ML11129A188

TS 11-06, Emergency Technical Specification Change for One-Time Extension of Surveillance Requirements Associated with Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation

ML11129A188
Person / Time
Site:	Sequoyah
Issue date:	05/06/2011
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TS 11-06
Download: ML11129A188 (50)

Text

Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing May 6, 2011 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Facility Operating License No. DPR-77 NRC Docket No. 50-327

Subject:

TS 11-06, Emergency Technical Specification Change for One-Time Extension of Surveillance Requirements Associated with Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) is submitting a request for a Technical Specifications (TS) change for Facility Operating License No. DPR-77 for Sequoyah Nuclear Plant (SQN), Unit 1. The proposed TS change would allow a one-time, 31-day extension to monthly TS Surveillance Requirements (SRs) 4.3.1.1.1 and 4.3.2.1.1 for certain instrument functions of the Reactor Trip System and Engineered Safety Feature Actuation System, respectively.

The enclosure provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachment 1 to the enclosure provides the existing Unit 1 TS pages marked-up to show the proposed changes. Attachment 2 to the enclosure provides the existing Unit 1 TS pages retyped to show the proposed changes.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

The SQN Plant Operations Review Committee and the SQN Nuclear Safety Review Board have reviewed this proposed change and determined that operation of SQN in printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 May 6, 2011 accordance with the proposed change will not endanger the health and safety of the public.

Additionally, in accordance with 10 CFR 50.91 (b)(1), TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

TVA requests an expedited review of the proposed change and that this change be processed as an emergency TS change, in accordance with 10 CFR 50.91(a)(5). In order to preclude the shutdown of SQN, Unit 1, TVA requests approval of this TS change by May 13, 2011. Should the Nuclear Regulatory Commission approve this change, the approved TS change will be implemented by May 13, 2011. Performing the SRs referenced above in accordance with the current TSs would pose undue risks to unit operation and could result in an inadvertent shutdown of SQN, Unit 1. Current electrical grid conditions combined with an inadvertent shutdown of SQN, Unit 1, would result in an unstable electrical grid condition. The current electrical grid condition resulted from the severe storms and tornados in the region on April 27, 2011, and could not have been avoided. Additional information on the current electrical grid condition is provided in the enclosure.

There are no regulatory commitments associated with this submittal. Please address any questions regarding this request to Dan Green at 423-751-8423.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 6th day of May, 2011.

Respectfully, R. M. Kr h

Enclosure:

Evaluation of Proposed Technical Specification Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Resident Inspector - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT I NRC DOCKET NO. 50-327 EVALUATION OF PROPOSED TECHNICAL SPECIFICATION CHANGE

Subject:

TS 11-06, Emergency Technical Specification Change for One-Time Extension of Surveillance Requirements Associated with Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES ATTACHMENTS

1. Proposed Technical Specification Changes (Mark-Ups)

2. Proposed Technical Specification Changes (Final Typed)

3. Power System Operation Transmission Alerts E-1

1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Facility Operating License No. DPR-77 for the Tennessee Valley Authority (TVA) Sequoyah Nuclear Plant (SQN), Unit 1. The proposed amendment would revise the Technical Specifications (TSs) to allow a one-time, 31-day extension to monthly TS Surveillance Requirements (SRs) 4.3.1.1.1 and 4.3.2.1.1 for certain instrument functions of the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS), respectively.

TVA requests an expedited review of the proposed change in order to preclude the shutdown of SQN, Unit 1. Given the current electrical grid condition explained below, performing the SRs referenced above in accordance with the current TSs could pose undue risks to the grid due to an inadvertent shutdown of SQN, Unit 1. Current electrical grid conditions combined with an inadvertent shutdown of SQN, Unit 1, could result in an unstable electrical grid condition. Therefore, TVA desires to extend the SRs referenced. Currently, if these SRs are not performed and this proposed TS change is not approved by May 13, 2011, the SQN, Unit 1, TS will require the associated instrument Functions to be declared inoperable and the applicable TS actions taken. These TS actions will ultimately require a shutdown of the unit.

The current electrical grid condition resulted from the severe storms and tornados in the region on April 27, 2011, and could not have been avoided.

2.0 DETAILED DESCRIPTION 2.1 Description of Emergency Situation On April 27, 2011, the Tennessee Valley Region experienced severe weather that included multiple tornados. The severe weather resulted in unprecedented damage to the TVA transmission system, and, coupled with several power plants being offline, an unstable grid condition. TVA lost approximately 90 bulk transmission lines, and as of May 3, 2011, over 250 transmission line towers remain damaged and 34 transmission lines remain out of service. Although TVA has responded quickly and is actively working to restore the transmission system, it is expected that the widespread damage to the transmission system will ultimately take multiple weeks to restore.

Given the current grid condition, TVA's Power System Operations organization has determined that the loss of generation from SQN, Unit 1, could result in a transmission emergency that would force action up to and including load shed in order to protect the grid in North Central Alabama and Southeast Tennessee.

Additionally, given the current condition of the transmission system in Northern Alabama, the loss of generation from SQN, Unit 1, could affect the availability of off-site power being supplied to Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3. All three units at BFN tripped on April 27, 2011, due to an almost complete loss of off-site power as well as the loss of most if not all of the 500kV transmission system that BFN supplies. As a result, BFN declared a Notification of Unusual Event emergency on April, 27, 2011. Electrical power for BFN was supplied by the on-site Emergency Diesel Generators until BFN exited this emergency condition on May 2, 2011 after a second qualified source of power to the plant was restored. All three units at BFN remain shutdown until a sufficient portion of the 500 kV grid in that region is restored.

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As supporting documentation associated with the current grid condition, Attachment 3 contains TVA Power System Operation Transmission Alerts that provide information relative to grid conditions since the severe weather and subsequent transmission system damage that occurred on April 27, 2011.

TVA has requested an expedited review of the proposed change in order to preclude the shutdown of SQN, Unit 1, and subsequent emergency condition on the grid.

Failure to act in a timely manner would result in the shutdown of SQN, Unit 1.

This emergency situation could not have been avoided. The amount of transmission system damage that occurred was unprecedented and therefore could not have been anticipated or prepared for given the severity of the weather conditions experienced in the Tennessee Valley Region.

Additionally, the surveillance for which TVA is requesting a one-time extension was scheduled to be completed on April 28, 2011. TVA did not complete the required testing associated with the surveillance by the scheduled date of April 28, 2011, due to the degraded grid condition that initially developed on April 27, 2011.

2.2 Proposed Technical Specification Changes In accordance with TS SRs 4.3.1.1.1 and 4.3.2.1.1, each train or logic channel for RTS instrumentation and ESFAS instrumentation shall be demonstrated Operable by the performance of the Channel Check, Channel Calibration, and Channel Functional Test operations for the Modes and at the frequencies shown in Tables 4.3-1 and 4.3-2 of the TSs. Table 4.3-1 requires a Channel Functional Test of RTS Functions 20 (Reactor Trip Breaker), 21 (Automatic Trip Logic), and 23 (Reactor Trip Bypass Breakers) to be performed monthly such that each train or logic channel is tested at least every 62 days on a Staggered Test Basis. Table 4.3-2 requires ESFAS Automatic Actuation Logic Functions 1 .b (Safety Injection and Feedwater Isolation),

2.b (Containment Spray), 3.a.2 (Containment Isolation Phase "A" Isolation from Safety Injection), 3.b.2 (Containment Isolation Phase "B" Isolation from Safety Injection), 3.c.2 (Containment Ventilation Isolation), 4.b (Steam Line Isolation), 5.b (Turbine Trip and Feedwater Isolation), 6.b (Auxiliary Feedwater), and 9.b (Automatic Switchover To Containment Sump) to be performed monthly such that each train or logic channel is tested at least every 62 days on a Staggered Test Basis.

TVA currently performs the Channel Functional Test required by SQN, Unit 1, TS SRs 4.3.1.1.1 and 4.3.2.1.1 for the above listed Functions such that each train or logic channel is tested on a 62-day Staggered Test Frequency as part of Solid State Protection System (SSPS) testing. SQN, Unit 1, Train "B" SSPS testing was scheduled to be completed by April 28, 2011. Due to the severe weather experienced throughout the Tennessee Valley Region and the resulting degraded grid condition, the required surveillance was not completed by the scheduled due date. SQN, Unit 1, is currently operating within the surveillance extension period of 15 days allowed by TS 4.0.2. The surveillance extension will expire on May 13, 2011.

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TVA is requesting a one-time, 31-day extension of the monthly Channel Functional Test specified in TS SRs 4.3.1.1.1 and 4.3.2.1.1 for the above listed Functions.

Implementation of this change extends the required due date for the referenced surveillances from May 13, 2011 to June 13, 2011. The proposed change will require the addition of Footnote 13 to Table 4.3-1 and Footnote 3 to Table 4.3-2. The footnotes will read "The Surveillance interval may be extended on a one-time basis up to 31 days beginning May 13, 2011, and ending June 13, 2011." Footnote 13 of Table 4.3-1 will be applied to Functions 20, 21, and 23. Footnote 3 of Table 4.3-2 will be applied to Functions 1.b, 2.b, 3.a.2, 3.b.2, 3.c.2, 4.b, 5.b, 6.b, and 9.b.

2.3 Justification for the Change TVA is requesting this change in order to preclude a unit shutdown. Performance of these Channel Functional Tests associated with SRs 4.3.1.1.1 and 4.3.2.1.1 is considered to be an activity with significant generation risk. An example of the risk associated with this test is documented in Licensee Event Report 50-328/1990-008-00 for SQN, Unit 2. On April 10, 1990, SQN, Unit 2, was at 100 percent power when a reactor trip occurred during the monthly performance of SSPS Train "B" testing.

The reactor trip resulted from a General Warning alarm on both trains of the SSPS and was caused by surveillance test steps being performed out of sequence.

During testing of the process instrumentation system and nuclear instrumentation

'system bistables, each channel bistable is placed in a trip mode causing one SSPS input relay in Train "A" and one in Train "B" to de-energize. A contact of each relay is connected to a universal logic printed circuit card. This card performs both the reactor trip and monitoring functions. The contact that creates the reactor trip also causes a status lamp and an annunciator on the control board to operate.

During the performance of the SSPS Channel Functional Test, the tested train Reactor Trip Bypass Breaker is racked in for the test generating a half-trip signal.

The half-trip signal generates a General Warning alarm and any condition that results in a half-trip signal being generated on the other train will result in a reactor trip signal being initiated. Performance of this Channel Functional Test without placing the Reactor Trip Bypass Breaker in service will also result in a turbine trip.

The train being tested will have a General Warning. A single train General Warning will not cause a reactor trip, but both trains with a General Warning will cause a reactor trip. A General Warning can be caused by the following:

" Loss of any 15 or 48 volt power supply output within the train,

• A printed circuit card removed or not fully inserted in the train,

• On Logic Test Panel: Input Error Inhibit switch not in "Normal," Multiplexer Test switch not in "Normal," Logic A switch not in "OFF," or Memories switch not in "OFF,"

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• On Output Relay Test Panel: Mode selector switch not in "Operate,"

• Ground Return Fuse blown or removed, and

• Loss of 120 Vac power to slave relays.

Human error is the higher failure risk associated with performing the SSPS Channel Functional Test. Human errors associated with procedure use and adherence and/or operating the wrong equipment are the most probable failures which could cause a reactor trip during the performance of the SSPS Channel Functional Test. Examples of human errors resulting in a reactor trip or inadvertent equipment actuations are as follows.

0 The SSPS logic panel and Master Relays are required to be placed in test to allow the Channel Functional Test to be performed. Failure to place these switches in the test position may cause an undesired condition in the SSPS.

• Failure to hold the Feedwater Isolation Reset handswitches in reset while placing Input Error switch in NORMAL would result in a Feedwater Isolation when Tavg is below the required setpoint (550 degrees Fahrenheit).

• The Input Error Inhibit switches of both Trains "A" and "B" must not be placed in Inhibit simultaneously as a reactor trip will occur.

• Components inside the SSPS cabinets can be accidentally bumped generating a trip signal in the system during performance of the SSPS Channel Functional Test.

Due to the numerous human errors that can contribute to a plant trip during SSPS testing and the fact that a portion of the trip logic is intentionally completed for some portions of the SSPS surveillances, TVA considers it prudent and justifiable to extend the frequency of the monthly Channel Functional Test specified in TS SRs 4.3.1.1.1 and 4.3.2.1.1 for the Functions previously discussed in order to avoid an inadvertent plant shutdown and subsequent transmission system (i.e., grid) emergency.

3.0 TECHNICAL EVALUATION

The RTS automatically keeps the reactor operating within a safe region by shutting down the reactor whenever the limits of the region are approached. The safe operating region is defined by several considerations such as mechanical/hydraulic limitations on equipment, heat transfer phenomena and nuclear phenomena. Therefore, the RTS keeps surveillance on process variables which are directly related to equipment mechanical limitations and variables which directly affect the heat transfer capability of the reactor. Other parameters utilized in the RTS are calculated from various process variables. In any event, whenever a direct process or calculated variable exceeds a setpoint the reactor will be shut down in order to protect against either gross damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the containment.

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The ESFAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the Reactor Coolant System pressure boundary, and to mitigate accidents.

RTS and ESFAS consist of up to four redundant sensors and associated process protection circuitry and two redundant digital logic trains. The process protection circuitry monitors various plant parameters and provides inputs to SSPS. SSPS processes the inputs and generates an output signal to initiate the required safety function. The SSPS performs the decision logic for actuating a reactor trip or engineered safety feature actuation, generates the electrical output signal that will initiate the required safety function (trip or actuation), and provides the status, permissive, and annunciator output signals to the main control room.

Each SSPS train has a built in testing device that can automatically test the decision logic matrix functions and the actuation devices while the unit is at power. When any one train is taken out of service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed. The testing device is semi-automatic to minimize testing time.

The RTS and ESFAS Channel Functional Tests are performed monthly as part of SSPS testing. SQN, Unit 1, SRs 4.3.1.1.1 and 4.3.2.1.1, Functions 20, 21, and 23 of Table 4.3-1 and Functions 1 .b, 2.b, 3.a.2, 3.b.2, 3.c.2, 4.b, 5.b, 6.b, and 9.b of Table 4.3-2 require, in part, a Channel Functional Test to be performed monthly, such that each train or logic channel is tested at least every 62 days on a Staggered Test Basis, to ensure the Operability of RTS and ESFAS. The monthly Channel Functional Tests verify proper operation of the Reactor Trip Breaker, Reactor Trip Bypass Breaker and the Automatic Trip Logic in the RTS.

The monthly Channel Functional Tests also verify proper operation of the ESFAS Automatic Actuation Logic. The specific ESFAS Automatic Actuation Functions verified by the monthly Channel Functional Test are Safety Injection and Feedwater Isolation, Containment Spray, Containment Isolation Phase "A" Isolation from Safety Injection, Containment Isolation Phase "B" Isolation from Safety Injection, Containment Isolation, Steam Line Isolation, Turbine Trip and Feedwater Isolation, Auxiliary Feedwater, and Automatic Switchover to Containment Sump.

TVA has evaluated the proposed change and concluded that it is appropriate under the current circumstances. The proposed change was evaluated using a review of the performance history of the applicable Surveillances, a comparison to the current standard TSs approved by the NRC, and a probabilistic risk analysis (PRA). Additional information on each of these evaluations is provided below.

3.1 Evaluation of Surveillance Test History As part of the evaluation for the proposed change, a review of recent surveillance test history (i.e., test history from the past five years) was performed for SQN, Unit 1, SSPS Trains "A" and "B." For SSPS Train "A," this review evaluated the past performances of the associated Channel Functional Tests for TS SR 4.3.1.1.1 and 4.3.2.1.1 from January 6, 2006, to March 18, 2011. For SSPS Train "B," this review evaluated the past performances of the associated Channel Functional Tests for TS SRs 4.3.1.1.1 and 4.3.2.1.1 from February 10, 2006, to February 25, 2011. No E-6

failures of the SSPS Channel Functional Tests for SRs 4.3.1.1.1 or 4.3.2.1.1 have occurred, based on the review of past SSPS Channel Functional Tests, for the date ranges list above. Based on this review, TVA has concluded that recent surveillance test history supports the proposed TS change in that both SSPS trains have been demonstrated to be highly reliable in their capability to perform their required safety function.

3.2 Comparison to Standard Technical Specifications Requirements As part of the Westinghouse Owners Group Technical Specifications Optimization Program, Westinghouse Topical Report WCAP-15376 documented a risk analysis of RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times. WCAP-15367 recommended increasing surveillance test intervals SSPS Logic Functions from 2 months to 6 months and Reactor Trip Breaker surveillance test intervals from 2 months to 4 months. These changes were accepted by the NRC and have been included in NUREG 1431, "Standard Technical Specifications (STS) -Westinghouse Plants," Revision 3.0.

STS Table 3.3.1-1 and SR 3.3.1.4 requires Trip Actuating Device Operational Tests to be performed for Reactor Trip Breakers (RTBs) and Reactor Trip Bypass Breakers every 62 days on a Staggered Test Basis. Based on the definition of Staggered Test Basis in the STS, Reactor Trip Breakers and Reactor Trip Bypass Breakers are tested every 62 days such that each train is tested every 124 days.

STS Table 3.3.1-1 and SR 3.3.1.5 and Table 3.3.2-1 and SR 3.3.2.2 require Actuation Logic Tests of the RTS Automatic Trip Logic and ESFAS Automatic Actuation Logic Functions, respectively, to be performed every 92 days on a Staggered Test Basis. Based on the definition of Staggered Test Basis in the STS, RTS Automatic Trip Logic, and ESFAS Automatic Actuation Logic Functions are tested every 92 days such that each train is tested every 184 days.

The proposed change for the SQN, Unit 1, Reactor Trip Breakers, Reactor Trip Bypass Breakers, RTS Automatic Trip Logic, and ESFAS Automatic Actuation Logic Functions would allow a one-time 31-day extension to the existing Channel Functional Test Surveillance Frequency. For the Reactor Trip Breakers and Reactor Trip Bypass Breakers, performance of a SQN, Unit 1, TS Channel Functional Test is equivalent to performance of an STS Trip Actuating Device Operational Test. In addition, for the RTS Automatic Trip Logic and the ESFAS Automatic Actuation Logic, performance of a SQN, Unit 1, TS Channel Functional Test is equivalent to performance of an STS Actuation Logic Test. The existing Channel Functional Test Frequency is monthly with each train being tested every 62 days on a Staggered Test Basis. Based on the SQN, Unit 1, definition of Staggered Test Basis, Reactor Trip Breakers, Reactor Trip Bypass Breakers, RTS Automatic Trip Logic, and ESFAS Automatic Actuation Logic Functions are tested every 31 days such that each train is tested every 62 days. The proposed one-time total Surveillance interval for performance of the Channel Functional Tests for each of these Functions is as follows.

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• 62 days + 15 days (extension allowed by SR 4.0.2) + 31 days = 108 days Therefore, the proposed SQN, Unit 1, change is conservative with respect to the Surveillance Frequencies specified in the STS for testing of these Functions, i.e., 108 days as compared to STS Surveillance intervals of 124 days and 184 days.

3.3 Risk Evaluation The following PRA addresses the additional Internal Events risk to SQN for extending the TS-required testing interval for the RTS and SSPS. This section addresses the three-tier approach to the evaluation of risk-informed TS changes described in Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications."

The first tier addresses PRA capability and insights, and includes the risk assessment supporting risk impact. The second tier addresses avoidance of risk-significant plant configurations. The third tier addresses risk-informed plant configuration control and management.

WCAP-1 5376 provides a risk-informed assessment of extending surveillance test intervals for SSPS and the RTBs and is utilized as a reference for this evaluation.

WCAP-1 5376 also covers extending the allowed outage time of the SSPS and RTBs.

However, the assessment presented below only evaluates an extension of the Surveillance Test Intervals (STIs) from 62 days to 108 days.

Tier 1: PRA Capability and Insights Quality of the PRA:

The risk evaluation was completed using the current Draft Revision 5 SQN Computer Aided Fault Tree Analysis (CAFTA) model. The Draft SQN model has been Peer Reviewed by the Pressurized Water Reactor Owners Group and was determined to meet 94% of the American Society of Mechanical Engineers standard requirements at Capability Category II or better. Since the peer review, multiple Findings and Observations have been resolved in the process of developing the Final SQN Revision 5 Model of Record. The Draft SQN CAFTA model is the most accurate representation of the as-built, as-operated plant.

Scope of the PRA:

The ESFAS and the RTBs are explicitly modeled within the Draft SQN PRA model.

The RTBs are represented by basic events for fails to open. The ESFAS model is represented by fault trees for each of the features identified within the PRA model.

The ESFAS modeled features are as follows.

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Actuation FunCtion Parameter: Signal:

Low-Low SG Level (1 out of 4)

Low Pressurizer Pressure Motor Driven Auxiliary Feedwater Hig Containen Pressure

-High Containment Pressure Low Steamline Pressure Low-Low SG Level (2 out of 4)

Low Pressurizer Pressure Turbine Driven Auxiliary Feedwater ContainenPressure H Containment High Pressure Low Steamline Pressure Low Pressurizer Pressure Safety Injection High Containment Pressure Low Steamline Pressure Low Pressurizer Pressure Containment Vent Isolation High Containment Pressure Low Steamline Pressure Low Pressurizer Pressure Phase A Containment Isolation High Containment Pressure Low Steamline Pressure Low Pressurizer Pressure Main Feedwater Isolation High Containment Pressure Low Steamline Pressure Main Steam Isolation High-High Containment Pressure Phase B Containment Isolation High-High Containment Pressure Containment Spray High-High Containment Pressure Refueling Water Storage Tank Containment Sump High Level (RWST) Auto Swapover RWST Low Level Chanties to the PRA Model.-

1. Changes to Failure Rate Probabilities:

For the purposes of this analysis, four different failure rates were increased, the RTBs failing to open, the ESFAS relays failing to operate, undervoltage drive card failure to operate, and universal card failure to operate. In accordance with WCAP-1 5376-P-A, Revision 1, pages 8-19, the following formula was used to calculate the new values for the failure rates:

,'extended STI*

tent STI FP(extended STI) = FP(current STI) x currentSTI Where:

FP(extendedSTI) = The new failure probability based on having the surveillance test interval increased E-9

FP(currentSTI) = The failure probability currently used in the PRA model, based on the existing surveillance test interval extended STI = The purposed time frame for the new surveillance test interval currentSTI = The current time frame for the surveillance test interval Using the above equation with an extended STI time of 108 days, and a current STI time of 62 days, the change in failure rates is presented in the table below.

Origin~al. Evaluation Type Code-Type__Code_ DescriptionOrgnlEautn Description _ Failure Rate Failure Rate RLYFD Relay Fails to Operate 2.480E-05 4.340E-05 RTBFO Reactor Trip Breaker 1.550E-05 2.71 OE-05 Fails to Open UVDFR Undervoltage Drive Card 3.400E-04 5.950E-04 Fails to Operate Universal Card UNLFR Fai Fails to erat Operate 1 3.800E-05 6.650E-05

2. Changes to the Unavailability Model:

There were no changes made to the test and maintenance events in the PRA model for the RTBs or the SSPS trains.

3. Changes to the Common Cause Failure Probabilities:

The common cause failure probabilities were updated to reflect the new failure probabilities presented in the table above. The CAFTA Common Cause analysis tool updated the failure rates based on the alpha factors used in current PRA model.

Modeling of Initiating Events:

The ESFAS and RTB models are not modeled as support system initiating events (modeled as a fault tree) as failure of these components is subsumed into the reactor trip frequency.

Truncation Limits:

The Core Damage Frequency (CDF) is truncated at 1E-1 1 and the Large Early Release Frequency (LERF) is truncated at 5E-13. To verify the basic events, a truncation analysis on the component importance was performed by reviewing the E-10

basic events changed in this analysis, their Fussel-Vessley per decade, and their Risk Achievement Worth (RAW) per decade. The review of the basic event importance measures verified that those basic events impacted by the change in probabilities appeared in the cutsets, and that their respective impact was not being underestimated. Of these basic events that only appear in the later truncation levels, all have RAW values greater than 1.0, therefore the impact of these events is captured at the 1 E-1 1 truncation for CDF and 5E-1 3 for LERF.

Flood:

The internal flooding model does not need to be changed for this evaluation. Any flood event which would impact the ESFAS cabinets or the RTBs would affect both trains at the same time. Therefore no changes were made to the internal flooding model.

Fire:

The auxiliary instrument rooms at SQN, where the SSPS cabinets are located, are not areas that would be considered vulnerable to fire. These plant areas do not contain pumps, motors, breakers, oil sources or oil storage. These areas have full cross zoned ionization smoke and thermal detection, total flooding CO 2 suppression systems and a portable halon extinguisher. The control rod drive equipment rooms, where the RTBs are located, have complete area protection by automatic pre-action sprinkler system, cross-zoned ionization smoke detection, two dry chemical portable extinguishers and a hose station. The safe shutdown strategy for 10 CFR 50, Appendix R-type fires at SQN does not take credit for automatic actuation or repositioning of equipment by the SSPS and RTS.

Results:

The average test and maintenance model was used to determine the incremental conditional core damage probability and incremental conditional large early release probability for the one-time TS SR extension. RG 1.177 states that Incremental Conditional Core Damage Probability (ICCDP) of less than 5.OE-07 is considered small for a single TS STI change. An Incremental Conditional Large Early Release Probability (ICLERP) of 5.OE-08 or less is also considered small. The acceptability of increasing the STI for the reactor trip instrumentation Channel Functional Test for SQN, Unit 1, has been quantitatively evaluated using the SQN CAFTA PRA Model.

The SQN PRA organization has determined that the proposed 31-day extension (modeled as 46 days, i.e., the 15 days allowed by SR 4.0.2 plus the proposed 31-day extension) and a 108-day STI time are justified. Based on the results for ICCDP and ICLERP against RG 1.177, the calculated values determined for this analysis were less than the RG 1.177 values. A summary of the results are provided in the table below.

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.I~nc eental.

Rk,M tric PRA Result Adjusted ACDF Time Conditional PRA. Result Window VAlue Core Damage 1.21 E-05 1.26E-05 5.OOE-07 46 Days 6.30E-08 Large Early Release 1.63E-06 1.92E-06 2.90E-07 46 Days 3.65E-08 PRA Analysis Assumptions:

1. The test and maintenance activities in the time frame, the additional 31 days, are assumed to not be known, therefore the average test and maintenance model is used to calculate the ICCDP and ICLERP.

2. The mean failure rates for the components affected by the extension of the Surveillance interval are conservatively assumed to be increased by the ratio of the new time interval to the existing time interval between Surveillances.

Uncertaintiesand Sensitivities:

Assumption 1 is a completeness uncertainty based on the unavailability of all plant states during the time interval. To address the uncertainty associated with plant operating configurations, a sensitivity analysis was run to show what the risk metric increase would be given no testing or maintenance activities would be performed during the extension time frame. These values, presented in the table below, meet the thresholds outlined in Section 2.4 of Regulatory Guide 1.177.

PR-A Adj~usted Tim.e Incremental RiskAMetric ACDF Conditional Result PRA Result Window Value Core Damage 1.01 E-05 1.05E-05 4.OOE-07 46 Days 5.04E-08 Large Early Release 1.23E-06 1.50E-06 2.70E-07 46 Days 3.40E-08 The second assumption, the calculation of mean failure rate increase, used the methodology in WCAP-15376-P-A. This methodology represents a consensus model used by the members of the PWROG. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," states that any model uncertainty that is part of a consensus model may be screened from sensitivity analysis.

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Tier 2: Avoidance of Risk-Significant Plant Configurations The PRA evaluation did not exclude maintenance on any other plant equipment during this evaluation. Therefore, no maintenance activities are specifically excluded by the Tier 1 PRA evaluation to avoid any risk-significant plant configurations. SQN's site-specific configuration risk management program (CRMP) evaluated below in Tier 3 will be utilized to prevent any risk-significant plant configurations during the period of the proposed TS change.

Tier 3: Risk-Informed Configuration Risk Management SQN work management programs have processes in place to assess risk for equipment outages, both planned and unplanned. In particular, NPG-SPP-07.3, "Work Activity Risk Management Process," is used for all maintenance. Procedure NPG-SPP-07.1, "On Line Work Management," details the risk assessment guidelines utilizing the results of the site PRA and are described in site-specific Technical Instructions (TIs). Other safety considerations, such as TSs and the scope of risk significant Structures, Systems, and Components defined in the site specific Maintenance Rule TI that are not modeled in the PRA, are included in the site specific scheduling risk management program. These programs are utilized to determine which system, component, and Functional Equipment Group (FEG) combinations may be worked.

In general, risk is evaluated in accordance with SQN Procedure 0-TI-DSM-000-007. 1, "Risk Assessment Guidelines," and based upon the SQN PRA. Other safety considerations, such as TSs, maintenance rule risk significant SSCs, and significant changes in weather or offsite power availability, are considered in the site-specific CRMP and are used to determine which system, component, and FEG combinations may be worked online. In addition, an assessment of scheduled activities is performed before implementation of a work window. The assessment includes reviews for the following.

" The schedule is evaluated against the risk bases outlined in the SQN PRA.

• Maximizing safety (reducing risk) when performing online work.

" Avoidance of recurrent entry into a specific TS limiting condition for operation (LCO) for multiple activities. Activities that require entering the same LCO are combined to limit the number of times an LCO must be established, thus maximizing the equipment's availability.

• If the risk associated with a particular activity cannot be determined, site engineering is requested to perform a risk assessment.

• Implementing compensatory measures and requirements for management authorization for higher risk configurations.

E-13

The SQN CRMP described above ensures that risk-significant plant configurations will not be entered and appropriate actions will be taken when unforeseen events put the plant in a risk-significant configuration.

The PRA quantitative evaluation has concluded that this change meets the NRC criteria for having a small quantitative impact to plant risk. SQN's risk-informed CRMP will ensure that risk significant plant configurations will be properly assessed and avoided during the period of the extended surveillance interval requested by this TS change.

3.4 Conclusion WVA has evaluated the proposed TS change and concluded that it is appropriate under the current circumstances. The PRA quantitative evaluation has concluded that this change meets the criteria for having a small quantitative impact to plant risk and the results for ICCDP and ICLERP were less than the values prescribed in RG 1.177 for a TS change. The proposed change is also conservative with respect to the Surveillance Frequencies specified in the STS for testing of these instrument Functions. Additionally, an evaluation of recent testing history has been performed and demonstrates that both SSPS trains remain capable of performing their required safety function. If at any time prior to the expiration of the current grace period for the affected SRs the electrical grid is determined to be sufficiently stable, TVA will withdraw this request for an emergency TS change.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The regulatory bases and guidance documents associated with the systems discussed in this license amendment request are the following:

General Design Criteria (GDC) 2: Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety function. The design bases for these structures, systems, and components shall reflect:

1. Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated,

2. Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena, and

3. The importance of the safety functions to be performed.

E-14

GDC 13: Instrumentation and control shall be provided to monitor variables and systems over their anticipated ranges for normal operation for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 20: The Protection System shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC 21: The Protection System shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.

Redundancy and independence designed into the Protection System shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the Protection System can be otherwise demonstrated. The Protection System shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

GDC 22: The Protection System shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.

Design techniques, such as function diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

GDC 23: The Protection System shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

GDC 24: The Protection System shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the Protection System.

Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

GDC 25: The Protection System shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the E-15

Reactivity Control Systems, such as accidental withdrawal (not ejection or dropout) of control rods.

GDC 27: The Reactivity Control Systems shall be designed to have a combined capability, in conjunction with poison addition by the Emergency Core Cooling System, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

GDC 29: The Protection and Reactivity Control Systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

Regulatory Guide 1.22 discusses an acceptable method of satisfying GDC 20 and GDC 21 regarding the periodic testing of protection system actuation functions.

These periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident.

10 CFR 50.55a(h) requires that the protection systems meet IEEE 279-1971.

Section 4.2 of IEEE 279-1971 discusses the general functional requirement for protection systems to assure they satisfy the single failure criterion.

There will be no changes to the RTS or ESFAS instrumentation design such that compliance with any of the regulatory requirements and guidance documents above would come into question. As such, SQN, Unit 1, will continue to comply with all applicable regulatory requirements.

4.2 Precedent TVA identified the following documents related to similar permanent RTS and ESFAS Surveillance Frequency changes.

• A Safety Evaluation was issued by the NRC by letter dated December 20, 2002, approving WCAP-15376.

• Industry/Technical Specification Task Force (TSTF) STS Change Traveler 411 was developed to reflect the implementation of WCAP-1 5376 in the STS.

TSTF Traveler 411 was approved by the NRC by letter dated August 30, 2002.

" Southern Nuclear Operating Company submitted a License Amendment Request on January 27, 2005 for Vogtle Electric Generating Plant (VEGP),

Units 1 and 2, to adopt the relaxations that were generically approved in WCAP-1 5376-P-A, Revision 1. Amendments 145 and 125 were issued for VEGP approving the changes proposed in WCAP-1 5376-P-A, Revision 1.

E-16

Additionally, while only one specific precedent is cited here, due to the incorporation of TSTF-411 into NUREG -1431, several licensees have received approval of the permanent RTS and ESFAS Surveillance Frequency changes.

4.3 Significant Hazards Consideration The Tennessee Valley Authority (TVA) has concluded that the change to Facility Operating License No. DPR-77 for Sequoyah Nuclear Plant (SQN), Unit 1, to allow a one-time, 31-day extension to monthly Technical Specification (TS) Surveillance Requirements (SRs) 4.3.1.1.1 and 4.3.2.1.1 for certain instrument functions of the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS), respectively, does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation in accordance with 10 CFR 50.91(a)(1) of the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

The proposed changes do not result in any modifications to RTS or ESFAS hardware, design requirements, or functions. No system operational parameters are affected. The protection system will continue to perform the intended design functions consistent with the design bases and accident analyses. The proposed changes will not modify any system interfaces and, therefore, could not increase the likelihood of an accident described in the Updated Final Safety Analysis Report (UFSAR). The proposed amendment will not change, degrade or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR.

The proposed TS SR extension reduces the potential for inadvertent reactor trips and spurious actuations and, therefore, does not increase the probability of any accident previously evaluated. The proposed changes to the TSs do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the RTS and ESFAS signals. The RTS and ESFAS will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This change meets the acceptance criteria in Regulatory Guide 1.177. Therefore, since the RTS and ESFAS will continue to perform their functions with high reliability as originally assumed, and the increase in risk as measured by Core Damage Frequency, Large Early Release Frequency, Incremental Conditional Core Damage Probability, and Incremental Conditional Large Early Release Probability is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components from E-17

performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated.

Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes are consistent with the safety analysis assumptions and resultant consequences.

Therefore, this proposed change does not increase the probability or consequences of an accident previously evaluated.

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

The proposed amendment does not require any design changes, physical modifications or changes in normal operation of the RTS and ESFAS instrumentation. Existing setpoints will be maintained. The changes do not affect functional performance requirements of the instrumentation. No changes are required to accident analysis assumptions. The changes do not introduce different malfunctions, failure modes, or limiting single failures. The one-time change to the surveillance frequency does not change any existing accident scenarios nor create any new or different accident scenarios.

Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes.

Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. The signals credited in the accident analyses and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis. A PRA quantitative evaluation has concluded that the proposed change has a small quantitative impact on plant risk and meets the acceptance criteria contained in Regulatory Guide 1.177.

Therefore, it is concluded that this proposed change does not involve a significant reduction in the margin of safety.

Based on the above, TVA concludes that the proposed amendment does not involves a significant hazards consideration for SQN, Unit 1, under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

E-18

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

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. WCAP-1 5367-P-A, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," Revision 1, March 2003.

2. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," August 1998.

3. NUREG-1 855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2009.

4. NUREG 1431 "Standard Technical Specifications (STS) -Westinghouse Plants,"

Revision 3.0, June 2004 E-19

ATTACHMENT I Proposed Technical Specification Changes (Mark-Ups)

Technical Specification Pages 3/4 3-12 3/4 3-13 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37a 3/4 3-38

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED

15. Deleted

16. Undervoltage - Reactor Coolant N.A. R Q 1 Pumps

17. Underfrequency - Reactor Coolant N.A. R Q 1 Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. (1) (12)

B. Turbine Stop Valve Closure N.A. N.A. (1) (12) 1"**

19. Safety Injection Input from ESF N.A. N.A. R 1,2

20. Reactor Trip Breaker N.A. N.A. 1, 2, and *

21. Automatic Trip Logic N.A. N.A.

22. Reactor Trip System Interlocks A. Intermediate Range Neutron N.A. R N.A. 2, and

• Flux, P-6 B. Power Range Neutron Flux, N.A. N.A. N.A. 1 P-7 C. Power Range Neutron Flux, N.A. R N.A.

P-8 D. Power Range Neutron Flux, N.A. R N.A. 1,2 P-1 0 N.A. 1 E. Turbine Impulse Chamber N.A. R Pressure, P-13 N.A. 1 F. Power Range Neutron Flux, N.A. R P-9 G. Reactor Trip, P-4 N.A. N.A. ,and*

S2,

23. Reactor Trip Bypass Breaker N.A. N.A. M(11O)R(1 1) 1, 2, and

• SEQUOYAH - UNIT 1 3/4 3-12 Amendment No. 54, 63, 141, 304, 318

TABLE 4.3-1 (Continued)

NOTATION

• - With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

• • - Above the P-9 (Power Range Neutron Flux) interlock.

(1) - If not performed in previous 31 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) - Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate ifthe absolute difference greater than or equal to 3 percent. The frequency of this surveillance is every 31 EFPD. This surveillance is not required to be performed until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after thermal power is > 15% RTP.

(4) - Deleted.

(5) - Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8) - Deleted.

(9) - The CHANNEL FUNCTIONAL TEST shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function.

(10) - Local manual shunt trip prior to placing breaker in service. Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(11) - Automatic and manual undervoltage trip.

(12) - Prior to exceeding the P-9 interlock whenever the unit has been in HOT STANDBY.

I (13) - The Surveillance interval may be extended on a one-time basis up to 31 days beginning May 13, 2011 and ending June 13, 2011.

ADD SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114, 141,199, 304, 318

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. SAFETY INJECTION AND FEEDWATER ISOLATION

a. Manual Initiation N.A. N.A. R 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. 1,2,3,4

c. Containment Pressure-High S R Q 1,2,3

d. Pressurizer Pressure--Low S R Q 1,2,3

e. Deleted

f. Steam Line Pressure-Low S R Q 1,2,3

2. CONTAINMENT SPRAY

a. Manual Initiation N.A. N.A. R 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M(1J 1,2,3,4

c. Containment Pressure--High-High S R 1,2, May ,16, 0Q SEQUOYAH - UNIT 1 3/4 3-34 Amendment Nos. 47,141

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

3) CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual N.A. N.A. R 1,2,3,4

2) From Safety Injection N.A. N.A. M(1) 1,2,3,4 Automatic Actuation Logic

b. Phase "B" Isolation

1) Manual N.A. N.A. R 1,2,3,4

2) Automatic Actuation Logic N.A. N.A. M(1) 1,2,3,4

3) Containment Pressure-- High- S R Q 1,2,3 High

c. Containment Ventilation Isolation

1) Manual N.A. N.A. R 1,2,3,4

2) Automatic Isolation Logic N.A. N.A. 1,2,3,4

3) Containment Purge Air S R 1,2,3,4 Exhaust Monitor Radio-activity-High Marh 14, 1996 SEQUOYAH - UNIT 1 3/4 3-35 Amendment No. 47,168, 220

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

4. STEAM LINE ISOLATION

a. Manual N.A. N.A. R 1,2,3

b. Automatic Actuation Logic N.A. N.A. M(1)i 1,2,3

c. Containment Pressure-- S R Q 1,2,3 High-High

d. Steam Line Pressure--Low S R Q 1,2,3

e. Negative Steam Line S R Q 3 Pressure Rate--High

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water S R Q 1,2,3 Level--High-High

b. Automatic Actuation N.A. N.A. M(1 1,2,3 Logic

6. AUXILIARY FEEDWATER

a. Manual N.A. N.A. R 1,2,3

b. Automatic Actuation N.A. N.A. M(1 1,2,3 Logic June 26, 1993 SEQUOYAH - UNIT 1 3/4 3-36 Amendment Nos. 47, 63, 141, 168

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

8. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS

a. Pressurizer Pressure, P-1 1/Not N.A. R(2) N.A. 1,2,3 P-11

b. Deleted

c. Steam Generator Level, P-14 N.A. R(2) N.A. 1,2

9. AUTOMATIC SWITCHOVER TO CONTAINMENT SUMP

a. RSWT Level - Low S R Q 1,2,3,4 COINCIDENT WITH Containment Sump Level - S R Q 1,2,3,4 High AND Safety Injection (See 1 above for all Safety Injection Surveillance Requirements)

b. Automatic Actuation N.A. N.A. 1,2,3,4 Logic Septomber 14, 2006 SEQUOYAH - UNIT 1 3/4 3-37a Amendment No. 47, 63, 129, 141,182, 188, 207, 311

TABLE 4 .3-2 (Continued)

TABLE NOTATION (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(2) The total interlock function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

(3) The Surveillance interval may be extended on a one-time basis up to 31 days beginning May 13, 2011 and ending June 13, 2011. 13 ADD Septcmbcr 14, 2006 SEQUOYAH - UNIT 1 3/4 3-38 Amendment No. 47, 182, 188, 207, 311

ATTACHMENT 2 Proposed Technical Specification Changes (Final Typed)

Technical Specification Pages 3/4 3-12 3/4 3-13 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37a 3/4 3-38

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED

15. Deleted

16. Undervoltage - Reactor Coolant N.A. R Q Pumps

17. Underfrequency - Reactor Coolant N.A. R Q Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. (1) (12) 1"**

B. Turbine Stop Valve Closure N.A. N.A. (1) (12)

19. Safety Injection Input from ESF N.A. N.A. R 1,2

20. Reactor Trip Breaker N.A. N.A. M(5) (13) and 1, 2, and

• S/U(1)

21. Automatic Trip Logic N.A. N.A. M(5) (13) 1, 2, and *

22. Reactor Trip System Interlocks A. Intermediate Range Neutron N.A. R N.A. 2, and

• Flux, P-6 B. Power Range Neutron Flux, N.A. N.A. N.A. 1 P-7 C. Power Range Neutron Flux, N.A. R N.A. 1 P-8 D. Power Range Neutron Flux, N.A. R N.A. 1,2 P-10 E. Turbine Impulse Chamber N.A. R N.A. 1 Pressure, P-13 F. Power Range Neutron Flux, N.A. R N.A. 1 P-9 G. Reactor Trip, P-4 N.A. N.A. R 1,2, and *

23. Reactor Trip Bypass Breaker N.A. N.A. M(10) (13) 1, 2, and

• R(1 1)

Affpri 2, 209 SEQUOYAH - UNIT 1 3/4 3-12 Amendment No. 54, 63, 141, 304, 318,

TABLE 4.3-1 (Continued)

NOTATION

• - With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

• • - Above the P-9 (Power Range Neutron Flux) interlock.

(1) - If not performed in previous 31 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) - Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate ifthe absolute difference greater than or equal to 3 percent. The frequency of this surveillance is every 31 EFPD. This surveillance is not required to be performed until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after thermal power is > 15% RTP.

(4) - Deleted.

(5) - Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8) - Deleted.

(9) - The CHANNEL FUNCTIONAL TEST shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function.

(10) - Local manual shunt trip prior to placing breaker in service. Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(11) - Automatic and manual undervoltage trip.

(12) - Prior to exceeding the P-9 interlock whenever the unit has been in HOT STANDBY.

(13) - The Surveillance interval may be extended on a one-time basis up to 31 days beginning May 13, 2011 and ending June 13, 2011.

ApFil 2,2008 SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114, 141,199, 304, 318,

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. SAFETY INJECTION AND FEEDWATER ISOLATION

a. Manual Initiation N.A. N.A. R 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M(1) (3) 1,2,3,4

c. Containment Pressure-High S R Q 1,2,3

d. Pressurizer Pressure--Low S R Q 1,2,3

e. Deleted

f. Steam Line Pressure-Low S R Q 1,2,3

2. CONTAINMENT SPRAY

a. Manual Initiation N.A. N.A. R 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M(1) (3) 1,2,3,4

c. Containment Pressure--High-High S R Q 1,2, May41499, N

SEQUOYAH - UNIT 1 3/4 3-34 Amendment Nos. 47,141, -_

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

3) CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual N.A. N.A. R 1,2,3,4

2) From Safety Injection N.A. N.A. M(1) (3) 1,2,3,4 Automatic Actuation Logic

b. Phase "B" Isolation N.A. N.A. R 1,2,3,4

1) Manual

2) Automatic Actuation Logic N.A. N.A. M(1) (3) 1,2,3,4 I

3) Containment Pressure-- High- S R Q 1,2,3 High

c. Containment Ventilation Isolation

1) Manual N.A. N.A. R 1,2,3,4

2) Automatic Isolation Logic N.A. N.A. M(1) (3) 1,2,3,4

3) Containment Purge Air S R Q 1,2,3,4 Exhaust Monitor Radio-activity-High March 4, 1906 SEQUOYAH - UNIT 1 3/4 3-35 Amendment No. 47, 168, 220,

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

4. STEAM LINE ISOLATION

a. Manual N.A. N.A. R 1,2,3

b. Automatic Actuation Logic N.A. N.A. M(1) (3) 1,2,3

c. Containment Pressure-- S R Q 1,2,3 High-High

d. Steam Line Pressure--Low S R Q 1,2,3

e. Negative Steam Line S R Q 3 Pressure Rate--High

5. TURBINETRIPAND FEEDWATER ISOLATION S R Q 1,2,3

a. Steam Generator Water Level--High-High

b. Automatic Actuation N.A. N.A. M(1) (3) 1,2,3 Logic

6. AUXILIARY FEEDWATER N.A. N.A. R 1,2,3

a. Manual

b. Automatic Actuation N.A. N.A. M(1) (3) 1,2,3 Logic Juno 25, 1993 SEQUOYAH - UNIT 1 3/4 3-36 Amendment Nos. 47, 63, 141, 168, __

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

8. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS

a. Pressurizer Pressure, P-11/Not N.A. R(2) N.A. 1,2,3 P-11

b. Deleted

c. Steam Generator Level, P-14 N.A. R(2) N.A. 1,2

9. AUTOMATIC SWITCHOVER TO CONTAINMENT SUMP

a. RSWT Level - Low S R Q 1,2,3,4 COINCIDENT WITH Containment Sump Level - S R Q 1,2,3,4 High AND Safety Injection (See 1 above for all Safety Injection Surveillance Requirements)

b. Automatic Actuation N.A. N.A. M(1) (3) 1,2,3,4 Logic S.ptr F1A4, 2006 SEQUOYAH - UNIT 1 3/4 3-37a Amendment No. 47, 63, 129, 141,182, 188,207, 311, __

TABLE 4.3-2 (Continued)

TABLE NOTATION (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(2) The total interlock function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

(3) The Surveillance interval may be extended on a one-time basis up to 31 days beginning May 13, 2011 and ending June 13, 2011.

Scptcmbcr 14, 2006 SEQUOYAH - UNIT 1 3/4 3-38 Amendment No. 47, 182, 188, 207, 311, __

ATTACHMENT 3 Power System Operations Transmission Alerts April 27, 2011 Through May 6, 2011 A3-1

PSO/Transmission & Reliability Transmission Alert 1 - Large Area System Damage Wednesday, April 27, 2011 - 1000 CPT

• A severe weather system has damaged the TVA transmission system in Alabama and Mississippi.

" A total of nine 161 kV lines and six customer delivery points are currently out of service.

• The bulk electric system is secure and stable at this time.

• Severe weather will continue today as waves of intense thunderstorms track across the Valley. Additional tornadoes and damaging winds are expected.

• There is a high risk of severe weather, including tornadoes in north Alabama and adjacent areas of middle Tennessee, northeast Mississippi, and northwest Georgia. A moderate risk of severe weather covers much of the rest of TVA territory.

" This afternoon additional severe thunderstorms will follow behind the morning round of storms. Afternoon and early evening thunderstorms pose a risk of strong long-track tornadoes, especially in the high risk area.

" The storm will exit the service territory late tonight ending the threat.

" All non-essential transmission and generation maintenance activities are suspended until further notice.

" The TEOC will be activated in Support Mode at 1200 CPT today.

TransmissionAlert I - Large Area System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damagedor threatened reliabilityof TVA 's power supply over a large area with multiple outages to TVA or distributorsystems.

• Impacted area involves several Transmission Service Centers and/orCustomer Service Centers.

• All nonessentialpower switching operations are suspended.

• Area Emergency Team is activated.

" Transmission Emergency Operations Center(TEOC) is activated and staffed as appropriatein a Support or Control mode as determined by the Transmission Emergency Team Leader.

A3-2

PSO/Transmission & Reliability Transmission Alert 1 - Large Area System Damage Wednesday, April 27, 2011 - 1500 CPT

• The Transmission Emergency Operations Center remains activated in Support Mode.

• A total of twelve 161-kV lines remain out of service due to impacts from severe weather.

• The TVA service that remains out at this time is a portion of Cullman EC, East Mississippi EPA, and Philadelphia Utilities loads.

• Damage assessments continue in the impacted areas and the restoration activities are underway where conditions permit.

• The bulk electric system is secure and stable at this time.

TransmissionAlert 1 - Large Area System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged or threatened reliabilityof TVA's power supply over a large area with multiple outages to TVA or distributorsystems.

• Impacted area involves several Transmission Service Centers and/orCustomer Service Centers.

• All nonessentialpower switching operationsare suspended.

• Area Emergency Team is activated.

• TransmissionEmergency Operations Center(TEOC) is activated and staffed as appropriate in a Support or Control mode as determined by the Transmission Emergency Team Leader.

A3-3

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Thursday, April 28, 2011 - 0800 CPT

• A major tornado outbreak on Wednesday in the service territory has caused extensive damage to the TVA transmission grid.

" Over 90 bulk transmission elements were impacted by the storm and over 75 remain out of service.

• Crews are assessing damage, and restoration activities are in progress.

• All non-essential transmission and generation maintenance activities are suspended until further notice.

" Additional updates will be provided as conditions warrant.

TransmissionAlert 2 - Widespread System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damagedreliabilityof TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

" Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

• All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activated in a Controlmode and fully staffed.

A3-4

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Thursday, April 28, 2011 - 1700 CPT

" The Transmission Emergency Operations Center remains activated in Control mode.

• A total of seven transmission lines have been returned to service today.

• Damage assessments continue in the impacted areas, and the restoration activities are underway.

" The bulk electric system is stable at this time.

• All non-essential transmission and generation maintenance activities are suspended until further notice.

TransmissionAlert 2 - Widespread System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged reliabilityof TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

" Impacted areainvolves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or CustomerService Centers.

• All nonemergency operationsare terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency Operations Center(TEOC) is activated in a Control mode and fully staffed.

A3-5

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Friday, April 29, 2011 - 1000 CPT

" The Transmission Emergency Operations Center remains activated in Control mode.

" A total of nine additional transmission lines were returned to service overnight.

" A total of 70 bulk transmission lines remain out of service.

• Damage assessments continue in the impacted areas, andrestoration activities are underway.

" The bulk electric system is stable at this time.

" All bulk power system switching activities are suspended.

• Critical generator sensitive activities will be scheduled by TRO as system conditions allow.

TransmissionAlert 2 - Widespread System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damagedreliability of TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

" Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

• All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activated in a Control mode and fully staffed.

A3-6

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Saturday, April 30, 2011 - 1100 CPT

" The Transmission Emergency Operations Center remains activated in Control Mode.

" A total of six additional transmission lines and line sections were returned to service overnight.

" Damage assessments continue in the impacted areas and the restoration activities are underway.

• The bulk electric system is secure and stable at this time.

• All bulk power system switching activities are suspended.

• Critical generator sensitive activities will scheduled by TRO as system conditions allow.

TransmissionAlert 2 - Widespread System Damage

" Unusual transmissionsystem conditions (weather,power supply, or transmission problems) have damaged reliability of TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

" Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

" All nonemergency operationsare terminated.

• The CustomerResources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency Operations Center(TEOC) is activated in a Control mode and fully staffed.

A3-7

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Sunday, May 1, 2011 - 0900 CPT

• The Transmission Emergency Operations Center remains activated in Control Mode.

• A total of five additional transmission lines and line sections were returned to service overnight.

• A total of 55 transmission lines and line sections remain out of service.

• Damage assessments continue in the impacted areas and the restoration activities are underway.

• The bulk electric system is secure and stable at this time.

• All bulk power system switching activities are suspended.

• Critical generator sensitive activities will scheduled by TRO as system conditions allow.

TransmissionAlert 2 - WidespreadSystem Damage

" Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damagedreliabilityof TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

• Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

• All nonemergency operationsare terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activated in a Control mode and fully staffed.

A3-8

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Monday, May 2, 2011 - 1000 CPT

• The Transmission Emergency Operations Center remains activated in Control Mode.

• A total of six additional transmission lines and line sections were returned to service overnight.

• A total of 49 transmission lines and line sections remain out of service.

" Damage assessments continue in the impacted areas, and restoration activities are underway.

" The bulk electric system is secure and stable at this time.

• All bulk power system switching activities are suspended.

" Critical generator sensitive activities will be scheduled by TRO as system conditions allow.

TransmissionAlert 2 - WidespreadSystem Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged reliabilityof TVA's power supply over a widespreadarea with multiple outages to TVA or distributorsystems.

• Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

• All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• TransmissionEmergency Operations Center(TEOC) is activated in a Control mode and fully staffed.

A3-9

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Tuesday, May 3, 2011 - 0900 CPT

• The Transmission Emergency Operations Center remains activated in Control Mode.

• A total of 43 transmission lines and line sections remain out of service.

" Damage assessments continue, and restoration activities are underway.

• The bulk electric system is stable at this time.

" We continue to incrementally restore load in north central Alabama as transmission elements are returned to service.

" All non-restoration related bulk power system switching activities are suspended.

• Critical generator sensitive activities will be scheduled by TRO as system conditions allow.

Transmission Alert 2 - Widespread System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged reliabilityof TVA's power supply over a widespread area with multiple outages to TVA or distributorsystems.

• Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

• All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activated in a Control mode and fully staffed.

A3-10

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Wednesday, May 4, 2011 - 0900 CPT

• The Transmission Emergency Operations Center remains activated in Control Mode.

" Restoration activities are underway on damaged transmission facilities.

• The bulk electric system is stable at this time.

" We have restored most loads in north central Alabama.

• All non-restoration related bulk power system switching activities are suspended.

" Critical generator sensitive activities will be scheduled by TRO as system conditions allow.

TransmissionAlert 2 - Widespread System Damage

" Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damagedreliability of TVA's power supply over a widespreadarea with multiple outages to TVA or distributorsystems.

" Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/orCustomer Service Centers.

• All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activated in a Control mode and fully staffed.

A3-11

PSO/Transmission & Reliability Transmission Alert 2 - Widespread System Damage Thursday, May 5, 2011 - 0800 CPT

• The Transmission Emergency Operations Center remains activated in Control Mode.

• Restoration activities are underway on damaged transmission facilities.

• The bulk electric system is stable at this time.

• We have restored most loads in north central Alabama.

• All non-restoration related bulk power system switching activities are suspended.

• Critical generator sensitive activities will be scheduled by TRO as system conditions allow.

TransmissionAlert 2 - WidespreadSystem Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged reliabilityof TVA's power supply over a widespreadarea with multiple outages to TVA or distributorsystems.

• Impacted area involves heavy damage to the TVA power system involving a large number of Transmission Service Centers and/or Customer Service Centers.

" All nonemergency operations are terminated.

• The Customer Resources emergency center and TEMA offices in Nashville are staffed.

• Transmission Emergency OperationsCenter (TEOC) is activatedin a Control mode and fully staffed.

A3-12

PSO/Transmission & Reliability Transmission Alert I - Large Area System Damage Friday, May 06, 2011 - 1000 CPT

• Initial restoration activities have been completed in north Alabama and central Mississippi.

• Construction and repair work will continue for the next few weeks, and months in some cases, on damaged transmission facilities.

• The bulk electric system is stable at this time.

• The activation level of the Transmission Emergency Operations Center has been reduced from "Control" to "Support" Mode with field managers overseeing repair work at specific sites.

• Critical generator and transmission sensitive activities will be scheduled by TRO as system conditions allow.

TransmissionAlert 1 - Large Area System Damage

• Unusual transmissionsystem conditions (weather,power supply, or transmissionproblems) have damaged or threatenedreliabilityof TVA's power supply over a large area with multiple outages to TVA or distributorsystems.

" Impacted area involves several Transmission Service Centers and/or Customer Service Centers.

• All nonessentialpower switching operationsare suspended.

• Area Emergency Team is activated.

• Transmission Emergency Operations Center (TEOC) is activatedand staffed as appropriate in a Support or Control mode as determined by the TransmissionEmergency Team Leader.

A3-13