Main Steam Relief Valves. All systems responded as expected and all control rods fully inserted. The cause of the turbine trip was failure of the mechanical E-25

trip cylinder to latch when hydraulically reset. This event was reported to the NRC in LER 1999-003-00, dated 6/14/99.

• On May 24, 2000, with Unit 3 operating at 100 percent power, an invalid low reactor water level scram signal was generated while returning a feedwater level transmitter to service following scheduled calibration. The reactor scram caused reactor water level to decrease below the low level (level 3) and low-low level (level 2) setpoints. All emergency systems operated as expected in response to the scram, including initiation of HPCI and RCIC and insertion of all control rods. This event did not result in the loss of the normal heat removal path since the condenser remained available throughout the event and was used for decay heat removal and no Main Steam Relief Valves opened. The scram was the result of a pressure perturbation in the common variable sensing line shared with both channels of reactor protection system level transmitters.

This event was reported to the NRC in LER 2000-001-00 dated 6/22/00.

• On July 25, 2001, Unit 2 received an automatic scram from 100 percent power due to a main turbine trip from a power-load unbalance that occurred during Combined Intermediate Valve Testing. The vendor software contained a numerical error that resulted in an inadvertent turbine trip. The reactor scram caused the water level to go below the low level setpoint (level 3) which generated an additional scram signal and initiated PCIS. Following the initial pressure transient which peaked at 1148 psig, eight Main Steam Relief Valves opened. All systems responded as expected and all control rods fully inserted. This event was reported to the NRC in LER 2001-003-00 dated 9/21/01.

• On July 27, 2002, a Unit 2 main generator trip, main turbine trip, and reactor scram occurred from 100% power. All expected system responses were received, including the automatic opening of four safety relief valves. Actuation of PCIS groups 2, 3, 6, and 8 occurred due to the expected temporary lowering of reactor water level. The normal heat rejection path for the reactor remained in service.

Reactor water level was recovered to the normal operating range by the normal reactor water level control system. All systems responded as expected and all control rods fully inserted. This event was reported to the NRC in LER 2002-002-00.

• On July 8, 2004, a Unit 2 main turbine trip/reactor scram occurred from 100% power. All expected system responses were received, including the automatic opening of seven safety relief valves.

Electrical switching was in progress at the time of the scram and during this switching activity, the Unit 2 UPS 120VAC Bus was E-26

inadvertently de-energized briefly. The reactor scram occurred at this time due to a turbine control valve fast closure/turbine trip condition. The loss of the UPS power would not itself be expected to result in a turbine trip/reactor scram because of the fault-tolerant design of the main turbine EHC system logic. However, it was determined that one of two main generator output current signal channels in the EHC logic has been automatically bypassed previously by the system software during a separate power supply transient on a different plant distribution bus. The subsequent temporary interruption of the UPS bus caused the loss of the second main generator output current signal channel and the system logic indicated that a power-load unbalance (i.e., main generator load reject) condition existed. This event was reported to the NRC in LER 2004-001-00.

• On November 23, 2004, while Unit 3 was in steady state operation at 100 percent power, a main turbine trip and subsequent reactor scram occurred. All expected system responses occurred. A lightning strike occurred on the TVA 500-kV system approximately 40 miles distant from Browns Ferry. This strike resulted in a phase-to-ground fault on all three phases of the transmission line and the electrical power transient caused speed perturbations on both the Unit 2 and Unit 3 main turbines. The rate of speed change seen on Unit 3 was slightly greater than the maximum rate anticipated by the turbine control system logic and therefore, the turbine speed feedback signals, while valid, were designated as invalid by the logic. With all turbine speed feedback signals designated as invalid, a main turbine trip on loss of speed feedback occurred in accordance with system design and a reactor scram occurred due to the turbine trip. All expected system responses occurred. No safety relief valve operation occurred during the trip transient and post-trip review confirmed that peak reactor pressures remained below the nominal SRV lift setpoints. This event was reported to the NRC in LER 2004-002-00.

• On February 11, 2005, the Unit 3 reactor scrammed from 100%

power. The scram was caused by a simultaneous false trip signal generated to the main generator circuit breaker 234, switchyard circuit breakers 5264 and 5268, and a main generator trip. This signal was generated when a PK block (disconnect device 26W),

which had been pulled as part of a clearance for breaker 5264, was re-inserted as part of a switching order from the Load Dispatcher for returning the breaker to service. When the PK block 26W was inserted (out of sequence of the switching order), the associated current transformer circuit was momentarily grounded resulting in a false differential. The correct sequence of the switching order was to E-27

actuate the trip cutout switches for the differential trip functions prior to inserting any of the PK blocks. The generator trip resulted in a turbine trip and opening of the output breakers caused a power-load unbalance trip. The control valve fast closure caused the reactor to scram. All rods inserted. Reactor water level lowered, as expected, and was recovered by normal feed water flow. All expected Primary Containment Isolation System isolations were received along with the auto start of Control Room Emergency Ventilation, and the three Standby Gas Treatment trains. This event was reported in LER 2005-001-00.

As reflected in the events discussed above, BFN has experienced unplanned pressurization transients at approximately 3458 megawatts thermal (MWt). No abnormalities or deviations from predicted behavior were observed and no significant anomalies were seen in BFNs response to these large scale transient events. Since the BFN Unit 1 uprate included the same reactor vessel steam dome pressure changes as previously implemented for Units 2 and 3, this experience is applicable to EPU.

Industry Unplanned Power Uprate Related Transients To date, thirteen other plants have implemented EPUs:

• Hatch Units 1 and 2 (from 105% to 113% of OLTP)

• Monticello (from 100% to 106.3% OLTP)

• Muehleberg (i.e., KKM) (from 105% to 116% OLTP)

• Leibstadt (i.e., KKL) (from 105% to 117% OLTP)

• Duane Arnold (from 105% to 120% OLTP)

• Brunswick Units 1 and 2 (from 105% to 120% OLTP)

• Quad Cities Units 1 and 2 (from 100% to 117% OLTP)

• Dresden Units 2 and 3 (from 100% to 117% OLTP)

• Clinton (from 100% to 120% OLTP)

Southern Nuclear Operating Companys (SNOC) application for EPU of Hatch Units 1 and 2 was granted. BFN and Hatch are both BWR/4 with Mark 1 containments. Hatch Unit 2 experienced an unplanned event that resulted in a generator load reject from approximately 111% OLTP (98% of uprated power) in the summer of 1999. As noted in SNOCs LER 1999-005-00, no anomalies were seen in the plants response to this event. In addition, Hatch Unit 1 has experienced two turbine trips from 112.6% and 113% of OLTP (99.7% and 100% of uprated power) as reported in LERs 2000-004-00 and 2001-002-00, respectively. Again, the behavior of the primary safety systems was as expected. No new plant behaviors for either plant were observed. This indicates that the E-28

analytical models being used are capable of modeling plant behavior at EPU conditions.

The KKL power uprate implementation program was performed during the period from 1995 to 2000. Power was raised in steps from its previous operating power level of 3138 MWt (i.e., 104.2% of OLTP) to 3515 MWt (i.e., 116.7% OLTP). Uprate testing was performed at 3327 MWt (i.e., 110.5% OLTP) in 1998, 3420 MWt (i.e., 113.5% OLTP) in 1999 and 3515 MWt in 2000.

KKL testing for major transients involved turbine trips at 110.5% OLTP and 113.5% OLTP and a generator load rejection test at 104.2% OLTP.

The KKL turbine and generator trip testing demonstrated the performance of equipment that was modified in preparation for the higher power levels. Equipment that was not modified performed as before. The reactor vessel pressure was controlled at the same operating point for all of the uprated power conditions. No unexpected performance was observed except in the fine-tuning of the turbine bypass opening that was done as the series of tests progressed. These large transient tests at KKL demonstrated the response of the equipment and the reactor response. The close correlation to the predicted response provides additional confidence that the uprate licensing analyses consistently reflected the behavior of the plant.

Progress Energys Brunswick Units 1 and 2 were licensed to 120% of OLTP. BFN and Brunswick are BWR/4 plants with Mark I containments.

Brunswick Unit 2 experienced an unplanned event that resulted in a generator/turbine trip due to loss of generator excitation from 115.2%

OLTP (96% of uprated thermal power) in the fall of 2003. As noted in Progress Energys LER 2003-004-00, no anomalies were experienced in the plants response to this event. No new plant behaviors were observed. This indicates that the analytical models being used are capable of modeling plant behavior at EPU conditions.

BFN and the Exelon Generation Companys Quad Cities Units 1 and 2 and Dresden Units 2 and 3 units are BWR/3 or BWR/4 plants with Mark 1 containments. Dresden 3 has experienced several turbine trips and a generator load rejection from high uprated power conditions. In January of 2004, Dresden 3 experienced two turbine trips from 112.3% and 113.5% of OLTP (96% and 97% of uprated power) as reported in LERs 2004-001-00 and 2004-002-00, respectively. The plant response was as expected and no new plant behaviors were observed. This indicates that the analytical models used for transient analyses are capable of modeling plant behavior at EPU conditions. In May 2004, Dresden 3 also experienced a loss of offsite power which resulted in a turbine trip on Generator Load Rejection from 117% of OLTP (100% of uprated E-29

power). Control rods fully inserted, and system and containment isolations occurred as expected. Manual initiations proceeded in accordance with procedures. Plant response indicates that the analytical models being used are capable of modeling plant behavior at EPU conditions; however, there were several failures (unrelated to the analysis models) involving the Standby Gas Treatment System and an Emergency Diesel Generator output breaker. This was reported in LER 2004-003-00.

Data collected from testing and responses to unplanned transients for Hatch Units 1 and 2, Brunswick 2, Dresden 2 and 3, and KKL plants during post-EPU operation have shown that plant response has consistently been as planned, within expected parameters, and bounded by the plant transient and safety analyses. Based on the similarity in design of these units to BFN, it is reasonable to conclude that the response seen at these units would be comparable to that which would be seen at BFN.

The EPU test program has been evaluated and determined to adequately demonstrate that the SSCs will perform satisfactorily in service. As discussed earlier, no modifications are to be performed as part of EPU implementation that would cause BFN to behave significantly different from previous operating experience. Transient experience for other operating BWR plants for a wide range of power levels has shown a close correlation of the plant transient data to the predicted response and response was bounded by the plant safety analyses. It can be concluded that large transients, either planned or unplanned, have confirmed predicted plant response to that of transient modeling.

b. Introduction of New Thermal-Hydraulic Phenomena or Identified System Interactions As discussed earlier, no modifications are to be performed as part of EPU implementation that would cause BFN to behave differently from previous operating experience for turbine stop valve trip/turbine control valve trip event. Related to the valve closure, EPU will have no impact on the components important to achieving the desired thermal performance. RPS logic is unaffected and as discussed in PUSAR Section 2.5.1, overall control rod insertion times will not be significantly affected. The event does not produce any unique operating situations that have not been demonstrated previously by operation of the MSRVs, turbine bypass valves, EHC, feedwater and feedwater control systems and there is no functional dependency between the closure of the turbine stop/control valves and the successful operation of these systems. As discussed in the PUSAR, Section 3.1 the MSRV setpoints are increased E-30

consistent with the operating dome pressure increase. The uprated nominal MSRV setpoints ensure that adequate differences (simmer margin) between operating pressure and MSRV setpoints are maintained. Therefore, there is no effect on valve functionality (opening/closing). Transient analyses have confirmed that the overall system responses and thermal-hydraulic phenomena are not affected by operation at EPU conditions. Therefore, there are no new thermal-hydraulic or identified system interactions.

c. Facility Conformance to Limitations Associated With Analytical Analysis Methods The safety analyses performed for BFN used NRC-approved transient modeling codes. The NRC has accepted these for BWRs with a range of power levels and power densities that bound the requested power uprate for BFN. The codes have been benchmarked against BWR test data and have incorporated industry experience gained from previous transient modeling. Analyses use plant specific inputs and models all the essential physical phenomena for predicting integrated plant response to the analyzed transient. Thus, the codes will accurately and/or conservatively predict the integrated plant response to this transient at EPU power levels and no new information about transient modeling is expected to be gained from performing this large transient test.

d. Plant Staff Familiarization With Facility Operation and Trial Use of Operating and Emergency Operating Procedures As discussed in Section 10.6 of the PUSAR, some additional training is required to enable plant operation at EPU conditions and power level.

For EPU conditions, the scope of operator responses to transient, accident and special events are not affected.

e. Margin Reduction in Safety Analysis Results for Anticipated Operational Occurrences The limiting pressurization transient events at EPU are the MSIV closure and turbine trip with turbine bypass failure. Both events are analyzed with failure of direct scram (i.e., scram based on neutron flux instead of valve position). The BFN turbine trip analyses assume that the events initiate at 102% of EPU reactor thermal power, the reactor dome pressure is 1055 psig, one MSRV (with the lowest setpoint) is out-of-service (OOS) and uses BFN turbine stop valve closure times. The results of the EPU overpressure protection analysis for the turbine trip event are given in Figure 3-2 of the PUSAR. The peak EPU pressure for this transient is 1298 psig versus a peak pressure of 1342 psig for MSIV closure. This analysis demonstrates that the MSIV closure transient E-31

described above remains the limiting event in terms of margin to reactor pressure vessel peak pressure.

No new information with regard to transient modeling or the analysis results are expected to be gained from performing this large transient test. Therefore TVA concludes that performance of a turbine stop/control valve closure transient test would not yield sufficient information to warrant conducting this test.

f. Guidance Contained in Vendor Topical Reports The EPU license application was prepared following the guidelines contained in the NRC approved General Electric (GE) Company Licensing Topical Reports NEDC 32424P-A (ELTR1), February 1999, and NEDC 32523P-A (ELTR2), February 2000, and its Supplement 1, Volumes I and II. Appendix L, Section L.2.4 of ELTR1 currently discusses the need to perform large transient testing, specifically a Main Steam Isolation Valve (MSIV) closure test and a Generator Load Rejection test. Therefore, as indicated for each initial startup test in Table 1, a BFN plant-specific basis for either performing or taking exception to re-performing the original startup large transient tests is provided.

g. Risk Implications While this application is not "risk-informed," TVA believes, on a qualitative basis, that the benefits from performing these tests are negligible, when assessed against the risks of subjecting the plant to an otherwise unnecessary challenge. A scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. In addition, the risk posed by intentionally initiating a large transient event, although small, should not be incurred unnecessarily.

D. Evaluate the Adequacy of Proposed Transient Testing Plans

1. General Discussion BFN will approach the power increase associated with EPU in a controlled manner using testing that will be used to demonstrate that the plant operates within design parameters.

2. Specific Acceptance Criteria BFN does not propose to utilize any new types of testing to demonstrate performance at EPU conditions. BFN also does not intend to intentionally initiate a large transient test at high power such as a main steam isolation valve closure or turbine trip. The testing that will be performed will be based E-32

upon standard plant procedures such as post modification testing, Technical Specification surveillance testing and required startup tests. The approach to EPU conditions from current licensed power level will be conducted in a stepped fashion with appropriate holds as shown in Table 1 for evaluation of test data. The proposed testing will ensure the plant parameters respond as expected to various system perturbations.

In the event a plant system does not respond as expected, the test will be put on hold and the plant maintained in a safe condition until the issue is resolved. This is a standard practice and requires no special controls related to EPU.

By conducting the power ascension in a controlled, step fashion, the time a plant system is in an untested configuration above the current licensed power level is minimized. The EPU-related testing is expected to be conducted in an expeditious manner during restart from a refueling outage as permitted by grid and/or environmental conditions. Thus emphasis will be placed on the timely completion of testing activities. The EPU ascension will include the required Technical Specification testing and Quality Assurance requirements.

IV. CONCLUSION TVAs evaluation has confirmed that the planned scope of EPU testing coupled with the BFN Unit 1 Restart Testing Program is adequate and additional large transient testing is neither required nor prudent. Steady state testing confirms the important nuclear characteristics required for transient analyses. Technical Specification required surveillance testing (e.g., component testing, trip logic system testing, simulated actuation testing) demonstrates that the systems, structures and components (SSCs) will perform their functions, including integrated performance for transient mitigation as assumed in the transient analyses. The characteristics and functions of SSCs do not need to be demonstrated further in a large transient test. In addition, the limiting transient analyses (i.e., those that affect core operating and safety limits) are re-performed each cycle and are included as part of the reload licensing analysis.

Large transient testing is normally performed on new plants because experience does not exist to confirm a plant's operation and its response to transients.

However, these initial tests are not normally performed for plant modifications following initial startup because of the well-established quality assurance and maintenance programs including component and system level post modification testing. Large transient tests only challenge a limited set of systems and components. This situation results because the plant would respond to the transient using appropriate safety and non-safety systems and thus the plant response would be much milder than the limiting transients analyzed in the UFSAR. Also, it would not be expected that a limiting single failure would occur randomly in conjunction with the test. This situation of only challenging a limited E-33

number of systems also results because the transient is rapidly mitigated and the long term consequences are relatively benign and controlled by normal operator action. Actuation of a safety system in the long term phases of such an event typically does not occur until the reactor system conditions have returned to near normal conditions. In this near normal condition, the actuation closely resembles that which would occur during normal surveillance testing. No new significant information would be gained by imposing a large transient as opposed to surveillance testing and thus large transient tests impose an undue strain on the plant systems without sufficient return of information.

Based on the (a) similarity of the BFN design configuration and system functions at pre-EPU to post-EPU; (b) results of industry EPU experience and responses to unplanned transients; (c) the fact that past transient and safety analyses results correlate closely to results from actual transients; and, (d) the evaluation of unplanned transients for the pre-EPU BFN and other post-EPU plants that provide favorable comparison of plant responses, it is reasonable and justifiable that the effects at EPU conditions can be determined by existing requirements of the Technical Specifications and plant procedures. For large transients, the effect at EPU conditions can be analytically determined on a plant specific basis versus actual transient testing. The transient analyses performed for the BFN EPU demonstrate that all safety criteria are met and that EPU does not cause any previous non-limiting events to become limiting. No safety related systems will be significantly modified for the EPU. Some instrument setpoints were changed but the setpoint changes themselves do not measurably contribute to the response to large transient events. The associated post modification testing will confirm proper response of the equipment.

As has been shown, analyses for EPU provide the necessary assurance that sufficient margins to safety limits are maintained. Should any future large transients occur, BFN procedures require verification that the actual plant response is in accordance with the predicted response. Plant event data recorders are capable of acquiring the necessary data to confirm the actual versus expected response. The important nuclear characteristics required for transient analysis are confirmed by the steady state physics testing. Transient mitigation capability is demonstrated by equipment surveillance tests required by the Technical Specifications. In addition, the limiting transient analyses are included as part of the reload licensing analysis.

A scram and isolation from high power is an undesirable transient cycle on the primary system. Because past testing at BFN and evaluation of pre-EPU operational experience at BFN and post-EPU experience at other plants has shown that large transient plant responses are within the bounds of plant transient analyses, additional large transient testing involving scram from high power is not justified or necessary.

E-34

V. REFERENCES

1. TVA letter, T. E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Unit 1 -

Proposed Technical Specifications (TS) Change TS - 431- Request for License Amendment - Extended Power Uprate (EPU) Operation," dated June 28, 2004.

2. TVA letter, TE Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Unit 1 Response to NRCs Acceptance Review Letter and Request for Additional Information Related to Technical Specifications (TS) Change No. TS-418, Request for Extended Power Uprate Operation, dated February 23, 2005.

3. NRC Letter to TVA, "Browns Ferry Nuclear Plant Units 1 and 3 - Restart Test Program," dated August 30, 1994.

4. TVA letter to NRC, Browns Ferry Nuclear Plant (BFN) - Unit 1 Regulatory Framework for the Restart of Unit 1," dated December 13, 2002.

5. NRC Letter to TVA, "Regulatory Framework for the Restart of Browns Ferry Nuclear Plant Unit 1," dated August 14, 2003.

6. TVA letter to NRC, "Browns Ferry Nuclear Plant (BFN) - Restart Test Program (RTP) for Units 1 and 3," dated September 27, 1991.

7. TVA letter to NRC, "Browns Ferry Nuclear Plant (BFN) - Request for Additional Information Regarding the Restart Test Program for Units 1 and 3," dated February 18, 1992.

8. TVA letter to NRC, "Browns Ferry Nuclear Plant (BFN) - Update of Restart Test Program (RTP) for Units 1 and 3," dated December 28, 1992.

9. TVA letter to NRC dated July 19, 1993, "Browns Ferry Nuclear Plant (BFN) - Restart Test Program (RTP) Update for Units 1 and 3."

E-35

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER STP 1 Chemical and Radiochemical: Y Y Y Open Vessel Yes The purpose of the original test included X X X X X X Testing, (Standard (a) securing information and knowledge Before fuel loading, a complete Initial plant about the quality of the reactor coolant set of chemical and Heatup, procedure) chemistry, (b) determination that the radiochemical samples were Power sampling equipment, procedures, and taken to ensure that all sample Testing analytic techniques are adequate to stations were functioning supply the data required to demonstrate properly and to determine that the coolant chemistry met water initial water quality. quality specification and process Subsequent to fuel loading requirements, and (c) evaluation of fuel during reactor heatup and at performance, operation of demineralizers major power level changes, and filters, operation of the offgas system samples were taken and and calibration of certain process measurements made to instruments.

determine the chemical and radiochemical quality of reactor For test purpose (a), EPU testing will water and reactor feedwater, include sampling and measurements at amount of radiolytic gas in the selected power levels to determine 1) the steam, gaseous activities chemical and radiochemical quality of leaving the air ejectors, decay reactor water and feedwater and 2) times in the offgas lines, and gaseous release.

performance of filters and demineralizers. Calibrations For test purposes (b) and (c), the current were made of monitors in the Browns Ferry chemistry and plant stack, liquid waste system, and performance monitoring programs gather liquid process lines. information on plant equipment and system performance. This information is evaluated in order to maintain equipment, system and plant performance within process requirements, chemistry/radiochemistry specifications and guidelines and fuel warranty. The demonstration of the sampling equipment, procedures, analytic techniques, and operation validation E-36

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER need not be repeated. Therefore, this purpose of the original testing is not applicable to EPU and is not required.

STP 2 Radiation Measurements: Y Y Y Open Vessel Yes (EPU The purpose of the original test was (a) to X X X X X X Testing, startup test) determine the background radiation A survey of natural background Initial levels in the plant environs prior to radiation throughout the plant Heatup, operation for base data on activity site was made before fuel Power buildup and (b) to monitor radiation at loading. Subsequent to fuel Testing selected power levels to assure the loading, during reactor heatup protection of personnel during plant and at power levels of 25, 50, operation.

and 100 percent of rated power, gamma radiation level For test purpose (a), the demonstration of measurements and, where background radiation levels need not be appropriate, thermal and fast repeated. Therefore, this purpose of the neutron dose rate original testing is not applicable to EPU measurements were made at and is not required.

significant locations throughout the plant. All potentially high For test purpose (b), EPU testing at radiation areas were surveyed. selected EPU power levels will take gamma dose measurements and where appropriate, neutron dose measurements at specific limiting locations throughout the plant to assess the impact of the uprate on actual plant area dose rates.

STP 3 Fuel Loading: Y Y Y Open Vessel Yes The purpose of the original test was to X Testing (Standard load fuel safely and efficiently to the full Before fuel loading, control plant core size.

rods were installed and tested. procedure)

A neutron source of Current Technical Specifications and approximately 10 neutrons per approved plant procedures will effectively sec was installed near the govern the safe and efficient loading of center of the core. At least fuel for EPU implementation. The normal E-37

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER three neutron detectors refueling test program for open vessel calibrated and connected to testing will accomplish required testing high flux scram trips were and verification. During refueling, each located to produce acceptable fuel movement is verified to a pre-signals during loading. Fuel determined fuel assembly transfer form Loading was begun at the and assembly movement and subcritical center of the core and multiplication is monitored. Upon proceeded radially to the fully completion of core loading, a final configuration. As each cell verification of fuel assembly loading and was loaded, the following orientation is completed. During checks were performed: 1) refueling, subcriticality checks are Subcriticality check, 2) Control performed. Control rod function Rod Function test, 3) Fuel (withdrawal and insertion) is completed Loading check, 4) Repeat of with the core fully loaded. These Subcriticality check, and 5 procedures and actions meet the intent of Repeat of the Control Rod the original testing.

Function test. Shutdown margin demonstrations were performed periodically during fuel loading.

STP 4 Full Core Shutdown Margin: Y Y Y Open Vessel Yes The purpose of the original testing was to X (Standard demonstrate that the reactor would be This test was performed in the plant subcritical throughout the first fuel cycle fully loaded core at ambient procedure) with any single control rod withdrawn.

temperature in the xenon-free The core shutdown margin requirement is condition. Subcriticality was not changed by EPU. Shutdown margin demonstrated with the testing is performed during each refueling strongest rod fully withdrawn before any fuel handling over an open and a series of calibrated rods reactor vessel and the normal refueling pulled to a position calculated test program for heatup will accomplish to be equal to a shutdown required shutdown margin testing for margin specified to account for EPU.

expected reactivity changes These procedures and actions meet the E-38

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER during core lifetime. purpose of the original testing.

STP 5 Control Rod Drive System: Y Y Y Open Vessel Yes The purpose of the original test was to (a) X Testing, (Standard demonstrate that the CRD system The CRD tests performed Initial plant operated properly over the full range of verified that all control rod Heatup, procedure) primary coolant temperature and drives operated properly when Power pressures from ambient to operating and installed and also periodically Testing particularly that thermal expansion of during heatup to assure that core components does not bind or there is no significant binding significantly slow control rod movements, caused by thermal expansion and (b) to determine the initial operating of the core components. characteristics of the entire CRD system.

For test purpose (a), as stated in Section 2.5.1 of the EPU safety analysis report, no change is made to the control rods due to the EPU and the scram times are decreased by the transient pressure response. The normal refueling test program will accomplish control rod drive system testing and troubleshooting and CRDM scram insertion timing for EPU implementation.

For test purpose (b), as stated in Section 2.5.2 of the EPU safety analysis report, the CRD positioning and cooling functions are not affected by EPU.

Confirmation that the system meets TS requirements for operability for startup and power ascension is required by the refueling test program. As the EPU does not have an effect on the CRD System, the current verification per the refueling test program is valid for EPU operations.

E-39

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER STP 6 SRM Performance and Control Y Y Y Open Vessel Yes The purpose of the original testing was to NA Rod Sequence: Testing, (Standard demonstrate that the operational sources, Initial plant SRM instrumentation, and rod withdrawal The operational neutron Heatup, procedure) sequences provide adequate information sources were installed and Power to achieve criticality and increase power source range monitor count Testing in a safe and efficient manner. It also rate data taken during rod determined the effect of typical rod withdrawals to critical and movements on reactor power.

compared with stated criteria on signal and signal count-to- This was an initial startup test noise count ratio. requirement with regard to startup neutron sources to achieve initial Control Rod patterns were criticality in a safe and efficient manner recorded periodically as the for the control rod withdrawal sequences.

reactor was heated to rated Operation at EPU increases the upper temperature. As each rod end of the power operating domain.

group was completed during These changes in the higher end do not the power ascension, the significantly or directly affect the manner electrical power, steam flow, of operating or response of the reactor in and APRM response were the startup/low power range. The recorded. neutron monitoring is confirmed to be Movement of rods in a operating properly at low power via prescribed sequence was standard plant procedures/processes monitored by the Rod Worth The rod worth minimizer controls rod Minimizer and Rod Sequence patterns to ensure compliance with the Control System to prevent prescribed rod patterns. Therefore, unacceptable our-of-sequence additional SRM performance testing is control rod movements during not required for EPU startup or shutdown.

STP 9 Water Level Measurement: N Y Y Initial Yes The purpose of the original testing was to X X X X X X (Units Heatup, (Standard a) verify the calibration and b) verify the 2/3 The first part of testing Power plant agreement of the various narrow and only). measured the YARWAY wide range level indicators under various E-40

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER See reference to verify agreement Testing procedure) conditions.

STP 39 with the temperature correction for Unit factor used in calibration. The As part of the restart efforts, the Yarway

1) second part of the test temperature equalizing columns are consisted of determining the removed and the instruments replaced agreement of the water level with analog trip system devices. The instrumentation at two core hardware associated with this new flow rates and various heights. reactor water level instrumentation is not modified and normal operational water level and level setpoints (alarms/trips/

actuations) are not changed by EPU.

The demonstration of procedures and operational validation for EPU therefore need not be repeated.

STP 10 IRM Calibration/Performance: Y Y Y Open Vessel Yes (EPU The purpose of the original testing was to (Not a startup test, but will be done during first Testing, startup test) adjust the IRM system to obtain an controlled shutdown following APRM Initially, the IRM system was Initial optimum overlap with the SRM and calibration for EPU) set to maximum gain. After the Heatup, APRM systems.

APRM heatup calibration and Power after the first heat balance Testing The IRM overlap with the SRMs is not calibration of the APRM's, the affected by EPU. The APRMs will be re-IRM-APRM, overlap was referenced to read 100% at EPU checked and the IRM gains conditions. After the APRM calibration adjusted if necessary to for EPU, the IRM gains will be adjusted improve the IRM system as necessary to assure the IRM overlap overlap between the SRMs with the APRMs. This will be performed and IRMs. during the first controlled shutdown following APRM calibration for EPU.

STP 11 LPRM Calibration: Y Y Y Power Yes The purpose of the original testing was to X Testing (Standard calibrate the LPRM system.

The LPRM channels were plant calibrated to make the LPRM procedure) LPRM calibration is performed readings proportional to the periodically as specified in the Technical E-41

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER average heat flux in the four Specifications using approved plant corner fuel rods surrounding procedures. The normal refueling test each chamber at the chamber program will accomplish APRM elevation. The initial calibration calibration during startup to EPU in factors were obtained from accordance with the refueling test measurements of axial power procedure. The method and approach distribution, precalculated local used to perform LPRM calibration is not power distributions, and affected by EPU. These procedures and precalculated radial power actions will meet the intent of the original distributions. testing.

STP 12 APRM Calibration: Y Y Y Power Yes The purpose of the original testing was to X Testing (Standard calibrate the APRMs.

The APRM's were initially plant adjusted to maximum amplifier procedure) APRMs will be calibrated in accordance gain. A low power calibration with Technical Specifications during was performed based on heat startup from each refueling outage by the balance calculations during normal refueling test program. For EPU, reactor heatup. When reactor each APRM channel will be adjusted to power was high enough to be consistent with the core thermal obtain more accurate steady power, referenced to the EPU level as state heat balances to determined from the heat balance.

determine actual core thermal These procedures and actions will meet power, the APRM channels the intent of the original testing.

were calibrated to read percent of core thermal power.

STP 13 Process Computer: Y Y Y Prior to Yes The purpose of the original testing was to X X X X X X Startup, (Standard verify the performance of the process Following fuel loading, during Open Vessel plant computer under plant operating plant heatup, and the Testing, procedure) conditions.

ascension to rated power, the Initial nuclear steam supply system Heatup, Plant process computer data installation E-42

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER and the balance of plant Power and verification is confirmed during system process variables Testing startup from each refueling outage. The sensed by the computer normal refueling test program will became available. As station accomplish this testing during startup and process variable signals power ascension at EPU. This testing became available to the will meet the intent of the original testing.

computer, verification was made that the computer was receiving correct values of sensed process variables and that the results of performance calculations of the nuclear steam supply system and the balance-of-plant were correct.

STP 14 RCIC System: Y Y Y Initial Yes The purpose of the original testing was to X Heatup, (Standard verify the proper operation of the RCIC Flow tests of the RCIC System Power plant system over its required operating were performed at reactor Testing procedure) pressure range.

pressures between 150 and 1,020 psig. These tests were Unit 2/3 - As discussed in EPU safety designed to verify proper analysis report Sections 3.8 and 9.1.3 operation of the RCIC System, there is no change to the normal reactor determine time to reach rated operating pressure and the MSRV flow and adjust flow controller setpoints remain the same compared to in RCIC System for proper flow those associate with the 105% power rate. These tests were first uprate and the associated 30 psi performed with the system in pressure increase. For operation at EPU, the test mode so that there is no change to the maximum discharge was not routed to specified reactor pressure for RCIC the reactor vessel and the final System operation, no changes to the demonstration routed RCIC System performance parameters, discharge flow to the reactor and no effect on the maximum reactor vessel while the reactor was at pressure postulated to be present during system startup. Current testing is E-43

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER partial power. performed by testing with suction from the condensate storage tank with discharge back to the condensate storage tank. Therefore, no changes are required to meet the performance requirements for the RCIC System or to limit the maximum startup transient speed peak. RCIC System testing, including automatic starts from cold conditions is governed by Technical Specifications and approved plant procedures. During unit startup, the 150 psig test must be completed and RCIC declared operable. As EPU does not have an effect on the RCIC System, the current testing per the unit startup procedures is valid for EPU operations.

Therefore, additional specific system testing at EPU conditions is not required.

Unit 1 - During unit startup, RCIC reliability will be demonstrated by cold quick starts with the pump aligned in the normal EPU operating reactor pressure range. After the auto start portion of the test while the system is in operation, small step disturbances in flow command will be input to demonstrate satisfactory turbine control stability. Testing will be performed by testing with suction from the condensate storage tank with discharge to the reactor vessel.

Additional specific system testing at EPU conditions is not required.

E-44

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER STP 15 HPCI System: Y Y Y Initial Yes The purpose of the original testing was to X Heatup, (Standard verify the proper operation of the HPCI Flow tests of the HPCI System Power plant system over its expected operating were performed at reactor Testing procedure) pressure range.

pressures between 150 and 1,020 psig. These tests were Unit 2/3 - As discussed in the EPU safety designed to verify proper analysis report Sections 4.2.1 and 4.3, operation of the HPCI system, there is no change to the maximum determine time to reach rated specified reactor pressure for HPCI flow, and adjust the flow system operation, no changes to the controller in HPCI system for HPCI system performance parameters, proper flow rate. These tests and no effect on the maximum reactor were first performed with the pressure postulated to be present during system in the test mode so that system startup compared to those discharge was not routed to associate with the 105% power uprate the reactor vessel and the final and the associated 30 psi pressure demonstration routed increase. For operation at EPU, no discharge flow to the reactor changes are required to meet the vessel while the reactor was at performance requirements for the HPCI partial power. system or to limit the maximum startup transient speed peak. HPCI System testing, including automatic starts from cold conditions is governed by Technical Specifications and approved plant surveillance procedures. During unit startup, the 150 psig test must be completed and HPCI declared operable.

As EPU does not have an effect on the HPCI System, the current testing per the unit startup procedures is valid for EPU operations. Current testing is performed by testing with suction from the condensate storage tank with discharge back to the condensate storage tank..

Therefore, additional specific system E-45

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER testing at EPU conditions is not required.

Unit 1 - During unit startup, HPCI reliability will be demonstrated by cold quick starts with the pump aligned in the normal EPU operating reactor pressure range. After the auto start portion of the test while the system is in operation, small step disturbances in flow command will be input to demonstrate satisfactory turbine control stability. Testing will be performed by testing with suction from the condensate storage tank with discharge to the reactor vessel.

Additional specific system testing at EPU conditions is not required.

STP 16 Selected Process Y Y Y Power No (Low The purpose of the original testing was NA Temperatures: Testing power (a) to establish the proper setting for the condition low speed limiter for the recirculation Applicable reactor parameters not affected pumps and (b) to provide assurance that were monitored during the by EPU) the measured bottom head drain initial heatup, the initial temperature corresponds to the bottom cooldown, and after head coolant temperature during normal recirculation pump trips in operations.

order to determine that adequate mixing of the reactor The original test obtained RPV water was occurring in the temperatures during rapid heatup and lower plenum of the pressure cooldown to confirm thermal analysis vessel. The adequacy of the models. EPU does not significantly affect bottom-drain-line thermocouple RPV temperatures during rapid heatup or as a measure of bottom reactor cooldown. Since thermal analysis vessel temperature was also models were confirmed during the initial determined. startup testing and RPV temperatures are not significantly affected during EPU, this E-46

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER testing is not required for EPU conditions.

STP 17 System Expansion: Y Y Y Initial No (Unit The purpose of the original testing was to X Heatup, 2/3) (Low (a) verify the reactor drywell piping was System expansion checks Power power free and unrestrained in regard to thermal were made of major equipment Testing condition expansion, (b) verify that suspension and piping in the nuclear steam not affected components were functioning as supply system during heatup to by EPU) required, and (c) provide data for assure components are free to calculation of stress levels in nozzles and move as designed and Yes (Unit 1) weldments.

adjustments were made as (EPU necessary for freedom of startup test) Unit 2/3 - Full system expansion due to movement. thermal effects is experienced at low power conditions and does not increase in proportion to power level. Since EPU does not include a reactor vessel pressure increase nor a corresponding primary coolant temperature increase, the thermal expansion of drywell piping is not affected by EPU conditions.

Therefore, this test is not required for EPU.

Unit 1 - Due to the 30 psi reactor pressure increase (and associated temperature increase), system expansion checks will be made for major equipment and piping in the nuclear steam supply system during heatup to assure components are free to move as designed and adjustments will be made as necessary for freedom of movement.

STP 18 Core Power Distribution: Y Y Y Power Yes The purpose of the original testing was to X X X X X X (Standard (a) confirm the reproducibility of the TIP E-47

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER Core Power distribution, Testing plant system readings, (b) determine the core including power symmetry, was procedure) power distribution in three dimensions, obtained during the power and (c) determine core power symmetry.

ascension program. Axial power traces were obtained at There are no changes to the TIP system each of the TIP locations. The as a result of the EPU. TIP data and results of the complete set of Core Monitoring System data are taken TIP traces were evaluated to and analyzed to determine TIP determine core power asymmetry and core power symmetry symmetry. during each startup refueling test program in accordance with Technical Specifications and approved plant procedures. This testing meets the intent of the original testing.

STP 19 Core Performance: Y Y Y Initial Yes (EPU The purpose of the original testing was to X X X X X X Heatup, startup test) (a) evaluate the core thermal power and Unit 1 & 2 - Core power level, Power (b) evaluate core performance maximum heat flux, Testing parameters of Maximum Linear Heat recirculation flow rate, hot Generation Rate (MLHGR), Minimum channel coolant flow, MCHFR, Critical Power Ratio (MCPR), and fuel assembly power, and Maximum Average Planar Linear Heat steam qualities were Generation Rate (MAPLHGR).

determined at existing power levels. Plant and incore Core performance parameters will be instrumentation, conventional calculated during EPU to verify they heat balance techniques and remain within limits as part of a careful, core performance worksheets monitored approach to the maximum and nomograms were used. EPU power level. This monitored This was performed above 10 approach to EPU power levels meets the percent power and at various intent of the original testing.

pumping conditions and independent of the process computer functions.

Unit 3 - Core parameters were E-48

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER evaluated by manual calculations, the process computer, or the off-line computer program BUCLE.

STP 20 Electrical Output and Heat Y Y Y Power Yes (EPU Units 1, 2, 3 - The purpose of the original X Rate (U1/U2): Testing startup test) testing was to demonstrate that the plant Steam Production (U3): net electrical output and net heat rate requirements are satisfied. This test was The plant gross electrical used to demonstrate reactor vendor output was measured during warranty requirements and does not sustained operation pertain to the safe operation of the plant.

(maintained for 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />) at Plant electrical output versus reactor rated conditions to thermal power will be closely observed demonstrate that the during power ascension to determine guaranteed gross electrical plant heat rate.

output requirements were satisfied without exceeding the reactor power level warranty and to determine a preliminary net plant heat rate value.

For Unit 3 only, the steam production rate was measured during two 2-hour periods at conditions prescribed in the Nuclear Steam Generating System warranty.

STP 21 Flux Response To Rods: Y Y Y Power Yes The purpose of the original testing was to X X X X X X Testing (Standard demonstrate the stability of the core local Rod movement tests were plant power-reactivity feedback mechanism made at chosen power levels procedure) with regard to small perturbations in to demonstrate that the reactivity caused by rod movement.

transient response of the E-49

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER reactor to a reactivity Reactor core stability monitoring is perturbation was stable for the performed continuously by a dedicated full range of reactor power. A system and by Operations department centrally located rod was oversight. Core stability monitoring is moved and the neutron flux performed without moving control rods.

signal from a nearby LPRM Analytical stability evaluations are core chamber measured and reload dependent and are performed for evaluated to determine the each reload fuel cycle. The combination dynamic effects of rod of monitoring and evaluations fulfills the movement. intent of the original startup test.

STP 22 Pressure Regulator: Y Y Y Initial Yes (EPU The purpose of the original testing was to X X X X X X X Heatup, startup test) (a) determine the optimum settings for The pressure setpoint was Power the pressure control loop by analysis of decreased rapidly and then Testing the transients induced in the system by increased rapidly in steps and means of the pressure regulators, (b) the response of the system demonstrate the takeover capability of was measured. The backup the backup pressure regulator upon regulator was tested by failure of the controlling pressure increasing the operating regulator and to set spacing between the pressure regulator setpoint setpoints at an appropriate value, and (c) rapidly until the backup demonstrate smooth pressure control regulator took control. The load transition between the control valves and reference setpoint was bypass valves when reactor steam reduced, and the test repeated generation exceeds steam used by the with the bypass valves in turbine.

control. The response of the system was measured and The original EHC system has been evaluated and regulator modified to incorporate a digital control settings optimized. system with redundant channels which eliminates the need for a backup regulator. As discussed in EPU safety analysis report Section 5.3.13, pressure control operational testing will be performed during EPU power ascension.

E-50

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER For each defined test condition, the pressure control system response to pressure setpoint testing will be evaluated. EPU safety analysis report Section 5.3.11 also discusses that EPU startup ascension test or normal plant surveillance will be used to validate the TSV/TCV scram bypass interlock STP 23 Feedwater System: Y Y Y Power Yes (EPU The purpose of the original testing was to X X X X X X Testing startup test) (a) adjust the feedwater control system U1/U2/U3 - Reactor water level for acceptable reactor water level control, setpoint changes were used to (b) demonstrate stable reactor response evaluate and adjust the to subcooling changes, and (c) feedwater control system demonstrate the capability of the settings for all power and automatic core flow runback feature to feedwater pump modes. prevent low water level scram following U1 - Each feedwater pump the trip of one feedwater pump.

was operated through its flow For purposes (a) and (b), as discussed in range to verify acceptable EPU safety analysis report Section 5.2.2, pump linearity. Response time control system tests will be performed at on each feedwater pump was the appropriate plant conditions for that verified by changing the flow test at each of the power increments, to by 10 percent and measuring show acceptable adjustments and the turbine speed and flow operational capability. The feedwater response times. level control system will be tuned via post U2/3 - One of the three modification testing to account for the operating feedwater pumps increased EPU pumping capacity of the was tripped and the automatic feedwater system. The feedwater level flow runback circuit acted to control system is capable of controlling drop power to within the the reactor vessel level to prevent both capacity of the remaining level increases or decreases such that no pumps. new ECCS trips or actuations will occur based on vessel level. Therefore, this E-51

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER post modification testing will ensure proper operation of the system.

For purpose (c), the original Feedwater System startup testing included a feedwater pump trip test. EPU safety analysis report Section 9.1.3 states that the loss of one feedwater pump event only addresses operational considerations to avoid reactor scram on low reactor water level (Level 3). This requirement is intended to avoid unnecessary reactor shutdowns. This capability will be verified at a high reactor power condition. Based upon the upgrades to the condensate, condensate booster, and feedwater pumps, it is not expected that a recirculation pump runback will occur during this test.

STP 24 Bypass Valves: Y Y Y Power Yes The purpose of the original testing was to X Testing (Standard (a) demonstrate the ability of the One of the turbine bypass plant pressure regulator to minimize the valves was tripped open and procedure) reactor pressure disturbance during an closed. The pressure transient abrupt change in reactor steam flow and was measured and evaluated (b) demonstrate that a bypass valve can to aid in making final be tested for proper functioning at rated adjustments to the pressure power without causing a high flux scram.

regulator.

As stated in EPU safety analysis report Section 5.2.1.1, no modifications to the turbine control valves or the turbine bypass valves are required for operation at the EPU conditions. Confirmation testing will be performed during power E-52

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER operation.

STP 25 Main Steam Isolation Valves: Y Y Y Initial Not For purposes (a) and (b), the percent X Heatup, applicable power level at which a single MSIV can Fast full closure of each MSIV Power (Part a & b) be closed without a scram may change was performed at hot standby Testing due to EPU instrument-related changes.

and selected power levels to No (Part c) However, determination of this power determine a) the maximum level is a plant capacity consideration power conditions at which Yes (Part d) rather than a demonstration of proper individual valve full closure operation of a system(s) and this testing tests could be performed (Standard plant is not a requirement to safely implement without a reactor scram, b) functional checks (10 percent procedure) EPU.

closure) of each isolation valve For purpose (c), a simultaneous full were performed at selected closure of all MSIVs is a large transient power levels above the test. See Section III.C for a discussion of maximum power condition for test performance.

individual MSIV full closure, c) a test of simultaneous full For purpose (d), valve movement (i.e.

closure of all MSIV's was stroke) will be verified via a standard performed at about 100 plant procedure. Proper seating of the percent of rated thermal power valves will be conducted via performance and proper operation of the of the Appendix J local leak rate testing.

relief valves and the RCIC were shown, reactor process variable were monitored to determine the transient behavior of the system during and following each isolation test d) MSIV delay and movement times were determined and proper seating of the MSIVs demonstrated.

E-53

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER STP 26 Relief Valves: Y Y Y Power Yes The purpose of the original testing was to X Testing (Standard (a) verify the proper operation of the The main steam relief valves plant primary system relief valves, (b) were each opened manually. procedure) determine the capacity and response Capacity of each relief valve characteristics of the relief valves, and (c) was determined by the amount verify proper seating of the relief valves the bypass or control valves following operation.

closed to maintain reactor pressure. Proper reseating of Unit 2/3 - For purposes (a) and (c), the each relief valve was verified Main Steam Relief Valve Manual Cycle by observation of temperatures Test is performed once per operating in the relief valve discharge cycle in accordance with Technical piping. Specifications and approved plant procedures. During unit startup from U2/3 addition - At selected test refueling, each MSRV is verified to open conditions, each valve was and close when manually actuated at manually actuated and at least rated reactor pressure.

one valve was timed.

As described in EPU safety analysis U2 addition - Additional timing Section 3.1, no MSRV setpoint increase data was obtained in is needed because there is no change in conjunction with those the dome pressure or simmer margin.

transient tests which result in Therefore, there is no effect on valve automatic relief valve opening. functionality (opening/closing).

Therefore, Technical Specification testing of MSRV operation is considered appropriate for EPU.

For purpose (b), the data that was collected regarding relief valve capacity is still valid for the installed valves and is not impacted by operation at EPU conditions. Therefore it is not necessary to repeat this portion of the original test.

Unit 1 - As part of EPU implementation E-54

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER and the associated 30 psi pressure increase, the main steam relief valves will be reset to the higher set pressures.

For purposes (a) and (c), the Main Steam Relief Valve Manual Cycle Test is performed once per operating cycle in accordance with Technical Specifications and approved plant procedures. . During unit startup from refueling, each MSRV is verified to open and close when manually actuated at rated reactor pressure.

For purpose (b), the data that was collected regarding relief valve capacity is still valid for the installed valves and is not impacted by operation at EPU conditions. Therefore it is not necessary to repeat this portion of the original test.

STP 27 Turbine Stop and Control Y Y Y Power No A turbine stop/control valve closure or NA Valve Trips (U1): Testing generator load rejection at high power is Turbine Trip and Generator a large transient test. See Section III.C Load Rejection (U2/3): for a discussion of test performance.

Unit 1-The stop or control The ability to ride through a load rejection valves were tripped closed at within bypass capacity without a scram is selected reactor power levels performed at a low power level and is (50% and 100%). Neutron thus not affected by operation at EPU flux, feedwater flow and and therefore this test is not required to temperature, vessel water level be re-performed.

and pressure were monitored.

Responses of selected control valves, stop valves, relief valves, and bypass valves were recorded. The ability to E-55

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER ride through a load rejection within bypass capacity without a scram was demonstrated at low power (25%).

Unit 2/3 -The turbine stop valves were tripped at selected reactor power levels (50% and 100%) and the main generator breaker tripped in such a way that a load imbalance occurred. Several reactor and turbine operating parameters were monitored to evaluate the response of the bypass valves, relief valves, and reactor protection system. The peak values and change rates of reactor steam pressure and heat flux were determined.

The ability to ride through a load rejection within bypass capacity without a scram was demonstrated at low power (25%).

STP 29 Flow Control: Y Y N Power Yes The original testing purpose was to (a) X Testing (Standard determine the plant response to changes Various process variables were plant in the recirculation flow, (b) optimize the recorded while step changes procedure) setpoints of the flow controller, and (c) were introduced into the demonstrate the plant load following recirculation flow control capability in the different flow control system (increased and modes.

decreased) at chosen points on the 50, 75, and 100 percent Increased voids in the core during normal E-56

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER load lines. Up to 30 EPU power operations requires a slight percent/minute change in increase in recirculation drive flow to recirculation flow was made achieve the same core flow. Because from all flow conditions down to adequate core flow can be maintained the lower limit of the Master without requiring any changes to the Flow Controller and return. recirculation system and only a small Load following capability was increase in pump speed for the same demonstrated in all flow control core flow, the response to flow changes modes. will be similar to that displayed during the original startup testing.

U2 - Ramp changes were made at rates within the range System modifications since the original of 10 percent to 30 percent per plant configuration include a new flow minute. Load following control system and the installation of capability was demonstrated in Variable Frequency Drives (VFDs) for the the automatic and master Reactor Recirculation System pump manual flow control modes. motors. Reactor Recirculation System testing and tuning of the current flow control system is performed during each refueling startup during vessel hydro conditions and also at power conditions to analyze system response to speed demand of small and large changes. The testing performed during the normal refueling test program will meet the intent of the original test objectives during startup and power ascension.

STP 30 Recirculation System: Y Y Y Power Yes The original testing purpose was to (a) X Testing (Standard evaluate the recirculation flow and power U1/3-Single and both plant level transients following trips of one or recirculation pumps were procedure) both of the recirculation pumps, (b) to tripped at various power levels. obtain recirculation system performance Two pump trips were initiated data, and (c) to verify that no recirculation by tripping the MG set drive system cavitation will occur on the E-57

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER motors. One single pump trip operable region of the power-flow map.

at 50 percent power was initiated by opening the Increased voids in the core during normal generator field breaker and the EPU power operations requires a slight remaining single pump trips increase in recirculation drive flow to were initiated by tripping the achieve the same core flow. Because MG set drive motor. Reactor adequate core flow can be maintained operating parameters were without requiring any changes to the recorded during the transient recirculation system and only a small and at steady-state conditions. increase in pump speed for the same core flow, the response to flow changes U1-The jet pump resulting from either a single or two pump instrumentation was calibrated trip will be similar to that of original to read total flow. startup testing. No additional testing is required for EPU.

U1/2-MCHFR evaluations were made for conditions Verification of jet pump calibration is encountered during the accomplished by standard plant transient. procedures. See STP 35 for additional information.

U2-With the recirculation pumps operating at the speed The verification of the non-occurrence of corresponding to rated flow at cavitation is not impacted by EPU since rated power, power was this condition only occurs at a low reduced by inserting rods to 23 power/low flow condition. Therefore, it is percent power where the not necessary to repeat this portion of the recirculation pumps would original testing.

automatically run back to 20 percent speed and a check was made to determine if recirculation or jet pump cavitation occurred.

STP 31 Loss of Turbine-Generator and Y Y Y Power Yes The ability to mitigate a loss of offsite X Offsite Power: Testing (Standard power is performed at a low power level plant and is thus not affected by operation at E-58

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER The loss of auxiliary power test procedure) EPU and therefore this test is not was performed at 25 percent required to be re-performed. The ability (U2/3-20 to 30 percent) of of the electrical equipment to respond to rated power. The proper a loss of offsite power is demonstrated by response of reactor plant testing required by the Technical equipment, automatic Specifications for the onsite and offsite switching equipment, and the power distribution systems.

proper sequencing of the diesel generator loads were checked. Appropriate reactor parameters were recorded during the resultant transient.

STP 32 Recirculation M-G Set Speed Y Y Y Power Yes The original testing purpose was to (a) X Control (U1/2): Testing (Standard determine the speed control Recirculation Speed Control plant characteristics of the MG sets in the and Load Following (U3): procedure) recirculation control system, (b) obtain acceptable speed control system Several small step changes in performance, and (c) determine the speed demand were input at maximum allowable pump speed.

various pump speeds and appropriate recirculation loop This test determined the original as built transient signals recorded to characteristics of the Recirculation demonstrate response Control System. Increased voids in the performance over the full core during normal EPU power speed range with small speed operations requires a slight increase in demand step tests. recirculation drive flow to achieve the same core flow. Because adequate core (Note: The Recirculation M-G flow can be maintained without requiring Sets replaced with Variable any changes to the recirculation system Frequency Drives.) and only a small increase in pump speed for the same core flow, the response to flow changes will be similar to that of original startup testing.

E-59

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER System modifications since the original plant configuration include a new flow control system and the installation of Variable Frequency Drives (VFDs) for the Reactor Recirculation System Pumps.

Reactor Recirculation System testing and tuning of the current flow control system is performed during each refueling startup during vessel hydro conditions and also at power conditions to analyze system response to speed demand of small and large changes. The testing performed during the normal refueling test program will meet the intent of the original test objectives during startup and power ascension at EPU.

STP 33 Main Turbine Stop Valve N Y Y Power Yes The original testing purpose was to X Surveillance Test: testing (Standard demonstrate acceptable procedures for plant daily turbine stop valve surveillance tests At several test points, procedure) at a power level as high as possible individual main turbine stop without producing reactor scram.

valves were closed. The response of the reactor was Individual main turbine stop valves must recorded and the maximum be closed periodically during plant possible power level for operation as required for plant performance of this test along surveillance testing. As described in with the 100 percent power EPU safety analysis report Section 3.5.2, flow control line established. the TSV bounding closing time was Each stop valve closure was utilized in EPU analysis.

manually initiated and reset.

Rate of valve stroking and timing of the close-open sequence was chosen to minimize the disturbance E-60

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER introduced.

STP 34 Vibration: Y Y Y Power Yes The original testing purpose was to X X X X X X X (Unit 2/3 testing (Standard obtain vibration measurements on only) Vibratory responses were plant various reactor components to recorded at various procedure) demonstrate the mechanical integrity of recirculation flow rates at the system to flow induced vibration and temperatures below 150°F to check the validity and accuracy of the STP 90 using strain gages on in-core (Unit 1 analytical vibration model.

guide tubes, control rod stub only)- tubes, shroud support legs, Previous startup tests obtained vibration and jet pump riser braces; measurements on various reactor accelerometers on the pressure vessel internals to demonstrate recirculation loops and the mechanical integrity of the system displacement gages on the under conditions of flow induced shroud, steam separator and vibration, and to check the validity of the jet pumps. Portable vibration analytical vibration model.

sensor surveys were made on the recirculation loops and With the exception of the steam dryer, differential pressure the system flows associated with the measurements made across reactor vessel internals are unchanged the core plates, shroud head by operation at EPU conditions and thus and shroud wall. At hot, two- additional testing is not required. The phase flow conditions, similar small increase in normal flow through the measurements were made on recirculation loops are within the values the in-core guide tubes, previously tested. Changes in the shroud, jet pump riser and vibrations in the recirculation loops will be shroud head. The results of detected by the permanently installed vibration measurements made recirculation pump vibration monitoring.

at other BWR installations will Analysis of the reactor vessel internals at be considered in the final EPU power level was performed to selection of components to be ensure that the design continues to tested. comply with the existing structural requirements. Results of this analysis are provided in EPU safety analysis E-61

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER Section 3.3 and are discussed in EPU submittal Enclosures 10 as follows:

Flow induced vibration effects of the following components were evaluated per the requirements of NRC Regulatory Guide 1.20: Shroud, Jet Pumps, Jet Pump sensing lines, Steam Dryer, and Core Plate.

The following components have been evaluated and determined to be structurally adequate to withstand the effects of flow induced vibrations: Guide Rods, Top Head Instrument Nozzle, Head Spray Nozzle, Top Head Vent Nozzle, Core Spray Sparger, Core Spray Piping, Fuel Assembly, and Shroud Head Bolts, Steam line Nozzle, Water Level Instrument Nozzle, and Top Guide.

STP 35 Recirculation and Jet Pump Y Y Y Power Yes The original testing purpose was to X System Calibration: Testing (Standard perform a complete calibration of the plant installed recirculation system flow A closely controlled pressure procedure) instrumentation. The flow was applied simultaneously to instrumentation was recalibrated when an entire loop to obtain an the new digital recirculation flow control integrated calibration check of system was installed. Since the total the system instrumentation. core flow does not change for EPU Actual calibration of the jet conditions, the recirculation jet pumps will pump flow instrumentation was not require recalibration for EPU completed during hot conditions. If there are any indications or pressurized operation by requiring jet pump recalibration, it would comparison of the single and be performed using a standard plant double tapped pressure drops E-62

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER as a function of flow. procedure.

STP 36 Equalizer Open: Y N N Power No The original testing purpose was to (a) NA Testing explore the allowable operating range Testing was performed to and performance of the recirculation explore one pump operation system under conditions of one pump with the equalizer valve open. operation with the equalizer line valves Initial valve opening was made open, and (b) to develop operating at a high pump speed by rapid procedures for one pump equalizer open jogging until the inactive loop operation.

jet pumps go from reverse to forward flow. Successive valve The Unit 1 and 3 equalizer valves were openings and pump speed removed during the recirculation piping increases were schedules to replacement during the recovery. The avoid pump loop P, pump Unit 2 valves, although still existent, are speed and pump motor current restricted from being opened during limits. When the valve was full normal operation. Therefore, this test is open or when limits are no longer applicable and is not required reached the available for EPU.

operating region was explored and data obtained. The test was concluded by rapidly closing the equalizer valve while recording the transient.

STP 39 Water Level Verification in Y N N Initial Yes The purpose of the original testing was to X X X X X X (Unit 1 Reactor Vessel: Heatup, (Standard a) verify the calibration and agreement of only). Power plant the various narrow and wide range level See The first part of testing Testing procedure) indicators under various conditions and STP 9 measured the Yarway b) to demonstrate the ability of the for Units reference to verify agreement feedwater control system to regulate 2/3) with the temperature correction reactor water level.

factor used in calibration. The second part verified the ability As part of the restart efforts, the Yarway of the feedwater control system temperature equalizing columns are E-63

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER to regulate reactor water level removed and the instruments replaced at 50 percent flow/50 percent with analog trip system devices. The power and 100 percent hardware associated with this new flow/100 percent power. reactor water level instrumentation is not modified and normal operational water level and level setpoints (alarms/trips/actuations) are not changed by EPU. The demonstration of procedures and operational validation for EPU therefore need not be repeated.

Demonstration of the ability of the feedwater control system to regulate reactor water level is accomplished by standard plant procedures.

STP 70 Reactor Water Cleanup Y Y Y Initial Yes The original testing purpose was to X X X X X X X System: Heatup (Standard demonstrate the operability of the RWCU plant system under actual reactor operating Testing was performed to procedure) temperature and pressure.

demonstrate the operability of the reactor water cleanup As described in the EPU safety analysis system under actual reactor report Sections 3.10, RWCU system operating temperature and operation at EPU slightly decreases the pressure. temperature within the system and the FW iron input to the reactor increases as a result of the increased FW flow. The effects of EPU on the system function have been reviewed and it was determined the system can perform adequately during EPU with the original RWCU flow. Therefore, no changes are required to meet the performance requirements for the system.

Confirmation that the system meets TS E-64

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER requirements for operability for startup and power ascension is required by the refueling test program.

STP 71 Residual Heat Removal Y Y Y Power Yes The original testing purpose was to X System: Testing (Standard demonstrate the ability of the RHR plant system to (a) remove residual and decay Testing was performed to procedure) heat from the nuclear system so that demonstrate the ability of the refueling and nuclear system servicing RHR system to remove can be performed and (b) remove heat residual and decay heat from from the pressure suppression pool the nuclear system so that water.

refueling and nuclear system servicing can be performed, As described in the EPU safety analysis and remove heat from the report Sections 3.9, the RHR system is pressure suppression pool designed to restore and maintain the water. reactor coolant inventory following a LOCA and remove decay heat following reactor shutdown for normal, transient, and accident conditions. The EPU effect on the system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of heat discharged. The effects of the system functional capability have been reviewed and it was determined that the system functional basis continues to ensure that accident containment temperature limits are not exceeded and only plant availability is effected for normal reactor shutdown.

Therefore, no changes are required to meet the performance requirements for the system. Confirmation that the system meets TS requirements for operability for E-65

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER startup and power ascension is required by the Technical Specifications.

STP 72 Drywell Atmosphere Cooling Y Y Y Power Yes The original testing purpose was to verify X X X X X X X System: Testing (Standard the ability of the Drywell Atmosphere plant Cooling System to maintain design The Drywell Atmosphere procedure) conditions in the drywell during operating Cooling System was placed in conditions.

operation and its ability to maintain the temperature in the As discussed in the EPU safety analysis drywell. report Section 6.4.3, the change in vessel temperature is minimal and does not result in any significant increase in drywell cooling loads. The testing performed during the normal refueling test program will meet the intent of the original test objectives during startup and power ascension at EPU.

STP 73 Cooling Water Systems: Y Y Y Initial Yes (EPU The original testing purpose was to verify X X X X X Heatup, startup the performance of the RBCCW and the Testing was performed to Power test.) RCW systems is adequate with the verify that the performance of Testing reactor at rated condition.

the RBCCW and raw cooling water systems is adequate with As described in the EPU safety analysis the reactor at rated condition. report Sections 6.4.3 and 6.4.4, the EPU heat load increases for these systems are minimal and sufficient heat load removal capacity is available with the current system flow rates to accommodate heat removal at EPU normal operating conditions. Therefore, no changes are required to meet the performance requirements for the system and system testing is governed by E-66

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER approved plant procedures which remain valid for EPU operations.

STP 74 Off Gas System: N Y Y Power Yes The original testing purpose was to verify X X X X X X X Testing (Standard the proper operation of the Off Gas Testing was performed to plant system over its expected operating verify the proper operation of procedure) parameters and to determine the the Off Gas system over its performance of the activated carbon expected operating parameters adsorbers.

and to determine the performance of the activated As described in the EPU safety analysis carbon adsorbers. report Section 8.2.1, the EPU hydrogen flow rates and concentrations are still within the design limits of the system and the system components have sufficient design margin to handle the increase in thermal power for EPU with exceeding the system design limits of temperature, flow rates, or heat loads. Therefore, no changes are required to meet the performance requirements for the system. Confirmation that the system meets requirements for operability for startup and power ascension is required by a standard plant procedure.

STP 75 Reactor Shutdown from N N Y Power Yes The purpose of this test was to X Outside the Main Control testing (Standard demonstrate that the plant was designed Room: plant and constructed with adequate procedure) instruments and controls to permit safe With the plant operating at reactor shutdown from outside the main greater than 10 percent control room and maintain it in a safe generator output, the reactor condition, that the minimum number of was scrammed by closing the personnel required by the TS are MSIVs from the backup control adequate without affecting the safe, E-67

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER station. Operators manned continuous operation of the other units, their backup control stations as and that the plant emergency operating described in the emergency instructions are adequate.

operating instructions. The RCIC system was operated As described in the EPU safety analysis from the backup controls to report Section 10.6, EPU does not supply water to the reactor change any of the plant automatic safety vessel. Suppression pool functions. Also, after the applicable cooling was placed in automatic responses have initiated, the operation using the backup subsequent operator actions for plant controls. An extra licensed safety do not change for EPU. Additional operator remained in the main operator training will be required to control room to assure that the enable plant operation at the EPU power test was terminated and control level. Training required to operate the returned to normal in the event plant following EPU will be conducted of any unexpected conditions prior to restart of the units at the EPU occurred. The test was conditions. Standard plant procedures terminated when it was that are performed once per operating assured that the reactor could cycle ensure that if the control room be maintained in a safe hot becomes inaccessible, the plant can be standby condition from the placed and maintained in MODE 3 from backup controls. the backup control panel and the local control stations.

STP 92 Steam Separator-Dryer: Y N N Open Vessel Yes (EPU The original startup testing took samples X X X X X Testing, startup test) from within the vessel and the inlet at the Samples were taken from the Initial steam line for the determination of inlet and outlet of the steam Heatup, carryunder and carryover. EPU testing dryers, and the inlet at the Power will determine steam separator-dryer steamline at various power Testing moisture carryover. For this testing, MSL levels at chosen water levels moisture content is considered equivalent and recirculation flow rates. to the steam separator-dryer moisture The amount of carryunder was carryover. This MSL moisture content estimated from these samples test data will determine the steam and carryover was determined separator-dryer performance (i.e.,

E-68

TABLE 1 COMPARISON OF BFN INITIAL STARTUP TESTING AND PLANNED EPU TESTING ORIGI- ORIGINAL EPU TEST CONDITIONS ORIGINAL TEST NAL TEST UNIT ORIGINAL TESTING PERCENT OF 3293 MWT (OLTP)

DESCRIPTION EVALUATION / JUSTIFICATION /

TEST S/U TEST PLANNED (DERIVED FROM UFSAR NOTES1 NUM- PHASE FOR EPU SECTION 13.5) U1 U2 U3 <90 902 1003 105 110 115 EPU BER from Na-24 activities in moisture carryover) for the EPU and core samples taken from the outlet conditions in effect at the time of the test.

of the steam dryers. This testing will meet the intent of the original test objectives.

Notes:

1. In this table, EPU Safety Analysis Report refers to the PUSAR for Unit 1 and the PUSAR and/or the FUSAR for Units 2 and 3 as applicable.

2. Power relative to Current Licensed Thermal Power (CLTP).

3. Column applicable to Unit 1 only.

E-69

TABLE 2 COMPARISON OF SRP 14.2.1 TEST MATRIX AND BFN INITIAL STARTUP TESTS BFN Initial Startup Test?

SRP Power Ascension Test (Test Numbers)

Steady-State Power Ascension Testing Conduct vibration testing and monitoring of Yes reactor vessel internals and reactor coolant (34, 90) system components Measure power reactivity coefficients (PWR) or Yes power vs. flow characteristics (BWR) (30)

Steady-state core performance Yes (19)

Control rod patterns exchange No (Performed Analytically)

Control rod misalignment testing N/A (PWR only)

Failed fuel detection system No (No Applicable System)

Plant process computer Yes (13)

Calibrate major or principal plant control systems Yes (14, 15, 22, 23, 24, 29, 32)

Main steam and main feedwater system Yes operation (22, 23)

Shield and penetration cooling systems No (Standard Procedure)

ESF auxiliary and environmental systems No (Standard Procedure)

Calibrate systems used to determine reactor No thermal power (Standard Procedure)

E-70

TABLE 2 COMPARISON OF SRP 14.2.1 TEST MATRIX AND BFN INITIAL STARTUP TESTS BFN Initial Startup Test?

SRP Power Ascension Test (Test Numbers)

Chemical and radiochemical control systems Yes (1)

Sample reactor coolant system and secondary Yes coolant systems (1)

Radiation surveys Yes (2)

Ventilation systems (including primary Yes containment and steam line tunnel) (72)

Acceptability of reactor internals, piping, and Yes component movement, vibrations, and (17, 34, 90) expansions Transient Testing Relief valve testing Yes (26)

Dynamic response of plant to design load swings Yes (22, 23, 25, 27, 29, 30, 32)

Reactor core isolation cooling functional test Yes (14)

Dynamic response of plant to limiting reactor Yes coolant pump trips or closure of reactor coolant (30) system flow control valves (Reactor coolant recirculation pump trip test)

Dynamic response of the plant to loss of No feedwater heaters that results in most severe feedwater temperature reduction (Performed Analytically)

Dynamic response of plant to loss of feedwater Yes flow (23)

Dynamic response of plant for full load rejection Yes (Loss of Offsite Power Testing) (27, 31)

E-71

TABLE 2 COMPARISON OF SRP 14.2.1 TEST MATRIX AND BFN INITIAL STARTUP TESTS BFN Initial Startup Test?

SRP Power Ascension Test (Test Numbers)

Dynamic response of plant to turbine trip Yes (Turbine trip or generator trip) (27, 31)

Dynamic response of plant to automatic closure Yes of all main steam isolation valves (25)

E-72

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Main Turbine

• Replace HP Turbine diaphragms and No No No

• Turbine Balancing (if rotor buckets required)

• Replace HP Rotor/LP Rotors (Unit 1

• Overspeed Test only)

• Control and Stop Valve

• Replace springs, bonnets, washers, testing bellows, & bolting on 6 cross around

• Relief valve bench testing relief valves to permit increased set pressure

• Replace miter bend elbows in the condenser spray piping with long radius elbows to reduce back pressure Turbine Sealing

• Modify the size of the steam seal No No No

• Condenser Vacuum Steam unloader valves and associated piping to testing monitor steam seal allow the turbine sealing system to header pressure accommodate the larger steam flow

• Calibration of the Steam requirements Seal Header Pressure controller

• Inservice leak test E-73

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Condensate Pumps

• Replace 2 impellers in each of 3 pumps Yes Yes Yes

• Verification of pump flow

• Install 3 - 1250 hp motors and head

• Recalibrate relay settings

• Monitoring of pump and motor parameters (flow

• Recalibrate/replace pump & motor pressure, temperatures, instrumentation etc.)

• Modify HVAC ductwork

• Instrumentation calibration and functional testing Condensate Booster

• Replace 3 pumps Yes Yes Yes

• Verification of pump flow Pumps

• Install 3 - 3000 hp motors and head

• Recalibrate relay settings

• Monitoring of pump and motor parameters (flow

• Recalibrate/replace pump & motor pressure, temperatures, instrumentation etc.)

• Modify HVAC ductwork

• Instrumentation calibration and functional testing E-74

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Feedwater Pumps

• Replace 3 pumps Yes Yes Yes

• Balancing and Turbines

• Recalibrate pump instrumentation and

• Overspeed testing control system for increased flows at Controls Tuning EPU conditions

• Verification of pump flow

• Replace turbine/pump coupling and head

• Replace turbine rotor, diaphragms and

• Monitoring of pump and buckets turbine parameters (flow

• Recalibrate/replace turbine pressure, temperatures, instrumentation etc.)

• Instrumentation calibration and functional testing Moisture Separators

• Change vanes and add perforated plate No No No

• Moisture removal on moisture separators effectiveness testing

• Modify internal drains as needed

• Inservice leak test

• Installation testing (flow, temperature pressure, etc.)

E-75

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Feedwater Heaters

• Upgrade heater shell pressure No No No

• Relief Valve bench testing certification

• Installation Testing (flow,

• Rerate tube side pressure certification temperature, pressure, for FWH 3 etc.)

• Replace level transmitters on FWHs 1, 2

• Instrumentation

&3 calibration and functional

• Repair / replace 18 nozzles on FWHs 1, testing 2&3

• Inservice leak rest.

• Replace relief valves on FWHs 1, 2 & 3

• Relocate extraction steam nozzle &

shorten extraction steam line on FWH 3

• Install new impingement plate & steam duct inside FWH 3

• Reinforce / re-weld pass partition plates in all FWHs

• Install manway stiffeners on FWH 3 Main Condenser

• Replace #2, #3 and #4 bellows with No No No

• Installation Examination Extraction Steam upgraded bellows Bellows E-76

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Condensate

• Install 1 new vessel with valves & digital No No No

• Control system functional Demineralizers controls testing

• Upgrade controls on 9 existing vessels

• Initial installation Startup to digital (Unit 2 only) test (flow, temperature,

• Install digital control on 9 existing pressure, etc.)

vessels (Unit 1 only)

• Replace valves for increased reliability Steam Packing

• Install 24 piping & flow control valve to No No No

• Valve testing Exhauster Bypass accommodate increased condensate flows at EPU conditions Drywell Building Steel

• Modify building steel beams and Yes Yes No

• Applicable structural connections as required for load installation testing changes at EPU conditions Torus Attached Piping

• Modify supports and snubbers as Yes Yes No

• Applicable structural required due to EPU conditions installation testing Main Steam Supports

• Modify supports as required for load Yes Yes No

• Applicable structural changes due to EPU conditions installation testing

• Vibration monitoring of Main Steam and Feedwater Piping and components E-77

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Steam Dryer

• Perform identified modifications required No No No

• Testing is discussed in to maintain Dryer structural integrity at TVA's February 23, 2005 EPU Conditions letter, Reply 5.a (1)

Main Steam Relief

• Increase mechanical setpoint 30 psi with Yes Yes No

• Bench testing of setpoints Valves (Unit 1 only) 3% tolerance due to increased reactor

• Remote manual opening pressure at operating reactor pressure Motor Operated

• Modification to GL 89-10 MOVs to Yes Yes No

• MOVATS Valves (Unit 1 only) accommodate 30 psi pressure increase Reactor Recirculation

• Revise electrical protection system Yes Yes No

• Applicable Pump Motors setpoints instrumentation

• Revise temperature monitoring setpoints calibrations

• Assess additional heat load on plant

• Vibration monitoring HVAC & cooling water systems

• Controls tuning and

• Assess power cable voltage drop system operation during increase due to higher current vessel hydro

• Revise pump/motor vibration monitoring setpoints

• Re-rate pumps and motors for 120%

Power/105% Core Flow operating conditions E-78

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Jet Pumps

• Install sensing line clamps as required to Yes Yes No

• None required reduce vibration due to Recirculation Pump vane passing frequencies at EPU conditions Main Generator

• Recalibrate/replace pressure regulators No No No

• Field Installation testing System and pressure switches

• Instrumentation

• Increase generator hydrogen to 75 psig calibration and functional to operate at increased loads testing

• Rewind generator stator and generator

• Monitoring of system (i.e.,

field (Unit 1 only) voltage, amps, temperature) during power ascension Isolation Phase Bus

• Modify Isolation Phase Bus Duct Cooling No No No

• Verification of system Duct Cooling System to remove Bus Duct heat under flow, both air and water EPU conditions Main Bank

• Install 3-500 MVA transformers per unit No No No

• Field Installation testing Transformers

• Install 2-500 MVA spares (U1/2 & 3)

• Deluge spray down

• Upgrade oil and water deluge systems testing

• Upgrade relaying as needed

• Performance monitoring

• Replace 5 - breakers & disconnects ICS/SPDS

• Update as needed based on NSSS and No No No

• Performance monitoring BOP instrument changes E-79

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Reactor Recirculation

• Install ~ 35 temporary sensors on RR, No No No

• Collect and analyze Vibration Monitoring RHR, & RWCU in Drywell. vibration data (Unit 2)

• Install data acquisition center in Rx Bldg.

• Conduct testing program before EPU outage.

(Installed on Unit 2 and vibration data obtained during startup from U2C12 Outage.)

Vibration Monitoring

• Install temporary sensors based on No No No

• Collect and analyze ongoing analyses vibration data on selected

• Conduct testing program during power systems ascension Main Steam Isolation

• Replace MSIV poppets and modify Yes Yes Yes

• Stroke time testing Valves operators (Unit 1 only) as required to

• Applicable Technical reduce differential pressure across Specifications testing MSIVs at EPU conditions

• Performance monitoring

• Install 2-inch MSIV stems as required due to increased stem forces caused by EPU MS flow increase EHC Software

• New program inputs & logic for EPU Yes Yes Yes

• Verification of control conditions functions

• Turbine Valve setup

• Controls Tuning E-80

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs High reactor vessel

• Raise setpoint due to increased normal Yes Yes No

• Calibration per applicable pressure scram reactor vessel pressure Technical Surveillance setpoint (Unit 1 only) Specification value Requirements instructions High reactor vessel

• Raise setpoint due to increased normal Yes Yes No

• Calibration per applicable pressure trip for the reactor vessel pressure Technical Surveillance reactor recirculation Specification value Requirements instruction pump (Unit 1 only)

ATWS-RPT reactor

• Raise setpoint due to increased normal Yes Yes No

• Calibration per applicable vessel pressure reactor vessel pressure Technical Surveillance recirculation pump trip Specification value Requirements instructions setpoint (Unit 1 only)

Standby Liquid

• Raise discharge pressure requirement Yes Yes No

• Verified by Surveillance Control System pump due to increased normal reactor vessel Requirement testing discharge test pressure pressure (Unit 1 only)

Turbine Stop Valve -

• Lower percent reactor thermal power to Yes Yes No

• Calibration per applicable Closure and Turbine maintain trip at same absolute thermal Surveillance Control Valve Fast power Requirements instructions Closure, Trip Oil Pressure - Low bypass setpoint Neutron flux - High

• Lower percent reactor thermal power to Yes Yes No

• Calibration per applicable (set down) maintain trip at same absolute thermal Surveillance power Requirements instructions E-81

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Flow Biased

• Adjust slope and intercept for equation to Yes Yes No

• Calibration per applicable Simulated Thermal reflect change in maximum reactor Surveillance Power - High thermal power Requirements instructions RCIC and HPCI

• Raise discharge pressure requirement Yes Yes No

• Verified by Surveillance surveillance testing due to increased normal reactor vessel Requirement testing pump discharge pressure pressure (Unit 1 only)

Main Steam Line high

• Adjust delta P to account for higher flow Yes Yes No

• Calibration per applicable flow rate. Surveillance Requirements instructions ATWS Alternate Rod

• Raise setpoint due to increased normal Yes Yes No

• Calibration per applicable Injection (ARI) high reactor vessel pressure Technical Surveillance reactor pressure Specification value Requirements instructions setpoint (Unit 1 only)

Steam/Feedwater

• Increased flow rate to accommodate Yes No Yes

• Monitored to ensure plant Normal Flow Rate increased reactor thermal power output remains within anticipated operational limits Feedwater and

• Feedwater temperature increase Yes No No

• Change in temperature Reactor Vessel

• Reactor vessel temperature decrease accounted for in Temperature due to increased feedwater flow rate instrument calibration.

Changes E-82

Table 3 Browns Ferry EPU Planned Modifications, Setpoint Adjustments and Parameter Changes

Response

Impacts Action Required to Function Involves Activity Description Mitigate Plant Testing Important to Multiple Transient Safety SSCs Recirculation Pump

• Increased required recirculation pump No No No

• Verification of total core Flow Rate flow rate required to achieve total core flow flow Reactor Pressure

• 30 psi increase in normal operating Yes No No

• Main Steam Relief Valves (Unit 1 only) steam dome pressure Bench test

• Generic Letter 89-10 valves MOVATS tested

• Reactor Vessel pressure test

• HPCI/RCIC flow tests

• Standby Liquid Control System test pressure increased Primary Containment

• Peak pressure and temperature change See Testing See Testing See Testing

• Analytically demonstrated Pressure and Torus Column Column Column to achieve successful Temperature Post system operation (i.e.,

Accident primary containment, ECCS, etc) since actual post accident conditions cannot be simulated

• Integrated containment Leak rate testing

• Local leak rate tests

• ECCS Flow tests E-83