ML11356A233

10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report

ML11356A233
Person / Time
Site:	Sequoyah
Issue date:	12/16/2011
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
Download: ML11356A233 (25)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 December 16, 2011 10 CFR 50.59 10 CFR 72.48 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328, and 72-034

Subject:

10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report In accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), enclosed is the Sequoyah Nuclear Plant, Units 1 and 2, Summary Report regarding the implemented changes, tests, and experiments for which evaluations were performed in accordance with 10 CFR 50.59(c) and 10 CFR 72.48(c). The enclosure provides a summary of the evaluations which,I occurred since the previous submittal dated May 24, 2010.

There are no commitments contained in this letter. If you have any questions concerning this issue please contact G. M. Cook at (423) 843-7170.

Respec J.

Shea ager, Corporate Nuclear Licensing

Enclosure:

10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector-Sequoyah Nuclear Plant Printed on recycled paper

ENCLOSURE SEQUOYAH NUCLEAR PLANT 10 CFR 50.59 AND 10 CFR 72.48 CHANGES, TESTS, AND EXPERIMENTS

SUMMARY

REPORT

DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN) 22487 Engineering Design Change E22487A is being prepared for implementation of design output calculation SQS201 10. This calculation summarizes the various setpoints used in the Emergency Operating Procedures, as well as time critical operator actions. The following changes are included:

1) Minimum steam generator (SG) pressure setpoints which prevent injection of nitrogen into reactor coolant system (RCS): A minor change in the conservative direction is being made.

2) Auxiliary feedwater flow to each SG if all SGs are faulted: Setpoints will be revised to allow for the readability of the control room indicators.

3) Appendix 2 clarification in the definitions of time critical actions.

4) A minor change to the RCS temperature that can result in flashing at the residual heat removal (RHR) pump suction when it is realigned to the refueling water storage tank (RWST) or to the containment sump, including allowances for normal channel accuracy. The common piping to the containment spray pump suction from the sump is also considered.

5) Added new time critical actions that deal with required operator actions in the event of a loss of coolant accident (LOCA) occurring after RHR is aligned to the RCS for shutdown cooling in Mode 4.

These actions do not currently exist.

The described scenario involves the plant actions after a small break LOCA in Mode 4 has already occurred. Therefore, there will be no changes to frequency of an accident nor will any new accident be created. The NRC in TS change 07-05 (LCO 3.5.3) has recognized that manual actions are necessary to align the RHR system for emergency core cooling system (ECCS) use following a LOCA that may occur in Mode 4, after RHR has been placed in the shutdown cooling mode. Evaluations have been performed that evaluate the actions needed to re-align RHR for ECCS.

These evaluations include what the specific actions are, how long they will take to accomplish, and how long a time period is available for these actions to be completed. The evaluation has included examination for any change in the likelihood of occurrence of a malfunction, and for any new type of malfunction. None were identified. The analysis indicates that the ECCS function and the containment cooling function would be successful; therefore, the fission product barriers would not be adversely affected and there would be no increase in the consequences of any accident or malfunction.

Therefore, based on the results of the evaluation, the design change did not adversely affect plant operation or safety, or require changes to the TSs. Accordingly, prior NRC approval was not required for implementation of the design change.

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DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN) 22487 (Continued)

Only item 5 above screens in for a 10 CFR 50.59 evaluation.

The NRC has approved Technical Specification (TS) change 07-05. The change to TS Limiting Condition For Operation (LCO) 3.5.3 (titled "ECCS-Shutdown,"

applicability is Mode 4) contains the following:

"Note: An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the emergency core cooling systems ECCS mode of operation."

Item 5 above has determined when specific manual actions, other than basic valve alignments, are needed. The possible actions that might be required are actions that would cool the piping system to ensure that the RHR pump net positive suction head requirements are met. To have a complete analysis, item 5 also determined the time required to complete the cooling and valve alignments, the maximum time that is available to complete the actions, core cooling requirements, and containment cooling requirements.

The only situation where item 5 is applicable is after a LOCA has already occurred. The applicable LOCA is the only one that occurs after the RHR system has been aligned for shutdown cooling with the unit in Mode 4.

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DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN)

I S

22582 DCN D22582A replaces the buses from common station service transformer (CSST) B to the start board (these are the alternate supplies for all four start buses). These are redundant to the buses from CSST A and C. It also replaces the outdoor portion of the 2A and the 2B start buses. To support the replacement of the 2A and 2B start buses, an alternate power source is provided to the Unit 2 6.9 kilovolt (kV) unit boards. The source will be the Unit 1 start buses 1A and 1B through the alternate supply breakers in 6.9 kV unit boards 2A and 2C. As the 6.9 kV common boards are normally fed from the Unit 2 start buses, they will also be provided an alternate supply to these boards. A currently spare breaker in the Unit 1 6.9 kV unit board 1C will be used to supply the 6.9 kV common board A. The 6.9 kV common board B will be fed from the existing bus that connects it to the alternate supply breaker of 6.9 kV unit board 2A and 2B. The start board breakers that supply the 2A and the 2B start buses will be changed from a 'break before make' to a 'make before break' for manual transfers only. The automatic transfer is not impacted.

The interlock that prevents CSST B from feeding both start buses for Unit 2 will be removed. The interlock preventing the closure of the normal and the alternate supply breaker to the 6.9 kV unit boards 2A and 2D will be defeated to allow these boards to supply power to Unit 1 during its outage, The Unit 2 unit station service transformers (USSTs) will be disconnected on the high and low side from the bus ducts. The outdoor portion of the bus from the Unit 2 USSTs will be removed to support future modification work.

This 50.59 evaluation examines the impact of the changes made by DCN 22582A. Replacement of the start bus and CSST B bus was evaluated in the screening review as not adverse. The start bus breaker control circuit changes impact the manual transfer of the offsite power connection to the Unit 2 start buses. Alternate power is provided for Unit 2 from the Unit 1 start buses while Unit 2 is undergoing a refueling outage and as shown in calculations in the DCN there is no impact on Unit 1 operation since adequate power is provided to support required Unit 2 functions. The Unit 2 USST buses are disconnected between the Unit 2 USSTs and the Unit Boards and do not impact any accident analysis as the USSTs are not credited for offsite power supply. No new malfunctions were created by the design change and the consequences of postulated malfunctions were determined to be bounded by other previously analyzed events. Therefore, this design change did not result in more than a minimal impact on plant safety or require a change to the TSs. Accordingly, it was concluded that this change was acceptable and no prior NRC approval was required for implementation of the design change.

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DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN) 22419 DCN D22419A replaces the buses from CSST A to This 50.59 evaluation examines the impact of the changes the start board (these are the 1A and 2A normal made by DCN 22419A. The change impacts the offsite power supplies). These are redundant to the buses from connection to the start board for both units and impacts the CSST B. It also replaces the outdoor portion of the 1A connection for both offsite sources for Unit 1. Alternate power and the 1 B start buses. To support the replacement is provided for Unit 1 from the Unit 2 start buses while Unit 1 of the 1A and 1B start buses, an alternate power in undergoing a refueling outage. The Unit 1 USST buses are source is provided to the Unit 1 6.9 kV unit boards, disconnected between the Unit 1 USSTs and the unit boards.

The source will be the Unit 2 start buses 2A and 2B The design change did not adversely impact the reliability of through currently spare breakers in the Unit 2 6.9 kV the offsite power circuits, or increase the likelihood or unit board 2A and the 6.9 kV common board B. The consequences of a malfunction. The USST and its associated start board breakers that supply the 1A and the 1 B bus were not credited for accident mitigation and, as such, the start buses will be changed from a 'break before accident consequences were unaffected by the design make' to a 'make before break' for manual transfers change. Therefore, the design change did not impact safe only. The automatic transfer is not impacted. The operation of the plant or require a change to the TSs.

interlock that prevents CSST B from feeding both start Accordingly, prior NRC approval was not required for buses for Unit 1 will be removed. The Unit 1 USSTs implementation of the design change.

will be disconnected on the high and low side from the bus ducts. The outdoor portion of the bus from the Unit 1 USSTs will be removed to support future modification work.

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DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN) 22238 DCN D22238A is to replace the obsolete steam generator level control system (SGLCS) electronics with a state of the art digital distributed control system (DCS) manufactured by INVENSYS. The new system will be built with redundant-fault-tolerant processors, redundant input process signals, redundant power supplies (both sources and direct current (DC) supplies), redundant switched control networks, and redundant operator display units. The old analog equipment in the auxiliary instrument room racks will be replaced with the INVENSYS DCS components.

Field control processors (FCPs), fieldbus modules (FBMs), fieldbus communications modules (FCMs),

base plates, and cables. Also, a new engineering work station (EWS) rack and master DCS rack will be installed containing central processing unit, monitor, printer, keyboard, mouse, and keyboard-video-mouse (KVM) switches. In the main control room, several instruments will be removed and/or relocated on panel M-3 to make room for two operator display units (flat screens). Old manual/auto stations for the main feedwater regulating valve and bypass valves, and the main feed pump speed controls will be replaced by new manual/auto stations. New alarms that alert the operator to DCS statuses will be added to the annunciator panel on M-3. Input/output equipment (FBMs, FCMs, baseplates, and power supplies) will be added to panels M-10 and M-11 to service the equipment in panel M-3. Two new transmitters will be added to each: 1) the main steam header and 2) the feedwater header that will provide redundant inputs into the DCS.

The new digital DCS system replaces function-for-function the old analog SGLCS with many reliability improvements. The new system provides redundant inputs, redundant processors, networks, power supplies, etc. The new system is designated as "quality related" and is designed to meet quality related requirements. The reliability of the new system is expected to be superior to the old analog system. The modification did not negatively impact any system, structure, or component that is important to safety nor did it impact the consequences of any accidents or the frequency of their occurrence. Based on the evaluation, the new DCS would not cause a new type of malfunction or accident to be created and is expected to reduce the likelihood of SGLCS failures and their consequences by providing a more reliable and redundant control system. Therefore, plant operation and safety were not adversely affected by this design change and no TS change was required. Accordingly, the change did not require prior NRC approval for implementation.

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DESIGN CHANGE DESCRIPTION SAFETY ANALYSIS NOTICE (DCN) 22238 (Continued)

The components that are impacted directly with this modification are the electronic controls for the main feed pumps and the electronic controls for the main and bypass feedwater control valves. The old hardware median signal selectors (MSS) are being replaced by a software MSS. Discussion of that replacement is included in this evaluation. The potential credible failure modes that have been identified in the safety assessment are:

1) Failure of all feedwater regulating valves to close with the main feed pumps slowing to their minimum speed;

2) Control group 1 failing low with both main feed pumps slowing to their minimum speed with its associated control valves closed;

3) Control group 1 failing high, causing its associated regulating valves to fail open and the main feed pump speed increasing to the maximum revolutions per minute (RPMs).

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PROCEDURE DESCRIPTION SAFETY ANALYSIS 0-GO-7 Rev. 62 This change incorporates contingency actions to This 50.59 evaluation examines the proposed change to compensate for unavailability of centrifugal charging 0-GO-7. Both charging pumps will be out of service pumps (CCPs) during maintenance with the RCS simultaneously for maintenance work with the unit in Mode 5.

depressurized in Mode 5. The CCPs are normally RCS cooling, inventory control, and reactivity control will be used for the following purposes in cold shutdown:

performed/controlled by the RHR system. Provisions are included to preclude any possibility of an RCS dilution event.

1. RCS inventory control The procedure change provided adequate controls on RCS

2. RCS boration inventory and boration such that there was no impact on accident frequency, malfunction likelihood, or the To compensate for the loss of RCS inventory control consequences of any accident or malfunction events. No new and boration capability due to the CCPs being accidents or malfunctions were introduced by this change.

removed from service, this revision incorporates Accordingly, based on the results of the evaluation, the various contingency actions. These actions involve procedure change did not adversely affect safe operation of use of the RHR pumps to fill the RCS from the RWST the plant and no TS change was required. Therefore, it was (by opening the RWST isolation valve FCV-63-1 while concluded that the procedure change was acceptable and RHR suction from the RCS is in service), providing prior NRC approval was not required for implementation of the both inventory and boration. Use of the RHR system change.

for RCS inventory control or boration is inconsistent with the Final Safety Analysis Report (FSAR) design functions of the RHR system.

The FSAR describes the following accidents/transients occurring during cold shutdown:

RCS dilution and RCS/RHR overpressurization due to mass or heat input.

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PROCEDURE DESCRIPTION SAFETY ANALYSIS 0-SO-82-1 Rev. 35 0-SO-82-2 Rev. 34 0-SO-82-3 Rev. 38 0-SO-82-4 Rev. 34 50.59 evaluations were performed in support of procedure changes allowing the emergency diesel generator fuel oil duplex filters to be operated in single filter configuration. These changes directly affect equipment uniquely identified in the fuel oil systems and indirectly affect operation of the emergency diesel generators.

Each engine has two duplex filter assemblies, one dedicated only to the fuel oil priming pump and a second duplex filter for all fuel provided to the engine.

Both filter assemblies contain internal valves and a selector switch used to select between three modes of operation. The center position allows fuel to flow through both filter canisters in parallel. When the selector is moved to either side fuel is only allowed to flow through the filter canister selected, either L for the left canister or R for the right canister. The duplex filter assemblies are skid mounted equipment supplied by the vendor with each emergency diesel generator.

The position of the filter selector switch is not included in plant design output or current licensing basis documents including the FSAR.

Present configuration has all the fuel oil duplex filters selected to the center position. At each filter, fuel is being directed through both filter canisters maximizing the filter surface area. This configuration was previously selected to provide maximum filtering while allowing for a minimum number of configuration chanaes and filter maintenance.

The Electro Motive Division (EMD) engines used in the emergency diesel generators are designed to operate in either single or dual fuel filter configurations. An industry precedent exists to support operation in either mode with no increased risk of an accident or malfunction. The recommended filter replacement period is still approximately twice the expected worse case in-service time at the end of a 7 day emergency run if one filter is aligned (previously four times with two filters aligned). Annunciation is available based on fuel header pressure at the discharge of the engine mounted fuel filter.

Fouling is a gradual failure mode allowing time to initiate maintenance with existing procedures and processes. The procedure changes provided adequate controls to ensure that there was no increase in the frequency or consequences of an accident, and no new failure modes were created which could result in the likelihood or consequences of a malfunction to be increased more than minimal. Therefore, the procedure changes did not adversely affect safe plant operation or require a change to the TSs, and prior NRC approval was not required for implementation of the procedure changes.

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PROCEDURE DESCRIPTION SAFETY ANALYSIS 0-SO-82-1 Rev. 35 0-SO-82-2 Rev. 34 0-SO-82-3 Rev. 38 0-SO-82-4 Rev. 34 (Continued)

A change to the single filter configuration is desired in response to operating experience from Florida Power and Light (FPL) during the hurricanes which cut power to much of Florida in 2004. The utility operated several diesel generators similar to Sequoyah's design for extended periods. When fuel was added to the depleted fuel storage tanks while the engines were operating, increased sediment was stirred up from the normally stagnant bottom of the tanks. As a result of the abnormal sediment, the rate of filter fouling increased making filter replacement necessary to continue operating the diesel generators. FPL configures duplex filters with one canister in service and was able to change filters while the diesel generators operated continuously. If fouling occurs in both filter canisters simultaneously, changing to a single canister configuration to replace one canister will result in even further reduction of fuel. In the dual filter configuration it could be necessary to unload and possibly shutdown a diesel generator to replace fuel filters.

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PROCEDURE I DESCRIPTION

=SAFETY ANALYSIS AOP-C.04 Rev. 19 The chemical and volume control system (CVCS) has several safety related functions. The system provides normal and emergency borated water flow paths to the RCS during normal and accident conditions. The system also performs an ECCS function by providing high head safety injection flow during accident conditions. Additionally, the system provides filtered seal water to the reactor coolant pump (RCP) seals during all RCP operation modes. The CCPs are integral in all of these functions because they provide the motive force to transport borated/filtered water from the CVCS to the RCS.

Over the course of the Unit 2 Cycle 16 fuel cycle, it became evident that the 2B-B CCP discharge check valve (2-VLV-62-532) was allowing a small amount of backleakage, which was causing gas to accumulate in the discharge piping of the 2B-B CCP. As water leaked through the discharge check valve, it caused gas to come out of solution resulting in gas accumulation of approximately 0.03 cubic feet (ft3) per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift. Functional Evaluation (FE) 42823, revision 5, was written to evaluate this condition and it contained compensatory actions that the 2B-B CCP preferentially remain in service in order to eliminate the gas accumulation mechanism. If the 2B-B CCP was removed from service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, actions were established to require monitoring of the CCP discharge minimum flow piping and venting as required to remove any developed voids. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> window was established because it ensured that the vertical discharge piping above the 28-B CCP remained water solid.

The proposed activities were determined feasible and effective for maintaining the design functions of the CCPs.

The compensatory measures contained in the procedure did not increase the consequences of an accident or a malfunction, nor did they increase the frequency of an accident or the likelihood of a malfunction. The design functions of the CCPs were maintained consistent with the Updated Final Safety Analysis Report (UFSAR) by the performance of the compensatory measures such that no new accidents or malfunctions would be created. It was, therefore, concluded that the procedure change to implement the compensatory measures did not adversely affect plant operation or safety, or require a change to the TSs.

Accordingly, prior NRC approval was not required for implementation of the compensatory measures.

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PROCEDURE DESCRIPTION SAFETY ANALYSIS AOP-C.04 Rev. 19 At the end of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the vertical section of (Continued) discharge piping will still be full, which will ensure the 2B-B CCP pump casing remains water solid.

Implementation of these compensatory actions will assure that operability of the 28-B CCP is maintained.

The 2A-A CCP experienced some gas void accumulation also, but to a much lesser degree.

Five accident analyses were considered for evaluation of the compensatory measures. The accidents were a small and large break LOCA, a steam generator tube rupture, a faulted steam generator, and a spurious safety injection. The results of the evaluation were that the compensatory actions would be adequate to ensure that the CCP remained functional with respect to its design basis requirements. This evaluation was based upon the assumptions made in the accident analysis and that if the 28-B pump were stopped during an accident, then the gas accumulation in the piping could be stopped simply by swapping from the 2A-A CCP to the 2B-B CCP.

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PROCEDURE DESCRIPTION SAFETY ANALYSIS AOP-C.04 Rev. 19 Compensatory measures were established that would (Continued) enable the 2A and 2B CCPs to remain fully functional.

All automatic and manual pump start sequences were unaffected by the required compensatory measures.

These compensatory actions were implemented by a caution order placed on the 2B-B CCP handswitch in the Main-Control Room and a revision to AOP-C.04.

1. 2B-B CCP should preferentially remain in operation.

2. If the 2B-B CCP was idle, perform an ultrasonic test (UT) examination of the 2B-B CCP discharge piping once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If UT is not, or cannot, be performed, then vent the 2B-B CCP casing every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. Vent the 2B-B CCP casing within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the Inspection Services Organization (ISO) examinations reveal voiding in the vertical portion of the pump discharge piping. If no water is present in the vertical discharge piping, then full functionality of the pump is not ensured. Evaluate TS 3.5.2 for applicability.

4. Perform the ultrasonic testing for gas voids as directed by O-PI-ISO-000-001.A, at least monthly, on the 2A-A CCP.

5. Perform the 2A-A CCP casing venting as directed by 2-SI-OPS-062-040.A (monthly).

6. Perform the 2B-B CCP casing venting as directed by 2-SI-OPS-062-040.1B (monthly)..

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TEMPORARY DESCRIPTION SAFETY ANALYSIS ALTERATION CONTROL FORM (TACF) 1-11-005-063 Increasing pressure has been observed in the Sequoyah (SQN) residual heat removal (RHR) Unit 1 discharge piping since March, 4, 2011. The increasing pressure has been attributed to back leakage through secondary check valves 1-63-634 or 1-63-632 and from their respective primary check valves. The leakage has been determined to be from the RCS, although it is not yet certain which of the two routes contains the leakage. The RHR line must be frequently depressurized to keep from challenging the RHR discharge relief valves 1-VLV-63-626, 1-VLV-63-627 and 1-VLV-63-637, which are set at 600 pounds per square inch gauge. The leakage through the check valves is currently less than 0.1 gallons per minute, well within the TS limits.

TACF 1-11-005-063 documents an interim plant configuration for SQN Unit 1. A continuous vent will be installed under TACF 1-11-005-063 to reduce Operations staff burden of periodically venting the RHR discharge piping. This continuous vent will also reduce challenges to valves 1 -FCV-63-1 11, 1 -FCV-63-071, and 1-FCV-63-084 due to repeated venting actuations. When the continuous vent is in service the Operations staff will still be required to monitor the RHR pressure and make adjustments as necessary to keep the RHR pressure in the desired band. The continuous vent will remain in place until the leaking check valves can be repaired during the next refueling outaae (Unit 1 Cycle18).

The temporary continuous vent configuration was acceptable from a design, fabrication, testing, and performance perspective. The continuous vent was fully qualified to Tennessee Valley Authority (TVA) Class B requirements consistent with the remainder of the ECCS. The continuous vent has no function during an accident, so there was no impact on accident frequency or consequences. No new accidents or malfunctions were created by this temporary alteration, nor were any existing analyzed accidents or malfunctions adversely affected. As the new piping was routed in the same area as existing ECCS piping and the piping was fully qualified, there were no dose consequence changes to the control room or offsite. No significant changes to ECCS flow rates existed and there were no changes to the containment sump inventory. Based on the evaluation, the proposed temporary alteration provided adequate controls such that there was no adverse impact on plant operation or safety. It was, therefore, concluded that the temporary alteration was acceptable to implement in accordance with plant procedures without obtaining a TS change or prior NRC approval of the temporary alteration.

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TEMPORARY DESCRIPTION SAFETY ANALYSIS ALTERATION CONTROL FORM (TACF) 1-11-005-063 (Continued)

TACF 1-11-005-063 installs a check valve, two isolation/throttling valves, and a drain valve between 1-63-628 and 1-62-765. All of the new material is located in the Unit 1 690' pipe chase. All of the new material will be stainless steel and will be installed to TVA Class B requirements. The isolation/throttling valves will be 1/4 inch outside diameter (OD) tubing valves. The valves and connecting tubing were sized in this manner so that effective throttling of leakoff flow can be performed, and also to limit the flow that can pass through the temporary line to a value that obviously can have no adverse impact on delivered RHR/ECCS flow. The check valve will protect against an unanticipated loss of water from the volume control tank (VCT) suction supply to the CCPs. There is no normal flowpath that VCT water can take to be lost to the CCP suction through the new line. The check valve is merely a precaution against multiple failures or inadvertent system manipulations. The check valve is also spring loaded. The cracking pressure of the check valve will not allow RWST water to flow into the CCP suction, as a backup to manual control of the throttling valve by the Operators. The purpose of the throttling valve is to ensure that the RHR discharge pressure can be held elevated above the RWST head pressure to prevent the reactivity changes from the RWST borated water, as well as to maintain the integrity of the RCS water inventory procedure, 0-SI-OPS-68-137.0.

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TEMPORARY DESCRIPTION SAFETY ANALYSIS ALTERATION CONTROL FORM (TACF) 2-11-008-079 Work order 112203835 documents that during the stroke of Unit 2 fuel transfer tube valve (2-VLV-078-0610) the open limit switch 2-SW-079-0001 did not operate when the valve was fully open. This open limit switch illuminates the "valve open" light and is interlocked as a permissive for operation of the fuel transfer conveyor drive control system. An electrical jumper will be installed to bypass the open limit switch to permit operation of the fuel transfer conveyor drive system.

The jumper will be installed inside the fuel system pit side control console 2-PNLC-079-W1 on TB1 between points 301 (S1) and 302 (51). Wires will be lifted for "valve open" light. The jumper will allow the conveyor drive control system to operate as if limit switch 2-SW-079-0001 is closed and the fuel transfer tube valve is fully open. The "valve open" light will be de-energized and the conveyor drive control system will be available for operation without an interlock with the fuel transfer tube valve open limit switch. Lifting the wire going to

valve open" light defeats a false indication for the "valve open" light. The light is an important man-machine interface providing equipment configuration to operate the fuel transfer system. Personnel do not rely on interlocks to operate equipment. Interlocks are safety features. The conveyor drive system is controlled by SSW-4 (conveyor control switch) on/off switch. Procedure controls will require the following to verify the fuel transfer tube gate valve is fully open:

1. Action to count hand wheel turns when opening the valve.

The function of the fuel transfer valve open limit switch is to ensure that there will be no interference between the fuel transfer valve and the fuel transfer cart which could cause problems while refueling. This function of the fuel transfer valve limit switch is described along with many other fuel transfer system features in FSAR section 9.1. This function is not safety related. Electrically bypassing the malfunctioning limit switch and ensuring procedurally that the valve was open prior to moving fuel did not more than minimally increase the risk or consequences of any accident associated with fuel movement. After any manipulation to open the fuel transfer tube valve for fuel transfer, a fuel transfer cart with no fuel was required to be cycled through the fuel transfer tube valve to ensure there was no interference between the cart and the valve. These controls provided assurance that the likelihood and consequences of a malfunction would not be increased.

Therefore, based on the evaluation, the proposed temporary alteration would not adversely affect plant safety and could be implemented without a TS change or prior NRC approval.

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TEMPORARY DESCRIPTION SAFETY ANALYSIS ALTERATION CONTROL FORM (TACF) 2-11-008-079 (Continued)

2. After any manipulation to open the fuel transfer tube valve for fuel transfer, a fuel transfer cart with no fuel will be required to be cycled through the fuel transfer tube valve to ensure there is no interference between the cart and the valve.

The function of the fuel transfer valve limit switch is described in the Sequoyah FSAR, section 9.1. The function of the limit switch is to indicate the fuel transfer tube valve is fully open and to ensure that fuel transfer does not take place unless the valve is fully open. The latter purpose minimizes the chance of valve damage. In the event the cart strikes the valve, the thermal overload on the drive motor will shut down the fuel transfer system. The valve is made of robust, thick stainless steel metal plate and would not be damaged before the thermal overload operates.

Therefore, the functional capability for secondary containment isolation is not impacted. No alternate means of secondary containment isolation is necessary. Per NPG-SPP-09.5 "Temporary Alterations", a FSAR change is not required since this TACF will only be in place for the Unit 2 Cycle 17 refueling outage.

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CORE OPERATING DESCRIPTION SAFETY ANALYSIS LIMITS REPORT (COLR)

SQN Unit 2 Cycle 17, This activity consists of a change in method involving Based on the results of the 50.59 screening review, the COLR, and Cycle modifying the licensed fuel thermal performance proposed activity only required an evaluation.of the impact on Operation codes, TACO3 and GDTACO. This activity consists of evaluation methodologies described in the UFSAR. Each of the following elements:

the two identified elements of the methodology used was

1. The TACO3 and GDTACO methodology is evaluated. The use of TACO3 and GDTACO with an conservatively adjusted with a correction not identified additional penalty was found to be conservative relative to the in the associated topical reports or SERs to account design basis limits. The magnitude of the correction was for decreases in fuel pellet thermal conductivity with found to be appropriate based on the applicability of the burnup.

approved COPERNIC burnup dependent fuel pellet thermal

2. The magnitude of the correction is developed with conductivity methodology to the type of fuel used in the NRC-approved COPERNIC methodology, Sequoyah, Unit 2. Only the COPERNIC burnup dependent specifically, and limited to, the burnup dependent fuel fuel pellet thermal conductivity model was used in the pellet thermal conductivity model.

development of the adjustment to the TACO3 and GDTACO methodology. Therefore, the methodology was determined to be acceptable for use and may be implemented in accordance with plant procedures without obtaining a TS change or prior NRC approval.

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FINAL SAFETY DESCRIPTION SAFETY ANALYSIS ANALYSIS REPORT (FSAR)

FSAR Change This activity consists of a change in method involving Based on the results of the 50.59 screening review, the modifying the licensed fuel thermal performance proposed activity only required an evaluation of the impact on codes, TACO3 and GDTACO. This activity consists of evaluation methodologies described in the UFSAR. Each of the following elements:

the two identified elements of the methodology used was

1. The TACO3 and GDTACO methodology is evaluated. The use of TACO3 and GDTACO with an conservatively adjusted with a correction not identified additional penalty was found to be conservative relative to the in the associated topical reports or SERs to account design basis limits. The magnitude of the correction was for decreases in fuel pellet thermal conductivity with found to be appropriate based on the applicability of the burnup.

approved COPERNIC burnup dependent fuel pellet thermal

2. The magnitude of the correction is developed with conductivity methodology to the type of fuel used in the NRC-approved COPERNIC methodology, Sequoyah, Unit 1 and Unit 2. Only the COPERNIC burnup specifically, and limited to, the burnup dependent fuel dependent fuel pellet thermal conductivity model was used in pellet thermal conductivity model.

the development of the adjustment to the TACO3 and GDTACO methodology. Therefore, the methodology was determined to be acceptable for use and may be implemented in accordance with plant procedures without obtaining a TS change or prior NRC approval.

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COMMITMENTS DESCRIPTION SAFETY ANALYSIS NRC RG-1.23 Rev. 1 This change involves replacing the existing FSAR Committing to RG-1.23, revision 1, will have an overall benefit commitment to Nuclear Regulatory Commission to TVA, with minimal cost and no adverse impacts.

(NRC) Regulatory Guide (RG)-1.23, revision 0, with a Specifically, since TVA already operates its meteorological commitment to NRC RG-1.23, revision 1, replacing monitoring program according to American National the dewpoint sensor, and changing the sampling Standards Institute/American National Society (ANSI/ANS)-

intervals for some meteorological variables.

3.11 (2005), most of the differences between RG-1..23, revision.0, and RG-1.23, revision 1, are already TVA current practices.

However, implementing RG-1.23, revision 1, will impact the SQN meteorological monitoring program in the following three ways because of differences between RG-11.23, revision 0, and RG-1.23, revision 1.

1. Replace the dewpoint sensor:

The current dewpoint sensor (chilled-mirror) meets the RG-1.23, revision 0, accuracy but is obsolete, has a high lost record rate, and requires an excessive amount of maintenance. The proposed sensor (capacitive humidity) will satisfy RG-1.23, revision 1, but not RG-1.23, revision 0. The relaxation in dewpoint accuracy will have virtually no impact on SQN since the data are acceptable for identified applications.

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COMMITMENTS DESCRIPTION

[SAFETY ANALYSIS NRC RG-1.23 Rev. 1 (Continued)

2. Change the sampling intervals for some meteorological variables:

Only minor software changes are needed, since the current planned equipment is capable of collecting data for all variables at the desired rate. The summary data will continue to be arithmetic averages (using more data samples) and should be essentially indistinguishable from data using the current averaging methodology.

3. Install rain gauge wind shield:

The rain gauge at the SQN meteorological tower is not equipped with a wind shield because the precipitation of concern (wind-blown snow) is uncommon at SQN; however, the cost to install a wind shield is small, and the difference in precipitation (between a shielded and unshielded rain gauge) is expected to be within the normal variability. In the interest of standardization with other rainfall sampling sites, a wind shield will be installed.

Based on the evaluation of the above described changes, there would be no impact on the accidents and malfunctions described in the UFSAR, nor would new accidents or malfunctions be created. The described changes also would not increase the consequences of any accidents or malfunctions since only the sampling method is affected, which is assessed to be more reliable and to improve the quality of the meteorological data. Therefore, the commitment change to RG-1.23, revision 1, did not adversely affect plant operation or safety, or require a TS change. Accordingly, prior NRC prior approval was not required for implementation of the commitment change.

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DOCUMENT DESCRIPTION SAFETY ANALYSIS NUMBER/72.48 EVALUATION TRACKING NUMBER 72.212 Evaluation The NRC has reviewed and approved HI-STORM 100 TVA has re-performed the calculation for the design basis Report Rev. 4 Certificate of Compliance (CoC) 1014, Amendment probable maximum river flood. Calculation No. 5, for use in accordance with the general license CDQ000020080054, revision 0, documents a new PMF provisions of 10 CFR 72, Subpart K. Changes to the elevation for SQN. The new PMF level is elevation 722.0 feet HI-STORM 100 FSAR that are incorporated into with the current river configuration. This is an increase over revision 7 represent changes to analysis submitted to the previous PMF level of 719.6 feet. In addition, the PMF the NRC and changes by the certificate holder in level will change to elevation 723.3 feet when current accordance with the provisions of 10 CFR 72.48 and construction on the Chickamauga Lock is completed. There the authority granted to the certificate holder.

are other calculations to be performed associated with the evaluation of river flooding that will address the PMF level In accordance with 72.212 (b) (2) (ii), the licensee when the Chickamauga Lock construction is complete. Until shall evaluate any changes to the written evaluations then, interim information is being added to supplement the (i.e. documented in SQN's 72.212 written evaluation existing information in the 10 CFR 72.212 report and report) required by this paragraph using the calculation SQS20224 for the PMF level change from 719.6 requirements of 10 CFR 72.48(c). This review feet to 722.0 feet to document the impact on spent fuel stored evaluates the changes to the SQN 72.212 evaluation in dry casks at SQN.

report that are necessary to document the acceptability of a change to the probable maximum Problem evaluation report (PER) 162711 documents the flood (PMF) level for design basis flooding of spent problem associated with the revision to the flooding evaluation fuel stored in the Independent Spent Fuel Storage calculation as it relates to SQN. FE 43152, revision 4, for Installation (ISFSI).

PER 162711 provides the current FE for SQN which concludes that SQN is functionally operable. The corrective actions to PER 162711 will provide for the revising of the 10 CFR 72.212 report and calculation SQS20224 once all flooding calculations are completed. DCN 22404 will implement these changes.

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DOCUMENT DESCRIPTION SAFETY ANALYSIS NUMBER/72.48 EVALUATION TRACKING NUMBER 72.212 Evaluation For the interim, Holtec has provided supplemental analysis for Report Rev. 4 those Multipurpose Canisters (MPCs) loaded under (Continued)

Amendments 1 and 2 of decay heat loads up to 28.74 kilowatt (kW) and for MPCs to be loaded under Amendment 5 of decay heat loads up to 36.9 kW. Holtec document 1058218, revision 0, and Holtec Report HI-2104634, revision 0, document the reanalysis, respectively. The reanalysis was performed to address the new PMF level with a corresponding silt deposition of up to 4.6 inches for Amendments 1 and 2 and 5 inches for Amendment 5. For Holtec CoC Amendments 1 and 2, calculations performed in the prior analysis (calculation No. 3CG1S667, revision 3) remain valid with the new silt height of 5 inches for the change in SQN PMF, with additional justification documented under a Holtec Letter dated April 8, 2010. For MPCs to be loaded under Amendment 5, Holtec has determined the short term temperature limit to be 992 degrees Fahrenheit which does not exceed the limit of 1058 degrees Fahrenheit, as defined by the ISFSI FSAR, for the HI-STORM overpacks which experience the PMF, and the structural integrity of the MPC and overpack will not be compromised.

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DOCUMENT DESCRIPTION SAFETY ANALYSIS NUMBER/72.48 EVALUATION TRACKING NUMBER 72.212 Evaluation Similarly, Mactec has performed a. reanalysis of the effects of Report Rev. 4 the PMF on the SQN site. Mactec report 3043-10-1020 is an (Continued) addendum to the Mactec report 3043-03-1065 which provides the original 2004 analysis of the effects of the PMF in support of the ISFSI work at SQN. Mactec report 3043-10-1020 essentially concludes there is not a significant change in the original conclusions in Mactec report 3043-03-1065, with the exception of expected silt deposition. Silt deposition as a result of the new PMF is expected to increase from 3 - 4 inches to 3.25 - 4.5 inches.

The 50.59 evaluation assessed the above described changes and associated analyses and it was determined that the change to PMF level would not impact the frequency of an accident or likelihood of a malfunction previously evaluated in the ISFSI UFSAR. There would also be no increase the consequences of an accident or a malfunction, nor would the change create a new accident or malfunction not previously evaluated in the ISFSI UFSAR. No new evaluation methodologies were used. Therefore, the change in the PMF did not constitute a safety concern with respect to the ISFSI, and no license amendment was required.

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