ML11161A143

License Amendment Request for Approval of a Revision to the South Texas Project Fire Protection Program Related to the Alternative Shutdown Capability

ML11161A143
Person / Time
Site:	South Texas
Issue date:	06/02/2011
From:	Gerry Powell South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-11002643
Download: ML11161A143 (44)

Text

Nuclear Operating Company South Texas Project Electric Generating Station PO. Box 289 Wadsworth Texas 77483 June 2, 2011 NOC-AE-I 1002643 G25 10CFR50.90 10CFR50.48 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 License Amendment Request for Approval of a Revision to the South Texas Project Fire Protection Program Related to the Alternative Shutdown Capability Pursuant to 10 CFR 50.90, STP Nuclear Operating Company (STPNOC) hereby requests a licensee amendment for approval of a revision to the South Texas Project (STP) Fire Protection Program (FPP) related to the Alternative Shutdown Capability. Specifically, STPNOC requests the crediting of the performance of additional operations in the control room, including one automatic operation, in the event a fire requires evacuation. These operations will ensure compliance with Section III.L of 10 CFR 50, Appendix R. STPNOC has determined that reliance on the additional operations potentially create an adverse affect on the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, per Condition 2.E of the each Unit's Operating License, STPNOC is requesting approval by the Nuclear Regulatory Commission for making the changes to the STP FPP. The attached safety evaluation demonstrates that no significant hazards will result from this change.

This request provides the proposed change to the Operating Licenses as Attachments 1 and 2 of the enclosure to this letter.

The STPNOC Plant Operations Review Committee has reviewed and concurred with the proposed change.

AOO STI: 32827667

NOC-AE-1 1002643 Page 2 In accordance with 10 CFR 50.91(b), STPNOC is notifying the State of Texas of this request for license amendment by providing a copy of this letter and its enclosure.

Upon approval of this request, the approved change will be documented in the Fire Hazards Analysis Report. See licensing commitment described in Attachment 4 to the Enclosure. There are no other commitments in this letter.

It is requested that this license amendment request be approved by May 30, 2012 with a 60 day implementation period to provide time to revise STPNOC licensing documents.

If there are any questions regarding this amendment request, please contact Ken Taplett at (361) 972-8416 or me at (361) 972-7566.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 3Zue a. got(

G. T. Powell Vice President, Technical Support & Oversight

Enclosure:

Evaluation of the Proposed Change

NOC-AE-l 1002643 Page 3 cc:

(paper copy)

(electronic copy)

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-4125 Balwant K, Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8B1) 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Catherine Callaway Jim von Suskil NRG South Texas LP Ed Alarcon Kevin Pollo Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.

C. Mele City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services

NOC-AE-1 1002643 ENCLOSURE Evaluation of the Proposed Change

Subject:

License Amendment Request for approval of a revision to the South Texas Project (STP) Fire Protection Program (FPP) related to the Alternative Shutdown Capability.

1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

6.0 REFERENCES

Figure 1 - Control Room Control Panel Switch Orientation Figure 2 - Turbine Trip Circuitry Logic Diagram ATTACHMENTS:

1. Proposed Change to South Texas Project, Unit 1, Operating License No. NPF-76

2. Proposed Change to South Texas Project, Unit 2, Operating License No. NPF-80

3. Annotated Fire Hazards Analysis Report Page

4. List of Commitments

Enclosure NOC-AE-11002643 Page 1 of 29 Evaluation of the Proposed Change 1.0 Summary Description This evaluation supports a request to revise the South Texas Project (STP) Fire Protection Program (FPP) related to the Alternative Shutdown Capability.

The proposed change credits the performance of additional operations in the control room, including one automatic operation, in the event a fire requires evacuation. A thermal-hydraulic analysis demonstrates that these operations will ensure that the reactor coolant system (RCS) process variables remain within those values predicted for a loss of normal a-c power, as required by Section III.L. 1. of 10 CFR 50, Appendix R.

It is requested that this license amendment request be approved by May 30, 2012 with a 60 day implementation period. The proposed schedule is to allow the Nuclear Regulatory Commission (NRC) sufficient time to review and approve this amendment request and for STP Nuclear Operating Company (STPNOC) to revise licensing documents.

STPNOC previously submitted a license amendment request on February 4, 2008 (Reference 6.1) for a similar change. That amendment request credited the Electric Power Research Institute (EPRI)/Nuclear Energy Institute (NEI) cable fire tests (Ref 6.2) to provide a technical basis for concluding that the STP circuit cable design should not fail due to fire and result in spurious actuations during the time period it took to isolate the circuits in the control room and back-up the control room actions from stations outside the control room. That amendment request concluded that the operator actions were feasible and reliable. The amendment request was subsequently withdrawn (Reference 6.3) by STPNOC because it was determined that it was inappropriate to generically apply the NEI cable fire tests to a plant-specific application.

This license amendment request does not take credit for the EPRI/NEI cable fire tests. This amendment request demonstrates that a single spurious operation that could negate an action performed in the control room prior to isolating the affected circuit by transfer to alternate control stations will not result in the RCS process variables deviating beyond those values predicted for a loss of normal a-c power. An evaluation is provided to demonstrate that the operator actions performed in the control room prior to evacuation are feasible and can be reliably completed.

Enclosure NOC-AE-l 1002643 Page 2 of 29 2.0 Detailed Description 2.1 Current STP Alternative Shutdown Capability Alternate shutdown capability is provided to respond to a fire occurring within the main control room that results in evacuation. Following reactor trip from the control room, the transfer of control from the control room to the auxiliary shutdown panel and local control stations is accomplished from outside the control room using transfer switches. When transferred, these circuits are independent of and electrically isolated from the control room circuits. See Section 3.6 for more description of the circuit isolation capability.

The alternate shutdown capability provides the controls and direct reading indications to monitor the process variables necessary to perform reactivity control, reactor coolant makeup and inventory control, and reactor heat removal.

The NRC review of the STP alternative shutdown capability is documented in NUREG-078 1, "Safety Evaluation Report related to the operation of the South Texas Project, Units 1 and 2."

The following discussion regarding the performance of actions prior to control room evacuation is documented in Supplement 2 to NUREG-0781, dated January 1987 (Reference 6.4) as follows:

In addition to scramming the reactor from the control room, the applicant has included procedures for other actions that are to be performed before the control room is evacuated.

These actions, however, can be performed outside the control room regardless of circuit damage within the control room. They include tripping the reactor coolant pumps, closing the PORV block valves, isolating the steam generators, and securing the charging pumps. The above actions could prevent a very unlikely series of events, which include spurious actuations, the failure of specific automatic functions, and the operation of other specific automatic functions, from causing.RCS process variables to exceed those limits predicted for a loss of normal ac power.

The current description of the alternate shutdown capability is provided in Section 2.4.4 of the FHAR. Section 2.4.4. states:

Alternate shutdown capability is provided to respond to a large fire occurring within the main control room. Following reactor trip from the control room, the transfer of control from the control room to the auxiliary shutdown panel and local control stations is accomplished from outside the control room using transfer switches which are predominately located in the three redundant switchgear rooms. The remaining transfer switches are on the auxiliary shutdown panel in the train related diesel generator rooms and at the Essential Cooling Water Intake Structure ventilation fan MCCs. When transferred, these circuits are independent of the control room.

Enclosure NOC-AE-1 1002643 Page 3 of 29 2.2 Proposed Revision to STP Alternative Shutdown Capability This change proposes to credit the performance of the following manual operator actions in the control room prior to evacuation due to a fire for meeting the alternative shutdown capability.

1. Main steam line isolation

2. Closing the pressurizer power-operated relief valves (PORV) block valves

3. Securing all reactor coolant pumps

4. Feedwater isolation

5. Securing the startup feedwater pump

6. Letdown isolation

7. Securing the charging pumps In addition, this change proposes to credit the automatic trip of the main turbine upon the initiation of a manual reactor trip for meeting the alternative shutdown capability.

During a triennial inspection (Reference 6.5) of the STP Fire Protection Program (FPP),

STPNOC received a non-cited violation of Section III.L. 1 to 10 CFR Part 50, Appendix R because the facility's thermal-hydraulic analysis to demonstrate the alternative shutdown capability was inconsistent with actions credited in the STP FPP licensing basis. Specifically, the analysis inappropriately allowed four additional manual actions to be performed from the control room while the licensing basis only credited one manual action (i.e. initiation of a reactor trip) to be performed prior to evacuating the control room. Performing the additional actions inside the control room ensures that the RCS process variables remained within those values predicted for a loss of normal a-c power. The proposed amendment request is submitted to change the FPP licensing basis.

The STP FPP is described in the Fire Hazards Analysis Report (FHAR). The STP Alternate Shutdown Capability is described in Section 2.4.4 of the FHAR (See Section 2.1). Section 2.4.4 only credits the operator manual action to trip the reactor from the control room prior to evacuation. In addition, the STP fire safe shutdown analysis does not assume the occurrence of automatic operations within the fire area unless the automatic operation adversely impacts the response to the fire.

Section 5.4.4 of NRC Regulatory Guide (RG) 1.189, Revision 2, Fire Protection for Nuclear Power Plants, (Reference 6.6) states:

The only operator action in the control room before evacuation for which credit is usually given is reactor trip. For any additional control room actions deemed necessary before evacuation, a licensee should be able to demonstrate that such actions can be performed.

Additionally, the licensee should ensure that such actions cannot be negated by subsequent spurious actuation signals resulting from the postulated fire. The design basis for the control room fire should consider one spurious actuation [emphasis added] or signal to occur before

Enclosure NOC-AE-1 1002643 Page 4 of 29 control of the plant is achieved through the alternative or dedicated shutdown system. After control of the plant is achieved by the alternative or dedicated shutdown system, single or multiple spurious actuations that could occur in the fire-affected area should be considered, in accordance with the plant's approved FPP.

The proposed change credits the performance of certain operations within the control room until the operations are backed up from stations outside the control room. The operations are backed up outside the control room with alternative circuits by transferring control to local control stations outside of the control room as allowed by Appendix R,Section III.G.3. The transfer electrically isolates the circuits in the control room from the alternative shutdown circuits so that any circuit failures in the control room after transfer will not adversely affect the safe shutdown function. The proposed change assumes one spurious actuation to occur before control of the plant is achieved through the alternative or dedicated shutdown system. In addition, the proposed change credits an automatic trip of the main turbine upon initiation of the manual reactor trip.

In addition to manually tripping the reactor, the following operator manual actions are performed within the control room prior to evacuation:

1. Main steam line isolation

2. Closing the pressurizer power-operated relief valves (PORV) block valves

3. Securing all reactor coolant pumps

4. Feedwater isolation

5. Securing the startup feedwater pump

6. Letdown isolation

7. Securing the charging pumps Backup actions are performed outside the control room for the above actions within 10 minutes of initiating a reactor trip with the exception of backing up the tripping of the reactor coolant pumps which are performed within 30 minutes following reactor trip. The backup action to de-energize the reactor coolant pumps is not time critical since multiple spurious actuations would have to occur for an adverse impact to occur. Although the backup actions are performed outside the control room, the design basis thermal-hydraulic analysis (Reference 6.7) credits the actions performed prior to evacuation of the control room. In addition, the design basis thermal-hydraulic analysis demonstrates that any single spurious operation that negates an action performed in the control room will not result in RCS process variables beyond those values predicted for a loss of normal a-c power.

The annotated FHAR pages affected by this proposed change are provided in Attachment 3.

Upon approval of this request, the approved change will be documented in the STP FHAR. See the Licensing Commitment described in Attachment 4.

Enclosure NOC-AE-1 1002643 Page 5 of 29 3.0 Technical Evaluation 3.1 Regulatory Requirements The proposed change credits operations from the control room prior to evacuation in addition to a manual reactor trip. Performance of these actions ensure compliance with Section III.L of 10 CFR 50, Appendix R. The specific sections of this regulation addressed by the manual actions in the control room are:

III.L. 1 During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal a.c. power, and the fission product boundary integrity shall not be affected; i.e., there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary.

III.L. L.b Alternative or dedicated shutdown capability provided for a specific fire area shall be able to maintain reactor coolant inventory III.L.2.a The reactivity control function shall be capable of achieving and maintaining cold shutdown reactivity conditions.

III.L.2.b The reactor coolant makeup function shall be capable of maintaining the reactor coolant level above the top of the core for BWRs and be within the level indication in the pressurizer for PWRs.

III.L.5 Equipment and systems comprising the means to achieve and maintain cold shutdown conditions shall not be damaged by fire; or the fire damage to such equipment and systems shall be limited so that the systems can be made operable and cold shutdown can be achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Materials for such repairs shall be readily available on site and procedures shall be in effect to implement such repairs. If such equipment and systems used prior to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the fire will not be capable of being powered by both onsite and offsite electric power systems because of fire damage, an independent onsite power system shall be provided. Equipment and systems used after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be powered by offsite power only.

3.2 Description of the Operator Manual Actions Credited in the Control Room: The following operator manual actions are performed within the control room prior to evacuation in the event of a fire:

• Trip the reactor

• Initiate main steam isolation

• Close both pressurizer power-operated relief valve (PORV) block valves

• Trip all reactor coolant pumps

• Initiate feedwater isolation

Enclosure NOC-AE-1 1002643 Page 6 of 29

• Place the startup feedwater pump in PULL TO LOCK (i.e., secure feedwater)

• Isolate letdown Place centrifugal charging pumps in PULL TO LOCK (i.e., secure charging)

An evaluation of the control room operator actions was performed to demonstrate they are feasible and reliable. The evaluation assumes one spurious actuation or signal can occur before control of the plant is achieved through the alternative shutdown system. After control of the plant is achieved by the alternative shutdown system, single or multiple spurious actuations that could occur in the fire-affected area will not affect the fire safe shutdown capability because the circuits in the fire area (i.e. the control room) are electrically isolated from the circuits at the alternative shutdown stations. In addition, a defense-in-depth analysis was performed of the control room fire protection features.

Additional thermal-hydraulic-analysis was performed for selected actions. The thermal-hydraulic analysis quantified the impact of not performing these actions against the requirements of Appendix R. The actions that were evaluated are:

" Close both pressurizer power-operated PORV block valves

• Initiate feedwater isolation

" Place startup feedwater pump in PULL-TO-LOCK

" Place centrifugal charging pumps in PULL-TO-LOCK The thermal-hydraulic analysis was performed using the RETRAN computer code. The computer model used is similar to the model described in WCAP 14882-P-A (Reference 6.8).

The model was modified to reflect nominal conditions and setpoints as opposed to the worst case conditions and setpoints assumed in the safety analysis.

The additional actions to be credited are required for the following reasons:

Initiate main steam isolation-The evaluation credits the operator manual action to initiate main steam isolation prior to leaving the control room. This action is accomplished by the manipulation of a single switch in the control room (see Figure 1). Operators have demonstrated that these actions can be performed in less than 30 seconds of tripping the reactor. The purpose of closing these valves is to ensure an uncontrolled cool down of the RCS does not occur resulting in the indicated pressurizer water level going off-scale low in the event of a fire induced spurious opening of a secondary steam side valve such as the steam dump system. This manual action is required because the indicated pressurizer level going off-scale low is considered a violation of Appendix R,Section III.L.2.b.

Close both pressurizer power-operated PORV block valves - The thermal-hydraulic analysis credits the operator manual action to close the pressurizer PORV block valves. This action is accomplished by the manipulation of two switches in the control room (see Figure 1).

Operators have demonstrated that these actions can be performed in less than 60 seconds of tripping the reactor. The purpose of closing these valves is to protect against an uncontrolled

Enclosure NOC-AE-1 1002643 Page 7 of 29 depressurization of and loss of RCS inventory in the event of a fire induced spurious opening of a pressurizer PORV. With the pressurizer PORV block valves open, a fire induced spurious opening of the pressurizer PORVs would result in an Engineered Safety Features (ESF) safety injection signal in approximately 71 seconds due to a pressurizer pressure low signal. An ESF safety actuation signal is considered a process variable outside what would be predicted for a normal loss of a-c power. Therefore, this manual action is required to ensure compliance with Appendix R,Section III.L. 1.

Trip all reactor coolant pumps (RCP) -The evaluation credits the operator manual action to trip all RCPs. This action is accomplished by the manipulation of all four switches in the control room (see Figure 1). Operators have demonstrated that these actions can be performed in less than 120 seconds of tripping the reactor. The purpose of securing the RCPs is to ensure the RCS does not undergo an uncontrolled depressurization if a fire-induced circuit failure causes a pressurizer spray valve to spuriously open. With the RCPs running, a fire induced spurious opening of a pressurizer spray valve could result an ESF safety injection signal due to a pressurizer pressure low signal. An ESF safety actuation signal is considered a process variable outside what would be predicted for a normal loss of a-c power event and therefore a violation of Appendix R,Section III.L. 1. In addition, the centrifugal charging pumps are secured which provide seal water to the RCPs. Tripping the RCPs protects the RCP seals, which ensures compliance with Appendix R,Section III.L. 1 by protecting the primary coolant boundary. Therefore, this manual action is required to ensure compliance with Appendix R,Section III.L. 1 Initiate feedwater isolation - The thermal-hydraulic analysis credits the operator manual action to close all four main feedwater isolation valves. This action is accomplished by the manipulation of four switches in the control room (see Figure 1). Operators have demonstrated that these actions can be performed in less than 120 seconds of tripping the reactor. The purpose of closing the feedwater isolation valves is to protect against a spurious fire induced start of the startup feedwater pump. With the feedwater isolation valves open and a spurious fire induced start of the startup feedwater pump, the pressurizer level would go off scale low in approximately 4 minutes due to over cooling of the RCS. The pressurizer level going off-scale low would be a violation of Appendix R,Section III.L.2.b. In addition, the steam generator water level would go off scale high in approximately 4 minutes due to overfilling the steam generator. Overfilling the steam generators is considered a process variable outside what would be predicted for a normal loss of a-c power, which is a violation of Appendix R,Section III.L. 1. Therefore, this manual action is required to ensure compliance with Appendix R, Sections III.L. 1 and III.L.2.b.

Place startup feedwater pump in PULL-TO-LOCK - The thermal-hydraulic analysis credits the operator manual action to place the startup feedwater pump in PULL-TO-LOCK. This action is accomplished by manipulation of a hand switch in the control room (see Figure 1).

Operators have demonstrated that these actions can be performed in less than 120 seconds of tripping the reactor. The purpose of placing the startup feedwater pump in PULL-TO-LOCK is to prevent shrinking the indicated pressurizer water level off-scale low due to over cooling

Enclosure NOC-AE-l 1002643 Page 8 of 29 the RCS and to prevent over filling the steam generator due to a fire induced spurious opening of a main feedwater isolation valve.

With a loss of off-site power, electrical power is lost to the startup feedwater pump. The source of power to the startup feedwater pump is not diesel-backed. Therefore, the startup feedwater pump is not available to provide feedwater flow and this action is not required.

With offsite power available, auto starting of the startup feedwater pump will occur when the operators close the main steam isolation valves. Closing the main steam isolation valves will trip the main feedwater pump turbines because closing these valves secures steam to these turbines. The startup feedwater pump will then auto start due to a main feedwater pump turbine trip with offsite power available.

With the startup feedwater pump auto starting, a fire induced opening of a main feedwater isolation valve would result in the steam generator water level off scale high in approximately two minutes. The steam generator water level off scale high is considered a process variable outside what would be predicted for a normal loss of a-c power, a violoation of Appendix R,Section III.L. 1. In addition, the pressurizer water level may go off scale low in approximately 5 minutes after a reactor trip, which is a violation of Appendix R,Section III.L.2.b. Therefore, this manual action is required to ensure compliance with Appendix R, Sections III.L. 1 and III.L.2.b.

Isolate letdown - The evaluation credits the operator manual action to isolate letdown. This action is accomplished by the manipulation of two switches in the control room (see Figure 1). Operators have demonstrated that these actions can be performed in less than 120 seconds of tripping the reactor. The purpose of isolating letdown is to reduce the possibility of an uncontrolled loss of RCS inventory because charging (i.e. RCS makeup) is secured. An uncontrolled loss of RCS inventory is considered a violation of Appendix R,Section III.L. 1.b. This action also ensures the pressurizer level does not go off scale low due to fire induced spurious actions. Maintaining the pressurizer level on scale is required by Appendix R,Section III.L.2.b. Therefore, this manual action is required to ensure compliance with Appendix R Sections III.L..b and III.L.2.b.

Place centrifugal charging pumps in PULL-TO-LOCK - The thermal-hydraulic analysis credits the operator manual action to secure the charging pumps. This action is accomplished by the manipulation of two switches in the control room (see Figure 1). Operators have demonstrated that these actions can be performed in less than 120 seconds of tripping the reactor. The purpose of securing the charging pumps is to protect these pumps for use later in the event, prevent an uncontrolled depressurization of the RCS, and overfilling the pressurizer in the event of a fire induced spurious action. The charging pumps are required to achieve and maintain cold shutdown conditions as required by Appendix R, Sections III.L.2.a and III.L.5. An uncontrolled depressurization of the RCS is considered a process variable outside what would be predicted for a normal loss of a-c power, a violation of Appendix R,Section III.L. 1. The overfilling of the pressurizer is considered a violation of Appendix R,Section III.L.2.b. With the centrifugal charging pumps running, a fire induced

Enclosure NOC-AE-1 1002643 Page 9 of 29 spurious closure of the Volume Control Tank isolation valves would result in damage to these pumps. With the centrifugal pumps running, a fire induced spurious opening of a pressurizer spray valve would result in an ESF SI signal in approximately 2 minutes and pressurizer level off scale high in 8 minutes. Therefore, this manual action is required to ensure compliance with Appendix R, Sections III.L.1, III.L.2.a, III.L.2.b, and III.L.5.

Although the time to decide to evacuate the control room may be limited because of the unpredictability of the time from detection of a fire to the time the fire progresses to the point where the control room must be evacuated, operators will be pre-alerted to perform these actions because the actions are not required until the reactor is manually tripped. Operator walk down performance demonstrates that these actions can be performed in rapid succession following the initiation of the reactor trip to support the time line assumed in the evaluation and thermal-hydraulic analysis.

3.3 Description of the Automatic Turbine Trip This proposed change credits the automatic turbine trip upon a manual trip of the reactor. The crediting of automatic functions in the fire area is not normally allowed in fire safe shutdown analysis. STP reactor trip circuitry generates a signal to trip the main turbine upon initiation of a reactor trip. Once tripped, a fire-induced circuit failure will not result in a restart of the turbine.

Since the performance of a manual reactor trip has already been approved as part of the STP licensing basis, it is reasonable that the credited reactor trip circuitry would successfully function to trip the main turbine in addition to tripping the reactor. Additional justification for crediting this feature is provided in Section 3.5. An automatic turbine trip will prevent a rapid cool down of the RCS so that pressurizer level remains within the indicating range.

3.4 Single Spurious Actuation Analysis Any additional control room actions deemed necessary before evacuation should not be negated by subsequent spurious actuation signals resulting from the postulated fire such that the requirements of Appendix R are satisfied. The design basis for the control room fire should consider one spurious actuation or signal to occur before control of the plant is achieved through the alternative or dedicated shutdown system. After control of the plant is achieved by the alternative shutdown system, single or multiple spurious actuations that could occur in the fire-affected area should be considered in accordance with the plant's approved FPP.

The following provides an analysis of the results of a single spurious actuation effect on each of the control room actions before the action can be backed up by an action performed outside of the control room.

3.4.1 One Main Steam Isolation Valve Spuriously Opens.

A spurious opening of a main steam isolation valve could potentially result in an uncontrolled cool down of the RCS. However, an uncontrolled cool down of the RCS will not occur since a second fire induced spurious opening of other sources of steam relief such as the steam dump

Enclosure NOC-AE-1 1002643 Page 10 of 29 system and main turbine would be required for such a cool down. Therefore, the spurious opening of a main steam isolation valve will not result in an uncontrolled cool down of the RCS.

A loss of off-site power does not change the results of the evaluation for this scenario.

3.4.2 One Pressurizer PORV Spuriously Opens A spurious opening of pressurizer PORV could result in an uncontrolled depressurization of the RCS and loss of RCS inventory. However, an uncontrolled depressurization of the RCS will not occur since a second fire induced spurious opening of the pressurizer block valve would be required for such a depressurization. Therefore, the spurious opening of a pressurizer PORV will not result in an uncontrolled depressurization of the RCS and loss of RCS inventory. A loss of off-site power does not impact the results of the evaluation for this scenario.

3.4.3 One RCP Spuriously Starts The spurious start of an RCP could result in an uncontrolled depressurization of the RCS with off-site power available due to flow through a pressurizer spray valve. However, an uncontrolled depressurization will not occur because a second fire induced spurious opening of the pressurizer spray valve would be required for such a depressurization. Therefore, the spurious starting of an RCP will not result in the uncontrolled depressurization of the RCS with off-site power available. With off-site power not available, the spurious starting of an RCP cannot occur since power will not be available to the RCP.

3.4.4 One Feedwater Isolation Valve Spuriously Opens A spurious opening of a main feedwater isolation valve could result in the indicated pressurizer water level off-scale low due to over cooling of the RCS and overfill a steam generator due to excessive main feedwater flow. To prevent excessive main feedwater flow in the event of a spurious opening of a main feedwater isolation valve, the main steam isolation valves are closed, which secures steam to the turbine driven main feedwater pumps resulting in a loss of main feedwater flow. In addition, the startup feedwater pump is placed in PULL-TO-LOCK, ensuring flow from this feedwater source is secured. To initiate feedwater flow, a second fire induced spurious start of the startup feedwater pump would be required to initiate main feedwater flow.

Therefore, with all sources of main feedwater secured, the spurious opening of a main feedwater isolation valve will not result in the indicated pressurizer water level off-scale low or overfilling the steam generator. A loss of off-site power does not change the results of the evaluation for this scenario in that the startup feedwater pump cannot spuriously start because power to this pump is from a non-diesel backed power supply.

3.4.5 Startup Feedwater Pump Spuriously Starts The spurious start of the startup feedwater pump could result in the indicated pressurizer water level off-scale low due to over cooling of the RCS and overfilling the steam generators due to excessive main feedwater flow. The spurious start of the startup feedwater pump can only occur if off-site power is available since power to this pump is supplied by a non diesel backed power

Enclosure NOC-AE-11002643 Page 11 of 29 supply. Excessive main feedwater flow will not occur since a second fire induced spurious opening of a main feedwater isolation valve is required to permit flow to a steam generator.

Therefore, the spurious start of the startup feedwater pump will not result in the indicated pressurizer water level off-scale low and overfilling the steam generators.

3.4.6 Letdown Does Not Isolate The normal letdown flow is 120 gallons per minute. With an initial pressurizer water level at 25%, it would take 29.5 minutes for pressurizer water level to go off-scale low. Letdown isolation is achieved from outside the control room, through circuits isolated from the control room circuits, within 10 minutes of initiating the reactor trip. A loss of off-site power does not impact the results of the evaluation for this scenario. Therefore, sufficient time exists for the operators to take control of letdown from outside the control room to prevent the pressurizer level from going off-scale low.

3.4.7 Charging Pump Spuriously Starts If one centrifugal charging pump spuriously starts and automatic functions do not occur to provide an alternate source of makeup water, the inventory in the Volume Control Tank (VCT) would quickly be depleted, leading to air binding in the centrifugal charging pump. However, the second centrifugal charging pump remains available. For the case where one centrifugal charging pump spuriously starts and the automatic function to provide an alternate supply of water to the pump occurs, the pressurizer will go solid in 16 minutes. The operator action to transfer control of the charging pumps and isolate the control room circuit has been demonstrated to occur within 10 minutes. Therefore, sufficient time exists for the operator to take control of charging from outside the control room to prevent the pressurizer from going water solid. The above scenarios apply with or without off-site power because the centrifugal charging pumps are diesel backed.

3.4.8 Spurious Operations After Transfer to the Alternate Control Stations Transfer switches isolate the control room circuit from the Auxiliary Shutdown Panel (ASP) or Local Panel circuits to ensure that the effects of a fire in the control room will not impact the ability to safely shutdown the plant. Therefore, single or multiple spurious actuations that occur in the fire-affected area after transfer of safe shutdown control functions need not be considered.

See Section 3.6 for a more detailed description of the electrical isolation capability.

3.4.9 Single Spurious Actuation Analysis Conclusion The control room actions deemed necessary before evacuation are not negated by subsequent spurious actuation signals resulting from the postulated fire in a manner that would prevent the RCS process variables from staying within those values predicted for a loss of normal a-c power.

Because circuits in the control room are effectively isolated after control of the plant is achieved by the alternative shutdown system, single or multiple spurious actuations that could occur in the fire-affected area following transfer need not be considered.

Enclosure NOC-AE-1 1002643 Page 12 of 29 3.5 Automatic Turbine Trip Circuit Analysis This proposed change credits the automatic turbine trip upon a manual trip of the reactor. A logic diagram of the turbine trip circuitry is provided in Figure 2 to this Enclosure. When the reactor trip circuit breakers open in response to a manual reactor trip signal, an automatic turbine trip signal is generated by permissive P-16 relay located in each of two independent solid state protection system logic trains. It is unlikely that a fire-induced circuit failure would impact both independent channels. The P-16 relay generates a signal to the turbine electro-hydraulic control system cabinet (located outside the control room) to trip solenoids for repositioning valves in the turbine electro-hydraulic control system to dump oil pressure and allow the turbine throttle and governor valves to rapidly close under spring pressure thus securing steam flow to the main turbine. Once oil is unloaded, the fire-induced circuit failure can not fail in a condition where oil would be re-directed to re-open the turbine throttle and governor valves.

Based on the mechanics of this trip function, it is reasonable to assume that a turbine trip would be initiated as the result of a reactor trip and would not subsequently be negated by a fire-induced circuit failure.

3.6 Description of Electrical Isolation Capability If evacuation of the control room is required, the operators can establish and maintain the plant in a safe shutdown condition from outside the control room through the use of controls located at the ASP, transfer switch panels and other local control stations. These stations outside the control room provide the capability, in conjunction with limited local manual actions, for implementing cold shutdown from outside the control room.

The controls on the ASP are electrically isolated from those in the control room by transfer switches and fuses located on the transfer switch panels, with the exception of the controls associated with the turbine-driven AFW pump train and associated flow regulation. The transfer switches for the turbine-driven AFW pump and associated flow regulation controls are located on the ASP.

Six transfer switch panels are located in the Electrical Auxiliary Building with two of the panels located in each of their associated switchgear rooms on Elevation 10 ft, 35 ft, and 60 ft.

The switches and controls provided on the transfer switch panels are Class 1 E. Electrical and physical separation is maintained between the separation groups. The transfer switch panels provide control transfer between the control room and the ASP control circuits. In addition, control is provided on the transfer switch panels for equipment that requires one time or infrequent control during safe shutdown.

The transfer switches isolate the control room circuit from the ASP or Local Panel circuit to ensure that the effects of a fire in the control room will not impact the ability to safely shutdown the plant. The alternate shutdown stations have the capability of accepting a contact input from

Enclosure NOC-AE-1 1002643 Page 13 of 29 the transfer switch for the transfer of control from the control room to the ASP/Local Panels and vice-versa. Connection to the Operator Interface Modules (OIMs) are provided with adequate isolation or buffering such that a short, hot short, open circuit or ground through the non-active OIM or its cabling will not affect control by the active OIM or the functioning of the remainder of the system. The active OIM shall be selected through a separate transfer switch.

There are three types of control circuit configurations that are implemented at STP.

(1) The first control circuit configuration uses a transfer switch to perform the "transfer" and "isolation" functions. When the transfer switch is plaCed in the ASP/Local position, two sets of switch contacts operate in a break-before-make configuration. The first set of contacts open to electrically isolate the control room's control circuit. A second set of contacts close to transfer the component's control to the auxiliary location.

(2) The second control circuit configuration uses Qualified Data Processing System (QDPS) to perform the "transfer" and "isolation" functions. The transfer switch provides a "control position" input signal to QDPS which then transfers and isolates the component's control capability.

(3) The third control circuit configuration uses a transfer switch to perform the transfer function. The isolation function is performed by a combination of a transfer switch and fuses. The transfer switch functions as described in (1) above. The fuses are sized to electrically isolate the control room circuit before a loss of any upstream interrupting device that will cause a loss of a common power source.

3.7 Feasibility and Reliability of the Control Room Operator Actions This proposal credits the performance of operator actions within the control room until they can be backed up outside the control room with alternative circuits by transferring control to local control stations outside of the control room. Therefore, a feasibility and reliability assessment of the operator actions is provided.

NUREG-1852, "Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire," (Reference 6.9) provides technical guidance to assist in determining that operator manual actions are feasible and can be performed reliably in response to fire. The NUREG report provides criteria for analyzing the feasibility and reliability of operator manual actions to achieve safe shutdown. The following provides the STPNOC analysis of these criteria for justifying the operator manual actions specified in this proposal.

Enclosure NOC-AE-11002643 Page 14 of 29 3.7.1 Criterion 1: Analysis Showing Adequate Time Available to Perform the Actions to Address Feasibility Synopsis of criterion from NUREG-1852 Adequate time must be available to allow the actions to be diagnosed and executed in order to achieve and maintain hot shutdown following a single fire. The plant's thermal-hydraulic response must be analyzed to validate that the actions ensure that the safety functions can be performed STPNOC Evaluation:

The initiation of a fire in the control room will be rapidly diagnosed by the control room operators. Progression of the fire and the need for evacuation should be readily apparent so that time will be available to prepare for performing the actions. Once a decision to evacuate the control room is made, the control room operator uses a procedure to direct the actions to be taken. The operator is staged to rapidly perform the actions because their initiation is not required until direction is made to manually trip the reactor. Operator walk down performance data indicates that these actions can be performed in less than 90 seconds (most recent performance demonstrated that the steps were performed in 58 seconds) following initiation of the reactor trip which meets the thermal-hydraulic analysis (i.e. less than 120 seconds).

In conclusion, adequate time is available to perform the operator manual actions thus demonstrating that the actions are feasible.

3.7.2 Criterion 2: Analysis Showing Adequate Time Available to Ensure Reliability Synopsis of criterion from NUREG-1852 For feasible actions to be performed reliably, it should be shown that there is adequate time available to account for uncertainties in estimates of the time available and in estimates of how long it takes to diagnose and execute the operator manual actions. Sources of uncertainty that were analyzed are discussed below.

STPNOC Evaluation:

The walk through demonstration indicates that there is a relatively low margin. The following discussion provides assurance that this margin can be met.

The initiation and progression of a fire in the control room will be rapidly diagnosed by the control room operators. Although the operator actions must be performed rapidly, the operator should have time to be ready to proceed when the decision is made to manually trip the reactor. The actions can be performed by a single operator in the control room at adjacent

Enclosure NOC-AE-l 1002643 Page 15 of 29 panels (see Figure 1). The actions involve performing operations performed in response to other emergency operations that the operators are routinely trained to perform.

In conclusion, adequate time is available to ensure that the actions can be performed reliably.

3.7.3 Criterion 3: Environmental Factors Synopsis of criterion from NUREG-1852 Environmental conditions may affect an individual's mental or physical performance such that they may be degraded The expected environmental conditions considered both the locations where the operator manual actions are performed and the access route to the area.

STPNOC Evaluation:

A fire in the control room that is progressing to a condition where control room evacuation is required will result in a stressful environment for the control room operators. However, the operators are in an area that they routinely perform plant operations and they are trained to perform in emergency conditions. The required actions for the control room evacuation are many of the same actions that operators routinely train on in the performance for other emergencies (i.e, the actions are not unique to control room evacuation).

The operators are already in the area and manning the panels where the actions are required to be performed. No special protective clothing is required to perform the actions. Sufficient emergency lighting exists in the control room. The actions can be performed using normal face-to-face communications. The actions can be performed by a single operator in the control room from adjacent panels In conclusion, the environmental conditions in the control room will not impede the performance of the required operator actions and support the feasibility and reliability evaluation.

3.7.4 Criterion 4: Equipment Functionality and Accessibility Synopsis of criterion from NUREG-1852 The equipment necessary to achieve and maintain post-fire hot shutdown is accessible, and not damaged or otherwise adversely affected by the fire and its effects.

STPNOC Evaluation:

The proposal assumes that a fire-induced circuit failure will result in a single spurious actuation prior to transfer of control to alternate station outside the control room. The controls to perform the actions are readily accessible at adjacent control board panels in the control room.

Enclosure NOC-AE-1 1002643 Page 16 of 29 The equipment necessary to achieve and maintain post-fire hot shutdown are outside the control room. The alternate circuits to operate this equipment are isolated from the control room circuits and will not be damaged or otherwise adversely affected by a fire in the control room.

In conclusion, the equipment necessary to achieve and maintain post-fire hot shutdown remains accessible.

3.7.5 Criterion 5: Available Indications Synopsis of criterion from NUREG-1852 The system or component needs to include diagnostic indications relevant to the desired operator manual actions. These indications include those necessary to detect and diagnose the location of the fire.

STPNOC Evaluation:

A fire in the control room will be rapidly detected by the control room operators. Operator actions are performed rapidly in the control room prior to evacuation without further diagnosis. The actions are backed up from outside the control room within a short period of time. Sufficient indication is available outside the control room to demonstrate that the actions were successful in ensuring that the RCS process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L. 1.

In conclusion, diagnostic instrumentation remains available to support the feasibility and reliability evaluation.

3.7.6 Criterion 6: Communications Synopsis of criterion from NUREG-1852 Equipment to support communications among personnel is required to ensure proper performance of operator manual actions. Communications equipment is needed to ensure that any activities requiring coordination are clearly understood and correctly accomplished.

STPNOC Evaluation:

The operator manual actions will be performed using face-to-face communications in the control room. The equipment used to establish communications with the ASP station are a combination of sound-powered telephone circuits and portable radios. None of the communications equipment outside the control room will be adversely affected by a fire within the control room.

Enclosure NOC-AE-11002643 Page 17 of 29 Therefore, communication capability remains available to support the conclusions of the feasibility and reliability evaluation.

3.7.7 Criterion 7: Portable Equipment Synopsis of criterion from NUREG-1852 Ifportable equipment is needed to perform operator manual actions, the equipment should be functional and accessible to the extent it is needed to successfully implement the operator manual action. The portable equipment should be controlled and it should be routinely verified that the portable equipment is indeed located where it is supposed to be and has not been misplaced or otherwise moved Personnel should be trained to use the special tools and equipment in the planned application. If the use of the portable equipment may slow down action implementation, the delay should be considered in the time estimated (and preferably included in the demonstration) to perform the desired actions.

STPNOC Evaluation:

Portable radios are used by Plant Operators to establish communications with the ASP station in order to transfer equipment needed to achieve safe shutdown and isolate the control room circuits. The radios are part of the normal equipment carried by the Plant Operators such that their functionality is routinely verified during the shift. Plant Operators are trained in the routine use of the portable radio equipment. This is the only portable equipment needed to perform the operator manual actions.

Therefore, portable equipment remains functional and accessible to support the conclusions of the feasibility and reliability evaluation.

3.7.8 Criterion 8: Personnel Protection Equipment Synopsis of criterion from NUREG-1852 Personnel Protection Equipment needs to be functional and accessible to the extent it is needed to successfully implement the operator manual action. Personnel must be trained on its use and the time to put on the personnel protection equipment should be considered in the time line to perform the operator manual action.

STPNOC Evaluation:

Personnel protection equipment is not required to perform the operator manual actions. The use of self-containing breathing apparatus is not expected to be required prior to evacuation of the control room. Therefore, the conclusions of the feasibility and reliability evaluation are supported.

Enclosure NOC-AE-11002643 Page 18 of 29 3.7.9 Criterion 9: Procedures and Training Written procedures should cover the operator manual actions that are required to be performed to achieve and maintain hot shutdown. The operator should receive training on these manual actions.

STPNOC Evaluation:

Fires in control room leading to evacuation are addressed by procedure. The actions are in the current plant procedure and familiar to the operators. The plant operations staff is trained on the use of this plant procedure through the licensed operator requalification program. The operator manual actions are straightforward and familiar to the operators. The step to place the startup feedwater pump to PULL-TO-LOCK was recently added to the procedure. Plant operators have been notified of this change. Formal training on the procedure to include the new step is planned for the next licensed operator requalification training cycle that begins in May 2011. Once the fire condition is diagnosed and control room evacuation is needed, the actions are performed in sequence without further diagnosis.

In conclusion, written procedures and training support the feasibility and reliability evaluation.

3.7.10 Criterion 10: Stafming Adequate numbers of qualified personnel should be on site at all times so that hot shutdown conditions can be achieved and maintained in the event of afire. Individuals needed to perform the operator manual actions should not have collateral duties, such as fire fighting or control room operation, during the evolution of the fire.

STPNOC Evaluation:

The proposed actions are performed by a single operator assigned to the control room. The operators assigned to the alternate control stations are on-shift personnel. The operator has no other responsibilities during the performance of these actions. Therefore, plant staffing remains adequate to support the feasibility and reliability analyses.

3.7.11 Criterion 11: Demonstrations A demonstration with at least one randomly selected but established crew should be performed to provide a degree of overall assurance that the operator manual actions can be performed within the analyzed time available.

Enclosure NOC-AE-1 1002643 Page 19 of 29 STPNOC Evaluation:

Training and practice on the control room evacuation procedure is done at a frequency consistent with that established in existing training programs on abnormal procedures in compliance with 10 CFR 50.120. A demonstration of the actions, including the new step of placing the startup feedwater pump in PULL-TO-LOCK, was completed by one randomly selected crew to validate that the actions can be performed within the required times consistent with the thermal-hydraulic analysis.

There is adequate time to ensure that the (RCS) process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L. 1.

3.7.12 Conclusion The proposed operator actions are feasible and can be reliably performed to ensure that the (RCS) process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L. 1.

3.8 Defense-in-Depth Analysis The concept of defense-in-depth, described in 10 CFR 50, Appendix R, is applied to fire protection in fire areas important to safety, with the following three objectives:

1.

Prevent fires from starting;

2.

Detect rapidly, control, and extinguish promptly those fires that do occur; and,

3.

Provide protection of structures, systems and components (SSC) important to safety so that a fire that is not promptly extinguished by fire suppression activities will not prevent the safe shutdown of the plant.

3.8.1 Fire Prevention The objective of fire prevention is not affected by crediting the additional operations within the control room prior to evacuation. The area of the plant where the fire could occur that would require the manual actions to be performed has combustible loading allowances and limitations on hot work or other activities that are similar to other plant areas.

Enclosure NOC-AE-11002643 Page 20 of 29 3.8.2 Detect, Control and Extinguish Fires This objective is also not affected by crediting the additional operations within the control room prior to evacuation. The control room is Fire Zone Z034 in Fire Area 1. Fire Zone Z032 (the Relay Cabinet Area of the Control Room) and Fire Zone Z083 (the Watch Supervisor's Office) are also part of Fire Area 1. Fire Area 1 is located at elevation 35 foot in the electrical portion of the Mechanical/Electrical Auxiliary Building.

Ionization smoke detectors are provided in each fire zone in Fire Area 1. The operating area of the control room is enclosed by a seismically designed suspended ceiling and architectural barriers, all of which are constructed of materials with a flame spread rating of 50 or less. Fire detection is provided throughout the control room, both above and below the suspended ceiling and in the safe-shutdown control cabinets to provide early warning of a fire for manual fire fighting. These detectors alarm at the local fire panel and inside the room itself. Detector spacing is such that the number of detectors employed is several times that required by NFPA 72E-1978.

Fire protection is effected through portable water and C02 extinguishers and hose streams from standpipes strategically located at the control room exits and adjacent to Fire Zone Z036.

Fire Zone Z036 is located immediately outside of the control room entrance door. The majority of cables can be effectively reached by these hose streams from the floor level. A seismic catwalk above the ceiling ensures access for manual suppression in the event of a fire in the cabling in the area above the ceiling. The Heating, Ventilation and Air Conditioning (HVAC) system return ducts contain smoke detectors which, upon activation, close dampers and divert airflow into a purge and cleanup mode. The system also has manual override capability. The space above the suspended ceiling is not used as an HVAC plenum and contains limited combustibles.

In situ combustible loading is IEEE 383 cable and ordinary Class A combustibles.

There are only 38 cable trays (of which approximately 20% are covered) above the suspended ceiling. These trays are 40-percent filled. The cable trays have been grouped into 5 clusters and separated to provide ready access for manual fire fighting efforts. All cable trays that enter this area terminate in this area. The cabling in these cable trays is designed to meet the requirements of IEEE-383 and consists of control and instrumentation circuits. There are no other combustible materials above the suspended ceiling except power cables for lighting which are encased in steel conduit.

The Control Room is continuously manned. Automatic fire suppression has not been provided in the control room as the use of manual suppression by trained personnel provides a high reliability against accidental introduction of fire protection agents into this safety-related area.

Considering the high density early warning detection provided, the wide spacing of the trays, the type of cables and size of trays, the full accessibility of manual hose streams, and the continuous manning of the control room below, the use of automatic systems in this room is neither justified nor necessary.

Enclosure NOC-AE-1 1002643 Page 21 of 29 Adequate emergency lighting is provided in all fire areas needed to operate required equipment.

Automatic Halon Suppression is provided in the relay portion of the Control Room (i.e. Fire Zone Z032).

Most fires within the control room would be readily detected and extinguished by the control room operators. As a backup to the operators, a dedicated on-site Fire Brigade is maintained at STP consisting of a minimum of five members per shift, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day trained to respond to and combat fires. Fire Brigade fire drill training results for plant areas in proximity to this fire zone indicate that Fire Brigade response to a fire is expected to be timely, with entry into the affected area within 15 to 30 minutes of initial alarm.

3.8.3 Protect SSCs so that Fires Will Not Prevent Safe Shutdown Existing fire protection regulations rely on passive fire protection through fire barriers that, when operable, have a high level of reliability to prevent the damage to redundant trains required for safe shutdown. In the event of a control room or relay room fire, cold shutdown can be achieved and maintained from the auxiliary shutdown panel, transfer switch panels or other local control stations and motor control cabinets. Transfer switches located outside the control room isolate the control room circuits from the ASP or Local Panel circuits to ensure that the affects of a fire in the control room will not impact the ability to safely shutdown the plant.

Although the Fire Area 1 fire zone boundary penetrations are not provided with rated penetration seals or HVAC dampers, the fire zone boundaries are significant obstructions which will limit zone-to-zone fire propagations. This facilitates containing and suppressing fires within a fire zone using manual fire fighting capabilities.

The walls, floors, and ceilings of Fire Area 1 are all 3-hour rated fire barriers with the exception of those portions of Fire Zones Z032 and Z034 that are adjacent to Fire Area 19. Fire Area 19 is a stairwell that has a 2-hour fire rating with Class B labeled doors and 1-1/2 hour rated fire dampers. The structural steel supporting the fire barrier is fireproofed to the same rating as that required of the barrier.

Doors and penetrations contained in these fire barriers have fire ratings compatible with that of the barrier. Ventilation duct penetrations in fire barriers are provided with 3-hour rated fire dampers installed in accordance with the manufacturer's instructions. In Fire Zone Z005, elevation 10' ventilation duct in the south wall is not provided with a three-hour fire damper.

To compensate for the lack of the fire damper, the duct that extends from the wall to the nearest fire damper is provided with three-hour fire rated coating. Smoke and heat removal are accomplished with portable exhaust fans and flexible ductwork.

Fire Area 1 has substantial boundaries between Fire Zones and encompassing the Fire Area to limit the spread of fire.

Enclosure NOC-AE-1 1002643 Page 22 of 29 3.8.4 Summary Considerable defense-in-depth features exist in Fire Area 1 such that it is extremely unlikely that a fire would result in evacuation of the control room. The control room is continuously manned such that adherence to fire prevention standards are continuously monitored. Fire Area 1 has substantial boundaries between Fire Zones and encompassing the Fire Area to limit the spread of fire. Numerous automatic detectors exist throughout the area. The automatic detectors combined with a continuously manned area allows for rapid detection and extinguishing of any fire and readily-available extinguishing agents makes it unlikely that a fire would result in evacuation of the control room.

3.9 Technical Evaluation Conclusion

Considerable defense-in-depth features exist in Fire Area 1 such that it is extremely unlikely that a fire would result in evacuation of the control room. Thermal-hydraulic analysis demonstrates that the proposed operations to be performed in the control room will ensure that the RCS process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L. 1. The analysis demonstrates that a single spurious operation before control of the plant is achieved through the alternative or dedicated shutdown system will not adversely impact the results of the analysis. Because circuits in the control room are effectively isolated after control of the plant is achieved by the alternative shutdown system, single or multiple spurious actuations that could occur in the fire-affected area need not be considered following transfer. The proposed operations are feasible and can be reliably performed and cannot be negated by subsequent spurious actuation signals resulting from the postulated fire that would result in RCS process variables beyond those values predicted for a loss of normal a-c power.

4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria Section III.L. 1 of Appendix R to 10 CFR 50 states, in part, alternative or dedicated shutdown capability provided for a specific fire area shall be able to (a) achieve and maintain sub-critical reactivity conditions in the reactor; (b) maintain reactor coolant inventory; (c) achieve and maintain hot standby conditions: (d) achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and (e) maintain cold shutdown conditions thereafter. During the post fire shutdown, the RCS process variables shall be maintained within those predicted for a loss of normal a-c power.

Section III.L.2 of Appendix R to 10 CFR 50 states, in part that the performance goals for the shutdown functions shall be:

a. The reactivity control function shall be capable of achieving and maintaining cold shutdown reactivity conditions.

Enclosure NOC-AE-11002643 Page 23 of 29

b. The reactor coolant makeup function shall be capable of maintaining the reactor coolant level above the top of the core for BWRs and be within the level indication in the pressurizer for PWRs.

c. The reactor heat removal function shall be capable of achieving and maintaining decay heat removal.

d. The process monitoring function shall be capable of providing direct readings of the process variables necessary to perform and control the above functions.

e. The supporting functions shall be capable of providing the process cooling, lubrication, etc., necessary to permit the operation of the equipment used for safe shutdown functions.

Section IiI.G.3 of Appendix R to 10 CFR 50 states, in part, alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room or zone under consideration, shall be provided where the protection of systems whose function is required for hot shutdown does not satisfy the circuit separation protection requirements of Section III.G.2.

Section 5.4.4 of NRC Regulatory Guide (RG) 1.189, Revision 2 states:

The only operator action in the control room before evacuation for which credit is usually given is reactor trip. For any additional control room actions deemed necessary before evacuation, a licensee should be able to demonstrate that such actions can be performed.

Additionally, the licensee should ensure that such actions cannot be negated by subsequent spurious actuation signals resulting from the postulated fire. The design basis for the control room fire should consider one spurious actuation or signal to occur before control of the plant is achieved through the alternative or dedicated shutdown system. After control of the plant is achieved by the alternative or dedicated shutdown system, single or multiple spurious actuations that could occur in the fire-affected area should be considered, in accordance with the plant's approved FPP.

NUREG-1852, "Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire," provides published technical guidance to assist in determining that operator manual actions are feasible and can be performed reliably in response to fire. The NUREG report provides criteria for analyzing the feasibility and reliability of operator manual actions to achieve safe shutdown.

The STPNOC License Condition 2.E specifies, STPNOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment No.

55 and the Fire Hazards Analysis Report through Amendment No. 19, and submittals dated April 29, May 7, 8 and 29, June 11, 25 and 26, 1987; February 3, March 3, and November

Enclosure NOC-AE-1 1002643 Page 24 of 29 20, 2009; January 20, 2010; and as approved in the SER (NUREG-0781 ) dated April 1986 and its Supplements, subject to the following provision:

STPNOC may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

STP was licensed after January 1, 1979 and is not required to meet Appendix R. The approved STP FPP was reviewed by the NRC and is documented in the STP FHAR. The STP Fire Hazards Analysis Report (FHAR) provides an analysis of how the safe shutdown strategy for each fire area meets regulatory requirements.

4.2 Precedent The NRC approved the use of operator actions for tripping both units, closure of the main steam isolation valves, closure of the feedwater discharge valves and tripping of the feedwater turbine prior to evacuating the control room for the Susquehanna Plant. (Reference 6.10) The NRC concluded that since all actions, including the manual trip of the reactor, could be accomplished in rapid succession by a single operator at one location, this approach provided a suitable means of precluding potential spurious operations that could affect the shutdown capability, while satisfying the concern for limiting the number of actions within the control room prior to evacuation.

The NRC approved the use of operator manual actions, where the circuit separation criteria of Appendix R,Section III.G.2 are not met in the STP Fire Hazards Analysis, in STP Amendments 186/173 (Reference 6.11) and Amendments 193/181 (Reference 6.12).

4.3 Significant Hazards Consideration STPNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response

No.

The design function of structures, systems and component are not impacted by the proposed change. The proposed change involves crediting operations in the control room prior to evacuation in the event of a fire in order to meet safe shutdown performance criteria. The proposed action will not initiate an event. The proposed actions do not increase the probability of occurrence of a fire or any other accident previously evaluated.

Enclosure NOC-AE-11002643 Page 25 of 29 The proposed operations are feasible and reliable and demonstrate that the unit can be safely shutdown in the event of a fire. No significant consequences result from the performance of the proposed operations.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response

No.

The design function of structures, systems and component are not impacted by the proposed amendment. The proposed change involves operations in response to a fire. They do not involve new failure mechanisms or malfunctions that can initiate a new accident.

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

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

Response

No.

Thermal-hydraulic analysis demonstrates that the proposed operations to be performed in the control room will ensure that the RCS process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L. 1. The analysis demonstrates that a single spurious operation before control of the plant is achieved through the alternative or dedicated shutdown system will not adversely impact the results of the analysis. After control of the plant is achieved by the alternative or dedicated shutdown system, circuits subjected to fire-induced circuit failures are isolated from the control stations such that the safe shutdown operations will not be compromised.

The need to perform the proposed operations can be readily diagnosed and the operations can be performed in rapid succession by control room operators at their normal control station.

The actions are straightforward and familiar to the operators. The actions have been verified that they can be performed through demonstration. The operations are backed up outside the control room such that assurance exists they should not be negated by subsequent spurious actuation signals from a postulated fire.

The automatic turbine trip action can reasonably be assumed to occur with the credited manual reactor trip action that is part of the current licensing basis.

Considerable defense-in-depth features exist in Fire Area 1 such that it is extremely unlikely that a fire would result in evacuation of the control room.

Enclosure NOC-AE-11002643 Page 26 of 29 The proposed operations are feasible and reliable and demonstrate that the unit can be safely shutdown in the event of a fire. The operations ensure that performance goals of Appendix R,Section III.L.2 are met. The achievement of these goals provide adequate margin from challenging any safety limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, STPNOC concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

5.0 Environmental Consideration STPNOC has reviewed the proposed amendment and determined that it does not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in the individual or cumulative occupational exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 1 OCFR51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 References 6.1 Letter from David W. Rencurrel, STPNOC, to NRC Document Control Desk dated February 4, 2008, "License Amendment Request for Deviation from Fire Protection Program Requirements." (TAC Nos. MD8112 and MD8113, ML080390483, NOC-AE-07002212) 6.2 EPRI Report No. 1006961, "Spurious Actuation of Electrical Circuits Due to Cable Fires: Results of an Expert Elicitation," dated May 2002 and NUREG/CR-6776, "Cable Insulation Resistance Measurements Made During Cable Fire Tests," dated June 2002.

6.3 Letter from Scott M. Head, STPNOC, to NRC Document Control Desk dated June 16, 2008, "Withdrawal of Proposed License Amendment Request for

Enclosure NOC-AE-11002643 Page 27 of 29 Deviation from Fire Protection Program Requirements." (TAC Nos. MD6694 and MD6695) (ML081820441) (NOC-AE-08002322) 6.4 NUREG-078 1, Supplement No. 2, Safety Evaluation Report related to the operation of South Texas Project, Units 1 and 2, dated January 1987.

6.5 South Texas Project Electric Generating Station - NRC Integrated Inspection Report 05000498/2006002 and 05000499/2006002, dated May 18, 2006.

(ML061390160) (ST-AE-NOC-06001496) 6.6 NRC Regulatory Guide 1.189, Revision 2, "Fire Protection for Nuclear Power Plants," dated October 2009.

6.7 South Texas Project Nuclear Calculation NC-7079, Revision 3, "Fire Hazards Analysis," dated May 26, 2010.

6.8 WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis, April, 1999 6.9 NUREG-1852, "Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire," USNRC, Final Report, October 2007 6.10 Letter from Chester Poslusny, NRC, to Robert G. Byram, Pennsylvania Power &

Light Company, dated October 21, 1997, "Evaluation of Fire Protection Issues, Safe Shutdown Methodology and Analysis of Associated Circuits, Susquehanna Steam Electric Station (SSES), Units 1 and 2." (TAC Nos. M90600 and M90601) 6.11 South Texas Project, Units 1 and 2 - Issuance of Amendments RE: Deviation from Fire Protection Program Requirements, dated September 16, 2008. (TAC Nos. MD6694 and MD6695) (ML082280465 and ML082280472) (ST-AE-NOC-08001805) 6.12 South Texas Project, Units 1 and 2 - Issuance of Amendments Approving Deviations from Fire Protection Program Requirements for Fire Areas 27 and 31, and Revision to License Condition 2.E, dated March 31, 2010. (TAC Nos.

ME0824, ME0825, ME1389 and ME1390) (ML100780075) (ST-AE-NOC-10001967)

Figure 1 Control Room Control Panel Switch Orientation V

I4. RCP TRIP SW F.

FEEDWATER I!

6. STARTUP FEEC

7. LETDOWN ISOI

• P-007 zcP-ea ZCP4B9 ZCP-eie CONTROL ROOM

=SW ITCH ISOLATION SWITCH PORV BLOCK VALVE SWITCHES (TWO)

I TCHES (FOUR)

SOLATION SWITCH (FOUR)

)WATER PUMP PULL-TO-LOCK SWITCH LATION SWITCHES (TWO) l PULL-TO-LOCK SWITCHES (TWO) 8.&CHARGING PU

NOTE, ALL SW I TCHES ARE ON BENCIO0ARD PART OF PANEL EXCEPT SWITCHES 3 b 7 WHICH ARE ON VERTICAL SECT I ON OF PANEL.

I SOUTH TEXAS PROJECT UNITS 181 2 CONTRO. ROOM CONTR. PELe.

SWITCH ORIENTATION FIGURE 1 REVISION I 00 0

z0

Figure 2 Enclosure NOC-AE-1 1002643 Page 29 of 29 Turbine Trip Circuitry Logic Diagram SSPS TRAIN R SSPS TRAIN S D.WP EMENGFNCY TRIP rtUID TURAJNN THROTII F

STOP VALME

Enclosure, Attachment 1 NOC-AE-11002643 Page 1 of 2 Enclosure, Attachment 1 Proposed Change to South Texas Project, Unit 1, Operating License No. NPF-76 (One Page)

SOUTH TEXAS LICENSE (4)

The facility has been granted a schedular exemption from Section 50.71 (e)(3)(i) of 10 CFR 50 to extend the date for submittal of the updated Final Safety Analysis Report to no later than one year after the date of issuance of a low power license for the South Texas Project, Unit 2. This exemption is effective until August 1990. The staffs environmental assessment was published on December 16, 1987 (52 FR 47805).

E.

Fire Protection STPNOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment No. 55 and the Fire Hazards Analysis Report through Amendment No. gxx, and submittals dated April 29, May 7, 8 and 29, June 11, 25 and 26, 1987; February 3, March 3, and November 20, 2009; January 20, 2010; and as approved in the SER (NUREG-0781) dated April 1986 and its Supplements, subject to the following provision:

STPNOC may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

F.

Physical Security STPNOC shall fully implement and maintain in effect all provisions of the physical security, training and qualification, and safeguards contingency plans previously approved by the Commission and all amendments and revisions to such plans made pursuant to the authority under 10 CFR 50.90 and 10 CFR 50.54(p).

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "South Texas Project Electric Generating Station Security, Training and Qualification, and Safeguards Contingency Plan, Revision 2" submitted by letters dated May 17 and 18, 2006.

G.

Not Used H.

Financial Protection The Owners shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

Amendment No.

Enclosure, Attachment 2 NOC-AE-1 1002643 Page 1 of 2 Enclosure, Attachment 2 Proposed Change to South Texas Project, Unit 2, Operating License No. NPF-80 (One Page)

(2)

The facility was previously granted exemption from the criticality monitoring requirements of 10 CFR 70.24 (See Materials License No.

SNM-1983 dated August 30, 1988 and Section III.E. of the SER dated August 30, 1988). The South Texas Project Unit 2 is hereby exempted from the criticality monitoring provisions of 10 CFR 70.24 as applied to fuel assemblies held under this license.

(3)

The facility requires a temporary exemption from the scheduler requirements of the decommissioning planning rule, 10 CFR 50.33(k) and 10 CFR 50.75. The justification for this exemption is contained in Section 22.2 of Supplement 6 to the Safety Evaluation Report. The staffs environmental assessment was published on December 16, 1988 (53 FR 50604). Therefore, pursuant to 10 CFR 50.12(a)(1), 50.12(a)(2)(ii) and 50.12(a)(2)(v), the South Texas Project, Unit 2 is hereby granted a temporary exemption from the schedular requirements of 10CFR 50.33(k) and 10 CFR 50.75 and is required to submit the decommissioning plan for both South Texas Project, Units 1 and 2 on or before July 26, 1990.

E.

Fire Protection STPNOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment No. 62 and the Fire Hazards Analysis Report through Amendment No. xx, and submittals dated April 29, May 7, 8 and 29, June 11, 25, and 26, 1987; February 3, March 3, and November 20, 2009; January 20, 2010; and as approved in the SER (NUREG-0781) dated April 1986 and its Supplements, subject to the following provisions:

STPNOC may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

F.

Physical Security The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "South Texas Project Electric Generating Station Security, Training and Qualification, and Safeguards Contingency Plan, Revision 2" submitted by letters dated May 17 and 18, 2006.

G.

Not Used Amendment No.

Enclosure, Attachment 3 NOC-AE-11002643 Page 1 of 6 Enclosure, Attachment 3 Annotated Fire Hazards Analysis Report Page Page 2-10 Page 3.2-6 Page 3.2-7 Page 3.2-7a Page 4-17

Enclosure, Attachment 3 NOC-AE-1 1002643 Page 2 of 6 STP FMAR 2.4.4 Alternate Shutdown Capability Alternate shutdown capability is provided to respond to a large fire occurring within the main control room. Following eso operationsperfored i..........

om the control room

'iapid succession', the transfer of control from the control room to the auxiliary shutdown panel and local control stations is accomplished from outside the control room using transfer switches which are predominately located in the three redundant switchgear rooms. The remaining transfer switches are on the auxiliary shutdown panel in the train related diesel generator rooms and at the Essential Cooling Water Intake Structure ventilation fan MCCs.

When transferred, these circuits are independent of the control room. Safe shutdown from outside the control room is discussed in FSAR Section 7.4.1.9.

If a loss of offsite power occurs, all three Class 1 E standby diesel generators receive an automatic start signal. No single control circuit failure due to a control room fire can disable all standby diesel generators. The sequencer circuits for the standby diesel generators are on separate paths outside the control room. The sequencers are located within their own fire area separated from the control room fire area. When control of a standby diesel generator is transferred to the local control station, the diesel will remain operating. Only one standby diesel generator is required to achieve safe shutdown.

! Credit is given to the following operations in the control room to meet the requirements of Appendix R, Section lI.L.1. Operators taking action to:

1. Trip the reactor

2. Isolate main steam

3. Close the pressurizer power-operated reief valves (PORV) block valves

4. Secure all reactor coolant pumps

5. Isolate feedwater

6. Prevent the startup feedwater pump from starting

7. Isolate letdown

8. Secure the charging pumps In addition, credit is taken for an automatic turbine trip in response to the reactor trip.

2-10 Amendment xx

Enclosure, Attachment 3 NOC-AE-11002643 Page 3 of 6 STP FHAR (Fire Zone Z032). Alternate shutdown capability is provided by the auxiliary shutdown stations, including the auxiliary shutdown panel, which are located in separate fire areas from the Control Room. Transfer switches are also located in separate fire areas. This precluded the need to provide separation for Trains A, B, C and D in the Control Room and Relay Cabinet Area.

D.

Redundant Safe Shutdown Assessment In the event of a Control Room or Relay Room fire, cold shutdown can be achieved and maintained from the auxiliary shutdown panel, transfer switch panels or other local control stations and MCCs.

However, before control room evacuation, operator action should be taken from the control room to trip the reactor, secure the reactor coolant pumps and CVCS charging pumps, and close the pressurizer PORV block valves, the letdown isolation valves, MSIV, and MSIV bypass FW1V valves, prevent the startup feedwater pump from starting. With the exception of tripping the reactor and preventing the startup feedwater pump from starting, the above actions may be performed from an alternate shutdown location if they cannot be completed prior to evacuating the control room.

E.

Conclusions The loss of all circuits and equipment in Fire Area 1 is acceptable as safe shutdown functions can be controlled from the auxiliary shutdown stations which are in separate fire areas.

Circuits necessary to shutdown from the auxiliary shutdown locations would remain free of fire damage.

F.

Deviations from BTP APCSB 9.5-1 Appendix A and/or 10CFR50 Appendix R with Respective Justifications

1.

Appendix A:F.2 Deviation Cables located above a suspended ceiling.

Justification See FHAR 4.2, Comparison to Appendix A of APCSB 9.5-1, Section F.2

2.

Appendix A:F.2 and Appendix R:III.G.3 Deviation No fixed suppression in the control room complex.

Justification 3.2-6 Amendment 19

Enclosure, Attachment 3 NOC-AE-1 1002643 Page 4 of 6 STP FHAR See FMAR 4.2, Comparison to Appendix A of APCSB 9.5-1, Section F.2, FMAR 4.1, Comparison to 10CFR50 Appendix R, Section III.G, and FHAR pages 3.2-5 and 3.10-1.

3.

Appendix A: F.2 Deviation Control room is not separated from the relay room and watch supervisor's office by 3-hour barriers.

Justification See FHAR 4.2, Comparison to Appendix A of APCSB 9.5-1, Section F.2 and Fire Area 1, Section 3.2.

4.

Appendix A:D. 1.J Deviation Unlabeled doors in Fire Area boundaries.

Justification See FHAR 4.2, Comparison to Appendix A of APCSB 9.5-1, Section D. 1.j.

5.

Appendix A:D.1.d Deviation The control room Z034 contains a frame utilized to display current control room staffing which is constructed of untreated wood.

Justification The frame does not significantly add to the combustible loading of the zone.

3.2-7 Amendment 19

Enclosure, Attachment 3 NOC-AE-11002643 Page 5 of 6 STP FHAR

6.

Appendix A: M.G.3 In addition to reactor trip, credits a series of operations performed in the control room in rapid succession prior to transfer to the control to the auxiliary shutdown panel and other points of control for meeting the alternate shutdown capability.

Justification License Amendment No. xx for Unit 1 and License Amendment No. xx for Unit 2.

3.2-7a Amendment 19

4.1 COMPARISON OF STP UNITS WITH REQUIREMENTS OF APPENDIX R APPENDIX R REOUIREMENTS STP POSITION III.G. (Cont'd)

III.G. (Cont'd)

3. Alternative or dedicated shutdown capability and its associated circuits,2 independent of cables, systems or components in the area, room or zone under consideration, shall be provided:

a.

Where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section; or

b. Where redundant trains of systems required for hot shutdown located in the same fire area may be subject to damage from fire suppression activities or from the rupture or inadvertent operation of fire suppression systems.

In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration.

3. Alternate shutdown capability for all three trains is provided outside the control room fire area to respond to a large control roomfire._Followinga series of_

operations performed in the control room in rapid succession r..etr t* p f* m the eeatrc 9.T I. the transfer of control to the auxiliary shutdown panel and other points of control is accomplished from outside the control room fire area. These circuits when transferred are independent of the control room fire area.

A fixed fire suppression system has not been provided throughout the control room. A detailed justification for this deviation is presented in Section 3.10. The control room is continuously occupied and is provided with portable fire extinguishers inside the control room and fire hose stations near the entrances. Fire detection is provided in the control room and the relay portion of the control room is provided with an automatic Halon suppression system.

2 Alternate shutdown capability is provided by rerouting, relocating or modificating of existing systems; dedicated shutdown capability is provided by installing new structures and systems for the function of post-fire shutdown.

FQ 0~~

0

Enclosure, Attachment 4 NOC-AE-l 1002643 Page 1 of 1 List of Commitments The following table identifies those actions committed to by STPNOC in this document. Any statements in this document with the exception of those in the table below are provided for information purposes and are not considered commitments. Please direct questions regarding these commitments to Ken Taplett at (361) 972-8416.

Commitment Continuing Scheduled Compliance Completion Date The annotated FHAR pages provided in Attachment X

Prior to 3 to the License Amendment Request will be Implementation of documented in the STP FHAR, upon approval of the Amendment amendment request by the NRC.