ST-HL-AE-5414, Provides Addl Info Re Proposed Single ESF Train Mitigation Capabilities

Provides Addl Info Re Proposed Single ESF Train Mitigation Capabilities

ML20115E005
Person / Time
Site:	South Texas
Issue date:	07/09/1996
From:	Leazar D HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ST-HL-AE-5414, TAC-M92169, TAC-M92170, NUDOCS 9607160053
Download: ML20115E005 (7)

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The Light company South Texas Project Hectric Generating Station P.o. Box 289 5 Worth, Texas 77483 i Houston Lighting & Power July 5', 1996 ST-HL-AE-5414 I File No.: G20.02.01 I 10CFR50.90,50.92 U. S. Nuclear Regulatory Commission  !

Attention: Document Control Desk i Washington, D.C. 20555 South Texas Project Units I and 2 Docket Nos. STN 50-498; STN 50-499 Additional Information Regarding Proposed Single Engineered Safety Features 1 Train Mitigation Capabilities (TAC No. M92169/M92170)

Subsequent to the transmittal ofinfonnation to support the previously submitted Special Test j Exception of the Standby Diesel Generator, the Nuclear Regulatory Commission staff requested  !

additional information relating to single train event mitigation. The requested information and j response is provided in Attachment I to this letter. .

- 1 Should you have any questions, please contact Mr. A. W. Harrison at (512) 972-7298 or me at (512)972-7795.

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D. A. Leazar Director, 1C002S Nuclear Fuel and Analysis 1

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/1 OD l ' l Attachment 1 9607160053 960709 PDR ADOCK 05000498 P PDR

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Houston Lighting & Power Company ST-HL-AE-5414 South Texas Project Electric Generating Station File No.: G20.02.01 Page 2 l

Leonard J. Callan Rufus S. Scott l Regional Administrator, Region IV Associate General Counsel l U. S. Nuclear Regulatory Commission Houston Lighting & Power Company 611 Ryan Plaza Drive, Suite 400 P. O. Box 61067 l Arlington, TX 76011-8064 Houston, TX 77208 Thomas W. Alexion Institute of Nuclear Power Project Manager, Mail Code 13H3 Operations - Records Center i U. S. Nuclear Regulatory Commission 700 Galleria Parkway  ;

l Washington, DC 20555-0001 Atlanta, GA 30339-5957 David P. Loveless Dr. Joseph M. Hendrie  ;

Sr. Resident Inspector 50 Bellport Lane l c/o U. S. Nuclear Regulatory Comm. Bellport, NY 11713 ,

P. O. Box 910 Bay City, TX 77404-0910 Richard A. Ratliff  !

Bureau of Radiation Control J. R. Newman, Esquire Texas Department of Health  !

Morgan, Lewis & Bockius 1100 West 49th Street ,

1800 M Street, N.W. Austin, TX 78756-3189 Washington, DC 20036-5869 .

U. S. Nuclear Regulatory Comm.

M. T. Hardt/W. C. Gunst Attn: Document Control Desk City Public Service Washington, DC 20555-0001 P. O. Box 1771 '

San Antonio, TX 78296 J. C. Lanier/M. B. Lee J. R. Egan, Esquire City of Austin Egan & Associates, P.C.  !

Electric Utility Department 2300 N Street, N.W. l 721 Barton Springs Road Washington, D.C. 20037 Austin, TX 78704 l

Central Power and Light Company J. W. Beck ATTN: G. E. Vaughn/C. A. Johnson Little Harbor Consultants, Inc.

P. O. Box 289, Mail Code: N5012 44 Nichols Road Wadsworth, TX 77483 Cohassett, MA 02025-1166 J. L. Pellet U. S. Nuclear Regulatory Commission

l. 611 Ryan Plaza Drive, Suite 400

[ Arlington, TX 76011-8064 l

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Attachm:nt 1 ST-HL-AE-5414 i Page1of5 l A. Describe the operator actions and time frame to accomplish the following activities. Clarify l where the operator has to go to perform the action if outside the Control Room.

i l 1. Action to depressurize the Reactor Coolant System to a pressure for using the Low Head Safety Injection pump to mitigate the Small Break Loss of Coolant Accident:

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Response

l In the eveat cf degraded or inadequate core cooling, the Emergency Operating Procedures at the l South Texas Project provide direction for manual post-accident depressurization and cooldown. l These actions are based on core exit thermocouple temperatures and are found in procedures )

OPOP05-EO-FRC1, Response to Inadequate Core Cooling", and OPOP05-EO-FRC2, " Response to Degraded Core Cooling". If Auxiliary Feedwater flow and Steam Generator level are adequate, the j operator is instructed to lower Reactor Coolant System pressure via a secondary side cooldown using l the Steam Generators. If this is not possible, the operator is instructed to manually open a Pressurizer PORV to depressurize the Reactor Coolant System to a pressure that will allow injection via the Low Head Safety Injection pump. Either of these actions can be taken from the Control Room quickly

(<l5 minutes to perform actions) if necessary to mitigate a Small Break Loss-of-Coolant Accident.

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2. Action to isolate non essential flow to Component Cooling Water system:

Response:  !

l As stated on page 3 of 12 in letter ST-HL-AE-5208 dated November 22,1995, an analysis of the Component Cooling Water system following a Safety Injection signal indicates the remaining Component Cooling Water train will operate satisfactorily following a loss of power to any two Component Cooling Water trains. The Component Cooling Water has five flow paths that are isolated by a safety injection signal. Total Component Cooling Water flow for these five flow paths is approximately 12,400 gpm. Approximately 95% of the flow or 11,800 gpm is carried by two of the flow paths, the spent fuel pool coolers and the non-safety header. The isolation design for these two flow paths has redundant valves powered from all three Standby Diesel Generator trains. The configuration of these redundant valves ensures that if there is only one operable Standby Diesel Generator, only one of these flow paths will remain unisolated. This condition is only applicable for i Trains A or B, since Train C powers an isolation valve in both of these flow paths. In the case where a flow path is unisolated and it is desired to isolate the remaining flow path, an operator can be dispatched to manually close the remaining open isolation valve. This action would be performed by a Plant Operator in the Component Cooling Water header room of the Mechanical Auxiliary Building, which is located within 5 minutes of the Control Room. The open isolation valve could be manually closed within approximately 5 minutes by the operator.

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Attachm:nt i ST-HL-AE-5414 1 Page 2 of 5

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3. Action to load the Positive Displacement Pump on the Technical Support Center Diesel Generator and align to the Reactor Coolant Pump seals:

Resnnnse:

The Technical Support Center Diesel Generator will start automatically on the Loss of Offsite Power  ;

event and power its associated bus. If required, the Control Room Operator will then start the Positive Displacement Pump from the Control Room and supply Reactor Coolant Pump seal injecticn 8 via the normal path. These actions which are described in OPOP04-AE-0001, " Loss of any 13.8 kV ,

or 4.16 kV bus", could be complete in approximately 15 minutes. Ilowever, as discussed in the t response to question D, single train operation will still provide the necessary Component Cooling Water flow to the Reactor Coolant Pump thermal barrier to prevent seal damage.  !

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4. Action to align Auxiliary Feedwater to an intact Steam Generator:

Remnonne:

STP has four Auxiliary Feedwater (AFW) trains feeding four steam generators. Auxiliary Feedwater trains A , B, and C are powered from Safety Trains A, B and C respectively. Auxiliary Feedwater Train D is a Turbine Driven pump which feeds D steam generator. If only one Standby Diesel Generator is available, then the minimum of one motor driven Auxiliary Feedwater pump and the steam driven Auxiliary Feedwater will be available to feed the steam generators. If either of these Auxiliary Feedwater pumps feeds an intact steam generator, the acceptance limit for this accident, i

which is the pressurizer does not go water solid within 30 minutes with no operator action, will be met.

I Ifit becomes necessary for another steam generator to be supplied by either of the running Auxiliary Feedwater pumps, a Plant Operator can be dispatched from the control room to the Auxiliary Feedwater valve cubicle in the Isolation Valve Cubicle in approximately 5 minutes. The performance of the necessary valve manipulations will take approximately 10 minutes.

5. Action to isolate the containment isolation valves to the piping for the containment radiation monitors:

i Resnonse:

Emergency Operating Procedure OPOP05-EO-0000, " Reactor Trip or Safety Injection" will be revised to provide direction to manually isolate these lines if unable to do so from the control room.

A Plant Operator dispatched from the control room will be able to reach the Mechanical Auxiliary l Building 41' Piping Penetration in approximately 5 minutes and locally isolate the containment j radiation monitors in approximately 10 minutes. )

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Attachm:nt 1 ST-HL-AE-5414 l Page 3 of 5 )

l B. Discuss the mitigating actions for Train C of Fuel Handling Building HVAC:

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l Recnonce:

1 The Fuel Handling Building HVAC is designed to mitigate the consequences of a fuel handling l l accident as well as LOCA (leakage from ECCS components located in the Fuel Handling Building) l by limiting plant site boundary dose to within the guidelines of 10 CFR 100. This is accomplished by routing exhaust air from the spent fuel pool and the remainder of the Fuel Handling Building through ESF filter units containing HEPA filters and iodine removal carbon filter units if high levels of l l airbome radioactivity are detected in the exhaust air or Safety Injectica is actuated. i The Fuel Handling Building exhaust air filters consist of a two train arrangement powered from A and B ESF diesels. In the event that only C ESF diesel is available, action would be taken to cross-connect ESF 4.16 kV buses to provide power to either A or B Fuel Handling Building exhaust air filter trains.

The following describes the steps required to connect one South Texas Project Engineered Safety j Features train to another Engineered Safety Features train, via Emergency Bus IL, any two Auxiliary i Engineered Safety Features Transformers, and the associated motor operated switches. This cross-connect scheme would be utilized to provide Standby Diesel Generator-backed power to the Fuel Handling Building HVAC heaters (114 kW per train) should no other preferred offsite power source  ;

or SDG-backed source be available. This description summarizes required actions following the

postulated loss of AC power to Engineered Safety Features Buses A and B. The initial conditions for this scenario include

(1) SDG A and B are out of service; (2) SDG C is in the standby mode; and (3)

Class lE buses an supplied from offsite power. A loss of offsite power then occurs, followed by a I fuel handling accident or LOCA, the successful start of SDG C, and the successful loading of SDG C.

The :pecific steps required to cross-connect two Engineered Safety Features buses are detailed in Addendum 12 of STP procedure OPOP04-AE-0001, " Loss of Any 13.8 kV or 4.16 kV Bus". This description summarizes the steps for cross-connecting Engineered Safety Features Bus C to Engineered Safety Features Bus B. The cross-connect sequence includes opening the Engineered Safety Features Transformer ElB and ElC Motor Operated Switches (MOS) [#1] from the preferred offsite power source; closing the ElB and ElC MOS units [#1] to the Emergency Bus IL source; closing the Emergency Bus IL feeder breakers [#2] to Engineered Safety Features Buses B and C; and closing the Engineered Safety Features Bus B and C normal supply breaker [#3] to each 4.16 kV Engineered Safety Features Bus. Engineered Safety Features Bus B is then energized from the output ,

of SDG C. The FHB HVAC loads may then be manually actuated and energized by SDG C. l This cross-connect scheme was utilized in the docketed South Texas Project position for Technical Specification Amendments 71 and 60, requested in letter ST-HL-AE-4924 dated November 7,1994 and approved by the NRC in letter ST-AE-HL-94117, dated February 14,1995, i

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Attachment 1 ST-HL-AE-5414 Page 4 of 5 C. Clarify Control Room HVAC single train capability:

Resnome:

While the South Texas Project design bases for Control Room Envelope HVAC is two of three trains, the capabilities of a single train of the Control Room Envelope HVAC system supported by a single train of Essential Chilled Water powered by the same ESF diesel generator are the following. A l single train of Control Room Envelope HVAC will maintain designed temperature and relative '

humidity conditions. However, a single train of the Control Room Envelope HVAC system is not sufficient to maintain the control room envelope at 1/8" positive pressure relative to the surrounding l areas. While single train operation testing has demonstrated most configurations will provide a positive pressure in the control room envelope, it was determined approximately 30 cfm inleakage  ;

could occur under certain single train configurations. This unfiltered inleakage, plus the normally assumed 10 cfm from the use of other doors in the pressurized areas, would yield a total Control l Room inleakage of approximately 40 cfm. l l The results of a sensitivity study of the Control Room thyroid dose to variations of the in-leakage l between 10 cfm and 40 cfm are shown below. The doses represent the TID-14844 source term results

for one train of Control Room Envelope HVAC and consequently include a substantial amount of  ;

conservatism.

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THYROID DOSE RESULTS l

I IN I FAKAGE (CFM) DOSE (REM) 10 36 17 49 25 63 40 90 i

[The results noted above assumed a "once-filtered" inleakage of 250cfm. Results from the

tests indicated that this inleakage value could be as high as 300cfm for certain conditions.

l Use of the 300cfm value would slightly increase the above doses]

When considering the extremely low probability (< E-9)of this beyond design basis configuration,  :

Control Room dosage is considered acceptable given the conservatism in the calculation and that the doses are within the 10 CFR 50, Appendix A General Design Criteria limits when using the more

realistic Westinghouse gap activity source term.

l l Based on the above, South Texas Project considers that the Control Room Envelope HVAC system has sufficient capability in the single train configuration to justify the short allowed outage time

requested in the proposed Technical Specification.

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I Attachmrnt 1 ST-IIL-AE-5414 Page 5 of 5 D. Clarify Reactor Coolant Pump seal cooling capability if B Train ESF diesel is the only l operating diesel generator: '

Response

If the B Train ESF diesel is the only operating diesel generator following a LOOP, no centrifugal charging pump is available to provide seal injection to the Reactor Coolant Pumps. In the event of a loss of offsite power, the B train component cooling water system will provide adequate cooling flow ,

to the Reactor Coolant Pump thermal barrier heat exchanger to maintain the integrity of the Reactor Coolant Pump seals without Operator action. For additional defense in depth, the Positive Displacement Pump may be powered by the Technical Support Center Diesel Generator to provide seal injection to the Reactor Coolant Pumps in the event of a loss offsite power with only B SDG available.

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