Text

License Amendment Request to Revise Technical Specifications (TS) 3.3.2, ESFAS Instrumentation, and TS 3.5.4, Refueling Water Storage Tank (Rwst).
	ML080150161
Person / Time
Site:	Vogtle
Issue date:	01/09/2008
From:	Stinson L; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-07-2168
	Download: ML080150161 (44)
	v • d • e

.,

L M. Stinson (Mike) Southern Nuclear Vice President Operating Company, Inc.

Fleet Operations Support 40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 3521]1 TeI205.992.51B1 Fax 205.992.0341 SOUTHERN COMPANY A

January 9, 2008 Energy to Serve YtJur World" Docket Nos.: 50-424 NL-07-2168 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2, "ESFAS Instrumentation." and TS 3.5.4. "Refueling Water Storage Tank (RWST)"

Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC), hereby requests an amendment to Facility Operating License Nos. NPF-68 and NPF-81 for the Vogtle Electric Generating Plant (VEGP), Units 1 and 2, respectively.

The proposed amendment will revise VEGP Units 1 and 2 Technical Specification (TS) TS 3.3.2, "ESFAS Instrumentation," Table 3.3.2-1, "Engineered Safety Feature Actuation System Instrumentation," Function 7.b, and TS 3.5.4, "Refueling Water Storage Tank (RWST)," Surveillance Requirement (SR) 3.5.4.2.

The proposed change to TS 3.3.2 lowers the Nominal Trip Setpoint and corresponding Allowable Value of the Refueling Water Storage Tank (RWST)

Level - Low Low at which the semi-automatic switchover from the RWST to the containment emergency sump occurs. The proposed change to TS 3.5.4, SR 3.5.4.2 will increase the minimum required RWST borated water volume.

Revisions to the associated TS Bases based on the proposed changes are also included.

These proposed TS changes are required to meet commitments related to the resolution of issues identified in NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors," dated September 13, 2004. The changes to TS 3.3.2 and 3.5.4 will increase the emergency sump water levels and net positive suction head (NPSH) available for emergency core cooling (ECCS) at the time of switchover to cold leg recirculation.

These changes were necessitated by the recent identification of a design deficiency that, in the worst case, could result in the residual heat removal (RHR) containment sump screens being uncovered at the initiation of switchover of the ECCS suction from the RWST to the containment sumps following a design basis loss of coolant accident (LOCA). A request for extension of the date for

U. S. Nuclear Regulatory Commission NL-07-2168 Page 2 completion of all actions pursuant to GL 2004-02 was submitted by SNC on December 7,2007 (NL-07-1969). This request for extension addressed, in part, the need for these proposed changes to the TS. An extension to May 31,2008 to allow, in part, NRC review and approval, and implementation of these proposed changes was granted by the NRC by letter dated December 19, 2007.

The enclosed amendment request is subdivided as shown below:

Enclosure 1 provides a basis for the proposed changes.

Enclosure 2 includes the marked-up TS and TS Bases pages with the proposed changes.

Enclosure 3 includes the associated typed TS and TS Bases pages with the proposed changes incorporated for VEGP.

Southern Nuclear Operating Company requests approval of the proposed license amendments by April 14, 2008. The proposed changes would be implemented by May 31,2008.

Mr. L. M. Stinson states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely,

~a~

L. M. Stinson Vice President Fleet Operations Support .

),~ j 1-". !' I ,<I

Swo~ to and subscribed before me this !I!!- .......c'-'y'Cf day of -.J1'nCL!ua , 2008.

~. (JiLL Notary Public My commission expires: X ("'JT 5:; AJlo LMS/LPH/daj

U. S. Nuclear Regulatory Commission NL-07-2168 Page 3

Enclosures:

1. Basis for Proposed Changes

2. Markup of Proposed TS and TS Bases Changes

3. Typed Pages for TS and TS Bases Changes cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. T. E. Tynan, Vice President - Vogtle Mr. D. H. Jones, Vice President - Engineering RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Acting Regional Administrator Mr. S. P. Lingam, NRR Project Manager - Vogtle Mr. G. J. McCoy, Senior Resident Inspector - Vogtle State of Georgia Mr. N. Holcomb, Commissioner - Department of Natural Resources

Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifica,tions (TS) 3.3.2, "ESFAS Instrumentation," and TS 3.5.4, "Refueling Water Storage Tank (RWST)"

Enclosure 1 Basis for the Proposed Changes

Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2, "ESFAS Instrumentation," and TS 3.5.4, "Refueling Water Storage Tank (RWST)"

Enclosure 1 Basis for the Proposed Changes Table of Contents 1.0 Summary Description 2.0 Detailed Description 3.0 Background 4.0 Technical Evaluation 4.1 Impact of the Proposed Changes on the Safety Analyses 5.0 Regulatory Evaluation 5.1 Significant Hazards Consideration 5.2 Applicable Regulatory Requirements / Criteria 5.3 Precedent 5.3.1 Relevant NRC RAls 5.4 Conclusions 6.0 Environmental Consideration 7.0 References

Enclosure 1 Basis for the Proposed Changes 1.0 Summary Description The proposed amendment will revise Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Technical Specification (TS) TS 3.3.2, "ESFAS Instrumentation," Table 3.3.2-1, Function 7.b and TS 3.5.4, "Refueling Water Storage Tank (RWST)," Surveillance Requirement (SR) 3.5.4.2. The proposed change to TS 3.3.2 lowers the Nominal Trip Setpoint (NTS) and corresponding Allowable Values (AV) for the Refueling Water Storage Tank (RWST) Level - Low Low level at which the semi-automatic switchover from the RWST to the containment sump occurs.

The proposed change to TS 3.5.4, SR 3.5.4.2 will increase the minimum required RWST borated water volume from "~631 ,478 gallons" to

"~686,000 gallons." Revisions to the associated TS Bases based on the proposed changes are also included. These proposed TS changes are required to meet commitments to the resolution of issues identified in NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors," dated September 13, 2004.

These changes were necessitated by the recent identification of a design deficiency that, in the worst case, could result in the residual heat removal (RHR) containment sump screens being uncovered at the initiation of switchover of the emergency core cooling system (ECCS) suction from the RWST to the containment sumps following a design basis loss of coolant accident (LOCA). A request for extension of the date for completion of all actions pursuant to GL 2004-02 was submitted by Southern Nuclear Operating Company (SNC) on December 7,2007 (NL-07-1969) (Reference 1). This request for extension addressed, in part, the need for these proposed changes to the TS. An extension to May 31,2008, to allow, in part, NRC review and approval, and implementation of these proposed changes was granted by the NRC by letter dated December 19, 2007 (Reference 2).

2.0 Detailed Description Proposed changes to TS 3.3.2, TS 3.5.4, and SR 3.5.4.2 are summarized below.

TS 3.3.2. "ESFAS Instrumentation," Table 3.3.2-1 (page 6 of 7),

"Engineered Safety Feature Actuation System Instrumentation." Function 7.b:

The Nominal Trip Setpoint (NTS) for Function 7, "Semi-automatic Switchover to pontainment Sump," item b, "Refueling Water Storage Tank (RWST) Level - Low Low," will be revised from "275.3 in" to "213.5 in" and the AV will be revised from "~264.9 in." to "~216.6 in. and ~ 210.4 in." The AV is being changed from a one-sided value to a two-sided value.

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Enclosure 1 Basis for the Proposed Changes Currently, the one-sided AV is in the decreasing direction from the NTS, which decreases operator action time to complete the switchover to the containment sump decreases. This remains true for the revised NTS, so an appropriate AV in the decreasing direction is retained. However, because the primary purpose of the proposed change to the NTS is to increase the volume of water in containment at the initiation of ECCS suction switchover by delaying the initiation of switchover, an AV in the increasing direction of the NTS is necessary as well.

TS 3.5.4. "Refueling Water Storage Tank (RWST)." SR 3.5.4.2:

The RWST borated water volume to be verified in the SR will be changed from a value of"~ 631,478 gallons" to "~686,000 gallons." The 631,478 gallons and 686,000 gallons represent actual contained borated water volumes in the RWST.

In addition, the associated revisions to TS Bases are included.

3.0 Background

By letter dated August 31, 2005 (Reference 3), SNC submitted a combined response for Joseph M. Farley Nuclear Plant (FNP) and VEGP in response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors." In this letter, SNC committed to the installation of the VEGP Unit 1 and Unit 2 new post-LOCA (loss of coolant accident) containment sump recirculation screens, completion of required modifications, and implementation of required procedural changes by December 31,2007.

By letter to the NRC, dated June 22,2006 (Reference 4), SNC requested the completion of the modifications to mitigate downstream effects for VEGP Unit 1 to be extended until the completion of the Unit 1 2008 Spring outage. In a teleconference on June 30,2006 with the NRC, SNC was requested to provide an update of on-going activities and a clarification as to what activities were driving the extension request. By letter dated July 28, 2006 (Reference 5), SNC provided an update of on-going activities and a clarification as to what activities were driving the extension request. The extension request was approved by the NRC in a letter dated September 7, 2006 (TAC MC4727) (Reference 6).

New containment emergency sump screens were installed in VEGP Units 1 (1 RF13, Fall 2006) and 2 (2RF12, Spring 2007) that increased the available screen area from approximately 54 square feet to 765 square feet for each of the Residual Heat Removal (RHR) screens, an approximate 1400%

increase, and 'from approximately 54 square feet to 590 square feet for each of the Containment Spray System (CSS) screens, an approximate 1075% increase. The screens have a smaller mesh size, 3/32 inch El-3

Enclosure 1 Basis for the Proposed Changes diameter versus 1/8 inch square opening in the old screens. Current modeling shows that the water level in the containment building will be at 175 feet-8 inches at the initiation of sump recirculation. The design for the containment sump screens, as confirmed by testing by General Electric, requires that the containment sump water level be maintained at 177 feet or greater to assure screen performance with the assumed debris loading.

This water level assures that there is a minimum of three (3) inches of water over the top of the screens (elevation 176 feet-9 inches). The proposed changes to lower the semi-automatic switchover setpoint and increase the minimum RWST borated water volume are required because of the design of the new sump screens, in that the screens must be fully submerged.

These changes were necessitated by the recent identification of a design deficiency that, in the worst case, could result in the residual heat removal (RHR) containment sump screens being uncovered at the initiation of switchover of the ECCS suction from the RWST to the containment sumps following a design basis loss of coolant accident (LOCA).

By SNC letter (NL-07-1969) to the NRC, dated December 7,2007 (Reference 1), SNC requested an extension to the completion schedule to extend the completion of the corrective actions required by Generic Letter 2004-02 for VEGP Units 1 and 2 from December 31, 2007 to June 30, 2008. This would allow SI\JC to complete downstream effects modifications for Unit 1, complete the downstream effects evaluation for both units, receive and process the screen vendors' test reports for chemical effects testing, and allow sufficient time for the NRC review and approval of this TS amendment request. As stated in the letter, the TS amendment is needed to ensure that the new RHR sump screens would be fully submerged in the worst case accident scenario at the initiation of cold-leg recirculation. This extension request was subsequently approved by the NRC in a letter dated December 19, 2007 to May 31, 2008 (Reference 2).

4.0 Technical Evaluation Revisions to the RWST minimum inventory and semi-automatic switchover to containment sump setpoint will ensure that the ECCS and CSS sump screens will be fully SUbmerged at the time of the initiation of switchover to ECCS recirculation from the containment emergency sumps. The change in delivered water volume between the current RWST minimum inventory and the RWST semi-automatic switchover to containment sump setpoint will increase the amount of water available in the containment sumps at the beginning of ECCS switchover. The containment sump setpoint revision will result in an increase of approximately 69,000 gallons in delivered RWST water, and the RWST volume revision will result in an increase of approximately 54,500 gallons. Therefore, the total increase in delivered RWST water prior to semi-automatic switchover to containment sump is approximately 123,500 gallons.

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Enclosure 1 Basis for the Proposed Changes Not only will this increase in delivered RWST water volume increase the available NPSH, it will delay the start of switchover 'from approximately 8 minutes to several hours, depending on the LOCA break size.

The calculation of the maximum containment water levels is based on the full contained volume of the RWST, maximum inventory from the LOCA, and conservative modeling of the containment to maximize the flood level for the design and environmental qualification of equipment. This analysis is unchanged by the change in the semi-automatic switchover RWST setpoint, because it is based on the physical characteristics of the RWST (maximum contained volume) and does not use the switchover setpoint.

Therefore, the calculation of maximum containment water level is unchanged.

The proposed changes will have no adverse impact on any ECCS equipment.

TS 3.3.2, "ESFAS Instrumentation," Table 3.3.2-1 (page 6 of 7),

"Engineered Safety Feature Actuation System Instrumentation," Function 7.b:

During the ECCS injection phase, water is taken from the RWST and injected into the Reactor Coolant System (RCS) through the cold legs.

When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical, the sump screens are covered, and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation.

The RWST supplies both trains of the ECCS and the Containment Spray System (CSS) during the injection phase of a LOCA recovery. The RWST Level - Low Low signal automatically initiates opening of the RHR containment emergency sump isolation valves. Upon receipt of the RWST Level - Low Low signal. the operator is required to perform manual switchover steps to complete the switchover in an orderly and timely manner and in the proper sequence. Upon completion of the switchover steps, the operator will verify proper operation and alignment of all ECCS components. The ECCS switchover steps, as described in the VEGP FSAR, have not changed. The volume between "Low Low" (initiation of ECCS switchover process) and "Empty" (termination of ECCS switchover and initiation of CS switchover) is large enough to permit the operator adequate time to manually complete all the manual steps needed.

The revised RWST Level - Low Low nominal Setpoint will continue to ensure that ECCS switchover is completed prior to receipt of the RWST Empty alarm, The CSS switchover will continue to occur based on level indication. The switchover will commence after the receipt of the RWST Empty alarm. The revised RWST Empty alarm (from 10% to 8%) provides sufficient margin to ensure CSS switchover is complete and suction to the El-5

Enclosure 1 Basis for the Proposed Changes RWST is isolated without allowing air entrainment from the RWST into the ECCS or CSS pumps. The CSS switchover steps, as described in the VEGP FSAR, have not changed.

The calculation of minimum containment water levels is based on the minimum delivered volume of the RWST, minimum inventories from the spectrum of LOCA break sizes, and conservative modeling of the containment to minimize the containment level. This analysis is based on the physical characteristics of the RWST setpoint and the minimum required TS volume.

The proposed change to TS Table 3.3.2-1, Function 7, "Semi-automatic Switchover to Containment Sump," item b, "Refueling Water Storage Tank (RWST) Level - Low Low," will increase the volume of water pumped into containment by approximately 69,000 gallons prior to the initiation of switchover.

The AV for Function 7.b, has been changed from ".::=. 264.9 in." to "50 216.6 in. and.::=. 210.4 in." In addition, the Nominal Trip Setpoint for item b has been changed from "275.3 in" to "213.5 in."

The basic uncertainty algorithm used to determine the overall instrument uncertainty for the RWST Level- Low Low ESFAS (engineered safety features actuation systems) function is the square-root-sum-of-the squares (SRSS) of the applicable uncertainty terms. This is the same methodology as defined in WCAP-11269, Revision 1, 'Westinghouse Setpoint Methodology for Protection Systems - Vogtle Station" (Reference 7), as modified for term dependencies. This methodology combines the uncertainty components for a channel in an appropriate combination of those groups which are statistically and functionally independent. Those uncertainties which are not independent are conservatively treated by arithmetic summation and then combined via SRSS with the independent terms. This approach is consistent with NRC Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," and ISA Standard 67.04.01-2006, "Setpoints for Nuclear Safety-Related Instrumentation."

A calculation determined the bounding Channel Statistical Allowance (CSA) for the RWST Level- Low Low function. The AVs are determined by adding (or subtracting) the calibration accuracy, defined by the Rack Calibration Accuracy (RCA), of the tested channel during the Channel Operational Test (COT) to the Nominal Trip Setpoint (NTS) in the non conservative direction(s) (Le., toward or closer to the Safety Analysis Limit(s) (SAL)) for the application. The magnitude of the as left (calibration accuracy term) and the as found tolerances are the same and are specified in plant procedures.

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Enclosure 1 Basis for the Proposed Changes The proposed changes to TS Table 3.3.2-1 Function 7.b will be implemented via changes to the impacted NTS and AV. There are four (4)

RWST level channels. The semi-automatic switchover requires any two (2) out of four (4) channel bistables to actuate. Changes to these setpoints will be accomplished by rescaling each level transmitter to the new range and adjusting the trip point of each bistable in each level channel.

In conclusion, the proposed change to TS Table 3.3.2-1, Function 7.b is acceptable based on the evaluation discussed above. The TS Bases for 3.3.2 and 3.6.6 have been revised accordingly (Enclosures 2 and 3).

TS 3.5.4, "Refueling Water Storage Tank (RWST)," SR 3.5.4.2:

The RWST supplies borated water to the Chemical and Volume Control System (CVCS) during abnormal operating conditions, to the refueling pool during refueling, and to the ECCS and the Containment Spray System during accident conditions.

As part of this TS amendment request, the minimum RWST borated water inventory requirement will be increased from ~ 631 ,478 gallons to

~ 686,000 gallons. The proposed change to SR 3.5.4.2 will increase the volume of water pumped into containment by approximately 54,500 gallons prior to the initiation of switchover to the containment sump.

The RWST minimum water volume is calculated based on the tank parameters; as such, instrument uncertainty is not a factor. To ensure that the calculated volume is delivered, instrument uncertainties are procedurally included for verification of compliance with the TS limit.

During the ECCS injection phase, water is taken from the RWST and injected into the RCS through the cold legs. When sufficient water is delivered from the RWST to ensure that enough borated water has been added to maintain the reactor subcritical, the emergency sump screens are submerged, and the containment sumps have enough water to ensure that the required net positive suction head to the ECCS pumps is available, suction is switched to the containment sump for cold leg recirculation.

As discussed previously, the RWST supplies both trains of the ECCS during the injection phase of a LOCA. A motor operated isolation valve is provided for each pump to isolate the RWST from the ECCS once the system has been transferred to the recirculation mode. The recirculation mode is entered when pump suction is transferred to the containment sump following receipt of the RWST Level - Low Low signal. During normal operation in MODES 1, 2, and 3, the safety injection (SI) and residual heat removal (RHR) pumps are aligned to take suction from the RWST. During accident conditions, the RWST prOVides a source of borated water to the ECCS pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for El-7

Enclosure 1 Basis for the Proposed Changes reactor shutdown. Insufficient water in the RWST could result in insufficient cooling capacity when the transfer to the recirculation mode occurs.

Therefore, the proposed change to TS 3.5.4, SR '3.5.4.2 is acceptable based on the evaluation discussed above. The TS Bases for 3.5.4 and 3.6.6 have been revised accordingly (Enclosures 2 and 3).

4.1 Impact of the Proposed Changes on the Safety Analyses Non-LOCA Analyses Reducing the Nominal Trip Setpoint (NTS) for the semi-automatic switchover from the RWST to the containment sump, and increasing the minimum RWST borated water volume do not impact the Non-LOCA analyses because they are not credited in any of the Non-LOCA analyses.

Small Break LOCA Analysis The Small Break LOCA (SBLOCA) analysis models the minimum useable RWST volume available before the signal for switchover to cold leg recirculation is initiated; Le., the RWST water volume between the SI switchover alarm and the minimum operating level. The value used in the current SBLOCA analysis is 302,000 gallons. The minimum useable RWST volume available before the signal for switchover to cold leg recirculation is initiated will increase to 435,522 gallons. Therefore, the SBLOCA analysis is not impacted by reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume.

Large Break LOCA Analysis The Large Break LOCA (LBLOCA) analysis does not model the RWST volume. Therefore the LBLOCA analysis is not impacted by reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume.

LOCA Long Term Cooling Evaluation LOCA Long Term Cooling (LTC) analyses are the calculations performed to ensure that the core remains subcritical in the sump recirculation mode, the calculations to preclude boric acid precipitation in the core after a LOCA, and the calculations to confirm the capability for long term decay heat removal.

The analysis that is performed to ensure that the core remains subcritical after the switchover to sump recirculation is referred to as the Post-LOCA Subcriticality Analysis. The results of this analysis are confirmed every core reload cycle to demonstrate that the water contained in the sump after a El-8

Enclosure 1 Basis for the Proposed Changes LBLOCA contains sufficient boron to ensure that the core will remain subcritical for the long term. The limiting condition for this analysis is minimum RWST water volume less the volume in the reactor cavity at the time of hot leg switchover. An evaluation was performed and concluded that the minimum RWST volume assumptions, associated with reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume, are conservative with respect to, or unchanged from, those used in the current licensing basis calculations. Therefore, reducing the NTS for the semi automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume would not have an adverse impact on the Post-LOCA Subcriticality Analysis.

The post-LOCA LTC Analysis which is performed to preclude boric acid precipitation in the core is also referred to as the Hot Leg Switch-over (HLSO) analysis. The calculations that predict the buildup of boric acid in the core are primarily impacted by the vessel mixing volume and boron concentration assumptions for the various sources of water that drain to the containment sump. For Vogtle Units 1 and 2, the most recent HLSO analysis used a maximum RWST volume of 706,199 gallons. The increase in the maximum deliverable RWST volume to 731,000 gallons has only a small effect on the core boric acid concentration, and the effect is within the margin available within the current 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Therefore, reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume would not have an adverse impact on the HLSO analysis.

LTC analyses also ensure adequate core flushing flow (before and after the switchover to hot leg recirculation) and long term decay heat removal.

These analyses confirm the adequacy of the post-LOCA SI flow in both the cold and hot leg recirculation modes. The only effect of an increased RWST minimum water volume on these analyses would be a minor beneficial impact, since it would delay the switchover to sump recirculation slightly, such that the decay heat boil-off would be lower. Therefore, reducing the I\lTS for the semi-automatic switchover from the RWST to the containment sump, and increasing the minimum RWST borated water volume would not have an adverse impact on the LTC analyses.

In summary, reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume have been evaluated with respect to LOCA Long Term Cooling, and it is concluded that the changes are acceptable with respect to the current Vogtle Units 1 and 2 LOCA Long Term Cooling licensing basis analyses.

Steam Generator Tube Rupture (SGTR) Analysis Although the SGTR analysis includes modeling of SI flow, the RWST is included in the model only as a source for the injection flow to define its El-9

Enclosure 1 Basis for the Proposed Changes temperature. There is nothing included in the model to consider or track RWST inventory. It should be noted that during an SGTR, the RCS is essentially intact and the RCS pressure remains high, limiting the amount of mass injected into the RCS. Also, the emergency operating procedures (EOPs) require the operators to terminate SI flow as part of the SGTR mitigation actions. In the VEGP SGTR analysis, as discussed in UFSAR Section 15.6.3, SI flow is terminated approximately in an hour from the time it was initiated on low pressurizer pressure.

Radiological Dose Analyses (except for the LOCA dose analyses)

The radiological dose analyses (except for the LOCA dose analyses) do not model the RWST; therefore, they are not impacted by reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume.

LOCA Dose Analyses and Recirculation Fluid pH Control System The proposed change that increases the RWST borated water volume does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The minimum containment sump pH used in calculating the radiological consequences for a LOCA remains bounding.

The minimum and maximum pH values remain bounding; therefore, the required amount of trisodium phosphate (TSP) remains unchanged. The impact on the containment flood level, equipment qualification, hydrogen produced by the corrosion of galvanized surfaces and zinc based paints, and chloride induced stress corrosion remains within the limits assumed in the design and accident analyses.

LOCA Mass and Energy Release Analysis and LOCA Containment Response Analysis The RWST is not credited/modeled in the Short-Term LOCA Mass and Energy Release Analysis due to the short duration of the transient.

Therefore, the Short-Term LOCA Mass and Energy Release analysis is not impacted by reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume.

The Long-Term LOCA Mass and Energy Release Analysis and LOCA Containment Response Analysis are based on a minimum deliverable RWST water volume of 501,257 gallons for calculating the time for switchover to sump recirculation. Reducing the NTS for the semi-automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume is a benefit, since they will provide an increase in cold RWST water for steam condensation and containment spray and subsequently result in less limiting long-term containment peak pressure and temperature conditions.

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Enclosure 1 Basis for the Proposed Changes Steam Line Break (SLB) Mass and Energy Release Analyses and SLB Containment Response Analysis The SLB Inside Containment Mass and Energy Release Analysis and SLB Containment Response Analysis do not model the emptying of the RWST and switchover to sump recirculation. Therefore, the SLB Inside Containment Mass and Energy Release Analysis and SLB Containment Response Analysis are not impacted by reducing the NTS for the semi automatic switchover from the RWST to the containment sump, and increasing the minimum RWST borated water volume.

In addition, the SLB Outside Containment Mass and Energy Release Analysis does not model the emptying of the RWST and switchover to sump recirculation. Therefore, the SLB Outside Containment Mass and Energy Release Analysis is not impacted by reducing the NTS for the semi automatic switchover from the RWST to the containment sump and increasing the minimum RWST borated water volume.

Boration Capability Evaluation The Boration Capability Evaluation is performed to determine the minimum RWST volume required to borate the RCS to the applicable shutdown margin requirements. Increasing the RWST contained/useable volume and reducing the Level - Low Low Setpoint is conservative with respect to the boration capability evaluation results, since these changes will provide a greater RWST boration volume. Therefore, there will be no negative impact on the boration capability evaluation. The final details of this change will be addressed and updated as part of the normal reload process (e.g.,

addressed in the Reload Safety Analysis Checklist) and the results included in the next fuel cycle Reload Safety Evaluation (RSE) as appropriate.

5.0 Regulatory Evaluation 5.1 Significant Hazards Consideration According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment. would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

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Enclosure 1 Basis for the Proposed Changes SNC has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

TS 3.3.2, "ESFAS Instrumentation," Table 3.3.2-1 (page 6 of 7),

"Engineered Safety Feature Actuation System Instrumentation,"

Function 7.b:

No. The previously analyzed accidents are initiated by the failure of plant structures, systems, or components, The proposed change that decreases the Allowable Value and Nominal Trip Setpoint (NTS) of the semi-automatic switchover to containment sump (RWST Level

- Low Low) does not have a detrimental impact on the integrity of any plant structure, system, or component (SSC) that initiates an analyzed event. The change does not adversely affect the protective and mitigative capabilities of the plant, nor does the change impact the initiation or probability of occurrence of any accident. The SSCs will continue to perform their intended safety functions.

The minimum containment sump pH used in calculating the radiological consequences for a LOCA remains bounding. The offsite and control room doses will continue to meet the requirements of 10 CFR 100 (Reactor Site Criteria) and 10 CFR 50 Appendix A GDC 19 (General Design Criteria - Control Room).

The proposed AV and NTS for TS Table 3,3.2-1, Function 7.b were determined using an uncertainty methodology previously approved by the NRC for this application. These values provide adequate assurance that required protective and mitigative functions will be initiated as assumed in the transient and accident analyses.

Therefore, there is no significant increase in the probability or consequences of an accident previously evaluated.

TS 3.5.4, "Refueling Water Storage Tank (RWST)," SR 3.5.4.2:

No. The proposed change that increases the RWST borated water volume does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The RWST borated water volume is not an initiator of any accident previously evaluated. As a result, the probability of an accident preViously evaluated is not affected, The proposed change does not alter or prevent the ability of structures, systems, and components from performing their intended El-12

Enclosure 1 Basis for the Proposed Changes safety functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The impact on the containment flood level, equipment qualification, and containment sump pH remains within the limits assumed in the design and accident analyses. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

The proposed change will not alter the operation of, or otherwise increase the failure probability of, any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Based on the above discussions, the proposed TS changes do not involve an increase in the consequences of an accident preViously evaluated.

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

No. The proposed changes do not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. The possibility of a new or different malfunction of safety-related equipment is not created. No new accident scenarios, transient precursors, or limiting single failures are introduced as a result of these changes. There will be no adverse effect or challenges imposed on any safety-related system as a result of these changes.

Based on this evaluation, the proposed changes do not create the possibility of a new or different kind of accident from any accident previolJsly evaluated.

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

No. The proposed changes to the semi-automatic switchover to the containment sump RWST Level - Low Low AV and NTS and to the required RWST minimum borated water volume do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis El-13

Enclosure 1 Basis for the Proposed Changes acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside of the design basis. The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intending function to mitigate the consequences of an initiating event within the applicable acceptance criteria.

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. The minimum and maximum pH values remain bounding; therefore, the required amount of trisodium phosphate (TSP) remains unchanged.

The impact on the containment flood level, equipment qualification, hydrogen produced by the corrosion of galvanized surfaces and zinc based paints, and chloride induced stress corrosion remains within the limits assumed in the design and accident analyses.

There will be no effect on the manner in which the Safety Limits or Limiting Safety System Settings are determined, nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluations, SNC concludes that the proposed changes do not involve a significant hazards consideration under the criteria set forth in 10 CFR 50.92(c) and do not affect the probability of any event initiators. There will be no adverse changes to normal plant operating parameters, ESF actuation Setpoints, or accident mitigation capabilities.

5.2 Applicable Regulatory Requirements/Criteria The ECCS is designed to cool the reactor core and to provide additional shutdown capability following initiation of the following accident conditions:

Loss-of-coolant accident (LOCA) including a pipe break or a spurious relief or safety valve opening in the reactor coolant system (RCS) which would result in a discharge larger than that which could be made up by the normal makeup system.

Loss-of-secondary-coolant accident including a pipe break or a spurious relief or safety valve opening in the secondary steam system which would result in an uncontrolled steam release or a pipe break in the secondary feedwater system.

A steam generator tube rupture accident.

The NRC's regulatory requirements related to the cooling capability of the ECCS during a LOCA are set forth in Title 10 of the Code of Federal Regulations Section 50.46 (10 CFR 50.46), "Acceptance Criteria for ECCS for Light-Water Nuclear Power Reactors." This regulation requires that El-14

Enclosure 1 Basis for the Proposed Changes licensees design their EGGS systems to meet five criteria, one of which is to provide the capability for long-term cooling. Following successful initial operation, the EGGS must possess the capability to remove decay heat such that the core temperature is maintained at an acceptably low value for the extended period of time required by the long-lived radioactivity remaining in the core.

The acceptance criteria for the EGGS, as established by 10 GFR 50.46, will continue to be met:

a. Maximum fuel element cladding temperature is ~ 2200°F;

b. Maximum cladding oxidation is ~ 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is

~ 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;

d. Gore is maintained in a coolable geometry; and

e. Adequate long term core cooling capability is maintained.

The proposed increase in the minimum RWST borated water volume ensures that the EGGS will continue to meet applicable regulatory requirements related to long-term cooling.

A review of "General Design Griteria for Nuclear Power Plants," as specified in Appendix A to 10 GFR Part 50, was conducted to assess the potential impact associated with the proposed changes. The applicable General Design Griteria (GOG) were evaluated as follows:

GOG 1, with respect to quality standards and records

GOG 2, with respect to the RWST

GOG 4, with respect to environmental and dynamic effects design bases

GOG 5, with respect to the RWST

GOG 10, with respect to specified acceptable fuel design limits

GOG 17, with respect to electric power systems

GOG 27, with respect to combined reactivity control systems capability

GOG 30, with respect to quality of reactor coolant pressure boundary

GOG 35, with respect to emergency core cooling

GOG 36, with respect to inspection of emergency core cooling

GOG 37, with respect to testing of emergency core cooling

GOG 54, with respect to systems penetrating containment

GOG 55, with respect to reactor coolant pressure boundary penetrating containment

GOG 56, with respect to primary containment isolation El-15

Enclosure 1 Basis for the Proposed Changes The proposed changes will continue to comply with the regulatory requirements of the above stated General Design Criteria.

The RWST continues to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii): "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier."

5.3 Precedent By letter dated October 2, 2007, Diablo Canyon Power Plant submitted License Amendment Request 07-02, Revision to Technical Specification (TS) 3.5.4, "Refueling Water Storage Tank (RWST)" to the NRC, to increase the minimum required borated water volume from "~ 400,000 gallons (81.5% indicated level)" to"~ 455,300 gallons (93.6% level)." The letter stated that the proposed change is required to meet commitments related to the resolution of issues raised in NRC Generic Letter 2004-02.

By letter dated December 16, 2005, Comanche Peak Steam Electric Station (CPSES) submitted License Amendment Request 05-010, Revision to Technical Specifications 3.3.2, "ESFAS Instrumentation," 3.5.2, "ECCS Operating," and 3.6.7, "Spray Additive System" to the NRC, requesting TS changes required to meet the commitments made in the CPSES response to Generic Letter 2004-02. On October 5, 2006, the NRC issued Amendments No. 129 for Unit 1 and 129 for Unit 2, approving the Comanche Peak LAR (TAC Nos. MC9494 and MC9495).

5.3.1 Relevant NRC RAls The folloWing are relevant NRC Request for Additional Information questions from a recent SNC TS amendment request submittal to the NRC regarding a VEGP P-9 setpoint change. Following each question, is the SNC response relative to the proposed semi-automatic switchover to containment sump - RWST Level- Low Low Setpoint and Allowable Value changes.

To support NRC assessment of the acceptability of the LAR in regard to the setpoint change, SNC is requested to provide the following:

A. Provide documentation of the methodology used for establishing the limiting setpoint (or NSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing. Indicate the related Analytical Limits and other limiting design values (and the sources of these values).

SNC Response:

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Enclosure 1 Basis for the Proposed Changes The basic uncertainty algorithm used to determine the overall instrument uncertainty for the RWST Level- Low Low ESFAS (engineered safety features actuation systems) function is the square-root-sum-of-the squares (SRSS) of the applicable uncertainty terms. This is the same methodology as defined in WCAP-11269, Revision 1, 'Westinghouse Setpoint Methodology for Protection Systems - Vogtle Station" (Reference 7), as modified for term dependencies. This methodology combines the uncertainty components for a channel in an appropriate combination of those groups which are statistically and functionally independent. Those uncertainties which are not independent are conservatively treated by arithmetic summation and then combined via SRSS with the independent terms. This approach is consistent with NRC Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," and ISA Standard 67.04.01-2006, "Setpoints for Nuclear Safety-Related Instrumentation."

RIS 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Speci'fications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24,2006, contains a methodology that is acceptable to the NRC to determine the tolerance for As-Left and As-Found values following surveillance testing. For this instrument channel, Function 7.b, the As-Left and As-Found acceptance criteria are the plant standard, which is currently +/- 0.5%, for the 7300 racks. The Corrective Action Program is used to resolve any values that exceed the acceptance criteria limits.

The related Analytical Limit of 435,522 gallons is the minimum useable RWST volume available before the signal for switchover to cold leg recirculation is initiated, as stated in Section 4.1, "Small Break LOCA Analysis."

B. Provide a statement as to whether or not the [NTS] is a limiting safety system setting on which a safety limit (SL) has been placed as discussed in 10 CFR 50.36(c)(1)(ii)(A). If the [NTS] is not SL Related, explain the basis for this.

SNC Response:

The Nominal Trip Setpoint (NTS) for the semi-automatic switchover to the containment sumps is not a limiting safety system setting (LSSS) on which a SL has been placed as defined in 10 CFR 50.36(c)(1 )(ii)(A).

1. As stated in 10 CFR 50.36(c)(1 )(ii)(A), "Limiting safety system settings for nuclear reactors are settings for automatic El-17

Enclosure 1 Basis for the Proposed Changes protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded."

At VEGP, the semi-automatic switchover is not a completely automatic function by virtue of the fact that operator action is required to complete the switchover. At the end of the safety injection phase of a LOCA, the RWST will have reached the level at which the switchover of ECCS suction to the containment sumps will be automatically initiated. As the RWST continues to empty, operator action as described previously will be required to complete the switchover to the containment sumps. Since operator action is required to complete the switchover, this is not an LSSS on which a SL has been placed.

2. As discussed in VEGP TS Bases B 2.0, "Safety Limits,"

restrictions of the reactor core SL (TS Bases B 2.1.1) prevent overheating of the fuel and cladding, as well as possible cladding perforation that would result in the release of fission products to the reactor coolant. Automatic enforcement of the reactor core SLs is provided by the following functions:

a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;

c. Overtemperature t::.T trip;

d. Overpower t::.T trip;

e. Power Range Neutron Flux trip;

f. Reactor Coolant Flow trips (including undervoltage and underfrequencyof the reactor coolant pump buses); and

g. Main steam safety valves.

The RCS pressure SL (TS Bases B 2.1.2) protects the integrity of the RCS against overpressurization. The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices. More specifically, no credit is taken for operation of the following:

a. Pressurizer power operated relief valves (PORVs);

b. Main steam atmospheric relief valves;

c. Steam Dump System;

d. Rod Control System;

e. Pressurizer Level Control System; or

f. Pressurizer Spray System.

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Enclosure 1 Basis for the Proposed Changes Since the semi-automatic switchover to containment sump RWST Level - Low Low Setpoint is not considered in the above listed fUr;lctions, systems or instruments described in VEGP TS Bases B 2.0, it is not considered SL-Related in accordance with 10 CFR 50.36(c)(1 )(ii)(A).

C. If the [NTS] is determined to be SL-Related, please refer to the NRC letter to the NEI SMTF dated September 7, 2005 (ML052500004) which describes Setpoint-Related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-Related setpoints. Specifically: Part '~" of the Enclosure to the letter provides LCO notes to be added to the TS, and Part "B" includes a check list of the information to be provided in the TS Bases related to the proposed TS change.

1. Describe whether and how the SRTS suggested in the September 7 letter will be implemented. If you do not plan to adopt the suggested SRTS is adopted, then explain how compliance with 10 CFR 50.36 will be assured by addressing items C2 and C3, below.

SNC Response:

This is not applicable based on the answer to Part B above.

2. Describe how surveillance test results and associated TS limits are used to establish operability of the safety system.

Show that this evaluation is consistent with the assumptions and results of the setpoint calculation methodology. Discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the criteria for determining operability of the instrument being tested are located in a document other than the TS (e.g.

plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

SNC Response:

This is not applicable based on the answer to Part B above.

3. Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses. If the controls are located in a document other than the TS (e.g. plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

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Enclosure 1 Basis for the Proposed Changes SNC Response:

This is not applicable based on the answer to Part B above.

D. For setpoints that are determined to be non-SL-related, describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in your discussion information on the controls you employ to ensure that the as left trip setting after completion of periodic surveillance is consistent with your setpoint methodology.

Also, discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the controls are located in a document other than the TS (e.g., plant test procedure), describe how it is ensured that the controls will be implemented.

SNC Response:

For the NTS and AV semi-automatic switchover to containment sump, design calculations ensure that the Setpoint is selected for the correct operation of Plant VogUe. Engineering documents, drawings and manuals detail the Setpoint for site use. Plant operating and calibration procedures incorporate this Setpoint.

The RWST semi-automatic switchover function is monitored for drift.

When the Technicians find an instrument or instrument string out of the acceptable criteria limits, they enter this into the Corrective Action Program. The calibration process requires re-calibration of the instrument string back to within acceptable As-Left limits.

Should the Technician encounter any concerns in this re calibration, the Technician will enter this into the Corrective Action Program and repair the instrument loop. In addition, TS Table 3.3.2-1 contains the following Note, which is applied to each NTS:

"A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provide the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint." The addition of this Note was approved by the NRC in Amendments 101 and 79 to the Unit 1 and Unit 2 TS, respectively.

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Enclosure 1 Basis for the Proposed Changes 5.4 Conclusions In conclusion, in regard to the proposed TS changes and based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 Environmental Consideration This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described above, the proposed change involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change does not involve the installation of any new equipment, or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite.

Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed change does not introduce any new mode of plant operation. The minimum containment sump pH used in calculating the radiological consequences for a LOCA remains bounding.

Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and an environmental assessment of the proposed changes is not required.

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Enclosure 1 Basis for the Proposed Changes

1. NL-07-1969, VEGP U1/U2, "Generic Letter 2004-02 Response Extension Request," L.M. Stinson to NRC, dated December 7,2007.

2. NRC letter dated December 19,2007, S. P. Lingam to T. E. Tynan, "Vogtle Electric Generating Plant, Units 1 and 2 - Generic Letter 2004 02, 'Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors,'

Extension Request Approval (TAC Nos. MC4727, MC4728)."

3. NL-05-1264, Joseph M. Farley Nuclear Plant and Vogtle Electric Generating Plant. "September 2005 Response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," L.M. Stinson to NRC, dated August 31, 2005.

4. NL-06-1275, VEGP U1/U2, "Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," D.E. Grissette to NRC, dated June 22, 2006.

5. NL-06-1483, VEGP U1/U2, "Response to NRC RAI on SNC Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," D.E.

Grissette to NRC, dated July 28, 2006.

6. NRC letter dated September 7,2006, C. Gratton to D. E. Grissette, "Vogtle Electric Generating Plant, Unit 1, Approval of Generic Letter 2004-02 Extension Request (TAC No. MC4727)."

7. WCAP-11269, Revision 1, "Westinghouse Setpoint Methodology for Protection Systems - Vogtle Station," November 1986.

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Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2, "ESFAS Instrumentation," and TS 3.5.4, "Refueling Water Storage Tank (RWST)"

Enclosure 2 Markup of Proposed TS and TS Bases Changes

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 7)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT(i)

6. Auxiliary Feedwater (continued)

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

d. Trip of all Main 1 per pump J SR 3.3.2.6 NA NA Feedwater Pumps

7. Semi-automatic Switchover to Containment Sump

a. Automatic 1,2,3,4(h) 2 C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays

b. Refueling 1,2,3,4 4 K SR 3.3.2.1 Water Storage SR 3.3.2.4 Tank (RWST) SR 3.3.2.7 Level-Low Low SR 3.3.2.8 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and re Safety Injection (continued)

(g) When the Main Feedwater System is operating to supply the SGs.

(h) In MODE 4, only 1 train is required to be OPERABLE to support semi-automatic switcho er for the RHR pump that is req OPERABLE in accordance with Specification 3.5.3, ECCS-shutdown.

(i) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration to rance band provided the Trip Setp int value is conservative with respect to its associated Allowable Value and the channel is rea usted to within the established cali ation tolerance band of the Nominal Trip Setpoinl. A Trip Setpoint may be set more c servative than the Nominal Trip Set int as necessary in response to plant conditions.

~ 216.6 in.

and 213.5 in.

~ 210.4 in.

Vogtle Units 1 and 2 3,3.2-14 Amendment No. ~ (Unit 1)

Amendment No. ~ (Unit 2)

RWST 3.5.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or D AND not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 ------------------NOTE------------------

Only required to be performed when ambient air temperature is < 40°F.

Verify RWST borated water temperature is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

~ 44°F and ~ 116°F.

Verify RWST borated water volume is ~ ~B~

SR 3.5.4.2 gallons. I 7 days

~ 2tOO ppm SR 3.5.4.3 Verify RWST boron concentration is and ~ 2600 ppm. I 7 days SR 3.5.4.4 Verify each sludge mixing pump isoiaon valve 18 months automatically closes on an actual or mulated RWST Low-Level signal.

I 686,000 Vogtle Units 1 and 2 3.5.4-2 Amendment No. ~ (Unit 1)

Amendment No.lHJ (Unit 2)

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE e. Safety Injection - Steam Line Pressure - Low SAFETY ANALYSES. (continued)

LCO, and APPLICABILITY This Function is anticipatory in nature and has a typical lead/lag ratio of 50/5.

Steam Line Pressure - Low must be OPERABLE in MODES 1,2, and 3 (above P - 11) when a secondary side break or stuck open valve could result in the rapid depressurization of the steam lines. This signal may be manually blocked by the operator below the P - 11 setpoint. Below P - 11, feed line break is not a concern.

Inside containment, SLB will be terminated by automatic SI actuation via Containment Pressure - High 1, and outside containment SLB will be terminated by the Steam Line Pressure - Negative Rate - High signal for steam line isolation. This Function is not required to be OPERABLE in MODE 4, 5, or 6 because there is insufficient energy in the secondary side of the unit to cause an accident.

2. Containment Spray Containment Spray provides two primary functions:

1. Lowers containment pressure and temperature after an HELB in containment; and

2. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to:

Ensure the pressure boundary integrity of the containment structure; and

Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure.

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment. Water Is initially drawn from the RWST. When the RWST reaches the Tank Empty setpoint :;0, the 8

(continued)

Vogtle Units 1 and 2 B 3.3.2-14 IRe'.'. 1 9/021

ESFAS Instrumentation B 3.3.2 sWitched over BASES APPLICABLE 2. Containment Spray (continued)

SAFETY ANALYSES, LCO, and spray pump suctions are manually to the containment APPLICABILITY sump if continued containment spray is required. Containment spray is actuated manually or by Containment Pressure High 3.

a. Containment Spray - Manual Initiation The operator can initiate both trains of containment spray at any time from the control room by simultaneously turning the two containment spray actuation handswitches in the same channel. Because an inadvertent actuation of containment spray could have such serious consequences, two switches must be turned simultaneously to initiate containment spray. There are two sets of two switches each in the control room. Each set of two switches is a channel of CS Manual Initiation.

Simultaneously turning the two switches in either channel will actuate both trains of containment spray. Two Manual Initiation switches in each channel are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function.

b. Containment Spray - Automatic Actuation Logic and Actuation elays Automatic actuation logic and actuation relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b. Under specific conditions, a single inoperable actuation relay does not require that the affected automatic actuation logic and actuation relays function be declared inoperable. Specific guidance is provided in this section under the heading "Actuation Relays."

Manual and automatic initiation of containment spray must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and s fficient energy in the primary or secondary systems to pose a threat to containment integrity due to overpressure conditions. Manual initiation is also required in MODE 4, even though automatic actuation is not required. In (continued)

Vogtle Units 1 and 2 B 3.3.2-15 ReVision No. 01

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE b. Phase A Isolation Automatic Actuation Logic and SAFETY ANALYSES, Actuation Relays (continued)

LCO, and APPLICABILITY declared inoperable. Specific guidance is provided in this section under the heading "Actuation Relays."

Manual and automatic initiation of Phase A Containment olation must be OPERABLE in MODES 1,2, and 3, when

~ .

I there is a potential for an accident to occur. Manual initiation is also required in MODE 4 even though automatic actuation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA, but because of the large number of components actuated on a Phase A Containment Isolation, actuation is simplified by the use of the manual actuation handswitches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. In MODES 5 and 6, there is insufficient energy in the primary or secondary systems to pressurize the containment to require Phase A Containment Isolation. There also is adequate time for the operator to evaluate unit conditions and manually actuate individual isolation valves in response to abnormal or accident conditions.

c. Phase A Isolation - Safety Injection Phase A Containment Isolation is also initiated by all Functions that initiate Sl. The Phase A Containment Isolation requirements for these Functions are the same as the requirements for their SI function.

Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

(continued)

Vogtle Units 1 and 2 B 3.3.2-18

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE b. Automatic Switchover to Containment Sump Refueling SAFETY ANALYSES, Water Storage Tank (RWST) Level Low Low Coincident LCO, and With Safety Injection (continued)

APPLICABILITY automatically. The operator must complete the switchover by manually closing the RWST suction valves.

The RWST is equipped with four level transmitters. These transmitters provide no control functions. Therefore, a two-out-of-four logic is adequate to initiate the protection function actuation. Although only three channels would be

, including instrument t---"C:1"T1'ti'cj'ent, a fourth channel has been added for increased uncertalnity. The reliability.

The setpoints for this fu on Table 3.3.2-1 are in I rom the RWST base. . setpoint Is equivalen . % of instrumin~t~s~~~~~

Allowable V . equivalent to of instrument r---------,

~ 28.5 % and S 295 %

s ...---

The transmitters are located in an area not affected by HELBs or post accident high radiation. Thus, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties.

Semi-Automatic switchover occurs only if the RWST low low level signal is coincident with SI. This prevents accidental switchover during normal operation. Accidental switchover could damage ECCS pumps if they are attempting to take suction from an empty sump. The automatic switchover Function requirements for the SI Functions are the same as the requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and reqUirements.

These Functions must be OPERABLE in MODES 1, 2, 3, and 4 when there is a potential for a LOCA to occur, to ensure a continued supply of water for (continued)

Vogtle Units 1 and 2 B 3.3.2-30 IRevision No g

RWST B 3.5.4 BASES ACTIONS E.1 and E.2 (continued)

Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 REQUIREMENTS (TI-10982)

The RWST borated water temperature should be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be within the limits assumed in the accident analyses band. This Frequency is sufficient to identify a temperature change that would approach either limit and has been shown to be acceptable through operating experience.

The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperature is ~ 40°F.

With ambient air temperatures ~ 40°F, the RWST temperature should not exceed the limits. Since ambient air temperatures do not exceed the RWST upper temperature limit, the requirement to verify RWST temperature only when the ambient temperature is below 40°F is acceptable.

SR 3.5.4.2 (L1-0990A&B, L1-0991 A&B, L1-0992A, Ll-0993 686,000 The ~tI§J114781 gallons",.,.,""""---r...-...----:---.

should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is normally stable and is protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience.

(contin ued)

Vogtle Units 1 and 2 B 3.5.4-7 ISeVISlon No. 01

Containment Spray and Cooling Systems B 3.6.6 BASES BACKGROUND Containment Spray System (continued)

The Containment Spray System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature and to reduce fission products from the containment atmosphere during a DBA. The RWST solution temperature is an important factor in determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat is removed from the containment sump water by the residual heat removal coolers.

Each train of the Containment Spray System provides adequate spray coverage to meet the system design requirements for containment heat removal.

The Containment Spray System is actuated either automatically by a containment High-3 pressure signal or manually. An automatic actuation opens the containment spray pump discharge valves, starts the two containment spray pumps, and begins the injection phase. A manual actuation of the Containment Spray System requires the operator to actuate two separate switches on the main control board to begin the same sequence. The injection hase continues until an RWST empty tank level alarm is rece' % level). When the RWST level reaches th an level, the operator manually aligns the 0 e recirculation mode. The Containment Spray em in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operating procedures.

Containment Cooling System Two trains of containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement, are provided. Each train of four fan units is supplied with cooling water from a separate train of nuclear service cooling water (NSCW). Air is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield in the lower areas of containment.

(continued)

Vogtle Units 1 and 2 B 3.6.6-2 IRev. 1 9/021

Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2, "ESFAS Instrumentation," and TS 3.5.4, "Refueling Water Storage Tank (RWST)"

Enclosure 3 Typed Pages for TS and TS Bases Change

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 7)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT(i)

6. Auxiliary Feedwater (continued)

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

d. Trip of all Main 1 per pump J SR 3.3.2.6 NA NA Feedwater Pumps

7. Semi-automatic Switchover to Containment Sump

a. Automatic 1,2,3,4(h) 2 C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays

b. Refueling 1,2,3,4 4 K SR 3.3.2.1 S216.6 in. 213.5 in.

Water Storage SR 3.3.2.4 and Tank (RWST) SR 3.3.2.7 <!; 210.4 in.

Level-Low Low SR 3.3.2.8 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection (continued)

(g) When the Main Feedwater System is operating to supply the SGs.

(h) In MODE 4, only 1 train is required to be OPERABLE to support semi-automatic switchover for the RHR pump that is required to be OPERABLE in accordance with Specification 3.5.3, ECCS-shutdown.

(i) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip 5etpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoinl. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3.2-14 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RWST 3.5.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or D AND not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 -------------------NOTE------------------------

Only required to be performed when ambient air temperature is < 40°F.

Verify RWST borated water temperature is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

~ 44°F and::; 116°F.

SR 3.5.4.2 Verify RWST borated water volume is ~ 686,000 7 days gallons.

SR 3.5.4.3 Verify RWST boron concentration is ~ 2400 ppm 7 days and::; 2600 ppm.

SR 3.5.4.4 Verify each sludge mixing pump isolation valve 18 months automatically closes on an actual or simulated RWST Low-Level signal.

Vogtle Units 1 and 2 3.5.4-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE e. Safety Injection - Steam Line Pressure - Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY This Function is anticipatory in nature and has a typical lead/lag ratio of 50/5.

Steam Line Pressure Low must be OPERABLE in MODES 1, 2, and 3 (above P - 11) when a secondary side break or stuck open valve could result in the rapid depressurization of the steam lines. This signal may be manually blocked by the operator below the P - 11 setpoint. Below P - 11, feed line break is not a concern.

Inside containment, SLB will be terminated by automatic SI actuation via Containment Pressure High 1, and outside containment SLB will be terminated by the Steam Line Pressure Negative Rate High signal for steam line isolation. This Function is not required to be OPERABLE in MODE 4, 5, or 6 because there is insufficient energy in the secondary side of the unit to cause an accident.

2. Containment Spray Containment Spray provides two primary functions:

1. Lowers containment pressure and temperature after an HELB in containment; and

2. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to:

Ensure the pressure boundary integrity of the containment structure; and

Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure.

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment. Water is initially drawn from the RWST. When the RWST reaches the Tank Empty setpoint 8%, the (continued)

Vogtle Units 1 and 2 B 3.3.2-14

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 2. Containment Spray (continued)

SAFETY ANALYSES, LCO, and spray pump suctions are manually switched over to the APPLICABILITY containment sump if continued containment spray is required.

Containment spray is actuated manually or by Containment Pressure High 3.

a. Containment Spray - Manual Initiation The operator can initiate both trains of containment spray at any time from the control room by simultaneously turning the two containment spray actuation handswitches in the same channel. Because an inadvertent actuation of containment spray could have such serious consequences, two switches must be turned simultaneously to initiate containment spray. There are two sets of two switches each in the control room. Each set of two switches is a channel of CS Manual Initiation.

Simultaneously turning the two switches in either channel will actuate both trains of containment spray. Two Manual Initiation switches in each channel are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function.

b. Containment Spray - Automatic Actuation Logic and Actuation Relays Automatic actuation logic and actuation relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b. Under specific conditions, a single inoperable actuation relay does not require that the affected automatic actuation logic and actuation relays function be declared inoperable. Specific guidance is provided in this section under the heading "Actuation Relays."

Manual and automatic initiation of containment spray must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and sufficient energy in the primary or secondary systems to pose a threat to containment integrity due to overpressure conditions. Manual initiation is also required in MODE 4, even though automatic actuation is not required. In (continued)

Vogtle Units 1 and 2 B 3.3.2-15

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE b. Phase A Isolation - Automatic Actuation Logic and SAFETY ANALYSES, Actuation Relays (continued)

LCO, and APPLICABILITY declared inoperable. Specific guidance is provided in this section under the heading "Actuation Relays."

Manual and automatic initiation of Phase A Containment isolation must be OPERABLE in MODES 1, 2, and 3, when there is a potential for an accident to occur. Manual initiation is also required in MODE 4 even though automatic actuation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA, but because of the large number of components actuated on a Phase A Containment Isolation, actuation is simplified by the use of the manual actuation handswitches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. In MODES 5 and 6, there is insufficient energy in the primary or secondary systems to pressurize the containment to require Phase A Containment Isolation. There also is adequate time for the operator to evaluate unit conditions and manually actuate individual isolation valves in response to abnormal or accident conditions.

c. Phase A Isolation - Safety Injection Phase A Containment Isolation is also initiated by all Functions that initiate SI. The Phase A Containment Isolation requirements for these Functions are the same as the requirements for their SI function.

Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

(continued)

Vogtle Units 1 and 2 B 3.3.2-18

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE b. Automatic Switchover to Containment Sump Refueling SAFETY ANALYSES, Water Storage Tank (RWST) Level- Low Low Coincident LCO, and With Safety Injection (continued)

APPLICABILITY automatically. The operator must complete the switchover by manually closing the RWST suction valves.

The RWST is equipped with four level transmitters. These transmitters provide no control functions. Therefore, a two-out-of-four logic is adequate to initiate the protection function actuation. Although only three channels would be sufficient, a fourth channel has been added for increased relia bility.

The setpoints for this function on Table 3.3.2-1 are in inches from the RWST base. The trip setpoint is equivalent to 29.0% of instrument span, including instrument uncertainty. The Allowable Values are equivalent to ~ 28.5% and S 29.5 % of instrument span.

The transmitters are located in an area not affected by HELBs or post accident high radiation. Thus, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties.

Semi-Automatic switchover occurs only if the RWST low low level signal is coincident with SI. This prevents accidental switchover during normal operation. Accidental switchover could damage ECCS pumps if they are attempting to take suction from an empty sump. The automatic switchover Function requirements for the SI Functions are the same as the requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

These Functions must be OPERABLE in MODES 1, 2, 3, and 4 when there is a potential for a LOCA to occur, to ensure a continued supply of water for (continued)

Vogtle Units 1 and 2 B 3.3.2-30

RWST B 3.5.4 BASES ACTIONS E.1 and E.2 (continued)

Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 REQUIREMENTS (TI-10982)

The RWST borated water temperature should be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be within the limits assumed in the accident analyses band. This Frequency is sufficient to identify a temperature change that would approach either limit and has been shown to be acceptable through operating experience.

The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperature is ~ 40°F.

With ambient air temperatures ~ 40°F, the RWST temperature should not exceed the limits. Since ambient air temperatures do not exceed the RWST upper temperature limit, the requirement to verify RWST temperature only when the ambient temperature is below 40°F is acceptable.

SR 3.5.4.2 (L1-0990A&B, L1-0991A&B, L1-0992A, L1-0993A)

The RWST water volume (686,000 gallons, 95% indicated level) should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is normally stable and is protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience.

(continued)

Vogtle Units 1 and 2 B 3.5.4-7

Containment Spray and Cooling Systems B 3.6.6 BASES BACKGROUND Containment Spray System (continued)

The Containment Spray System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature and to reduce fission products from the containment atmosphere dUring a DBA. The RWST solution temperature is an important factor in determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat is removed from the containment sump water by the residual heat removal coolers.

Each train of the Containment Spray System prOVides adequate spray coverage to meet the system design requirements for containment heat removal.

The Containment Spray System is actuated either automatically by a containment High-3 pressure signal or manually. An automatic actuation opens the containment spray pump discharge valves, starts the two containment spray pumps, and begins the injection phase. A manual actuation of the Containment Spray System requires the operator to actuate two separate switches on the main control board to begin the same sequence. The injection phase continues until an RWST empty tank level alarm is received (8% level). When the RWST level reaches the empty tank level, the operator manually aligns the system to the recirculation mode. The Containment Spray System in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operating procedures.

Containment Cooling System Two trains of containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement, are provided. Each train of four fan units is supplied with cooling water from a separate train of nuclear service cooling water (NSCW). Air is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield in the lower areas of containment.

(continued)

Vogtle Units 1 and 2 B 3.6.6-2

v t e Letter Sequence Request
TAC:MD7772, (Approved, Closed)
Initiation Request Acceptance... Administration Questions 28 January 2008 10 March 2008 Answers 6 February 2008 5 March 2008 22 May 2008 Results Approval Other: ML080880198
MONTHYEAR2008-03-10 [Table View]

NL-07-2168, License Amendment Request to Revise Technical Specifications (TS) 3.3.2, ESFAS Instrumentation, and TS 3.5.4, Refueling Water Storage Tank (Rwst).

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NL-07-2168, License Amendment Request to Revise Technical Specifications (TS) 3.3.2, ESFAS Instrumentation, and TS 3.5.4, Refueling Water Storage Tank (Rwst).

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