NOC-AE-0163, Application for Amends to Licenses NPF-76 & NPF-80, Replacing SG Water Level Trip Setpoint Differences

Application for Amends to Licenses NPF-76 & NPF-80, Replacing SG Water Level Trip Setpoint Differences

ML20236T510
Person / Time
Site:	South Texas
Issue date:	07/22/1998
From:	Cloninger T HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236T512	List: ML20236T514 NOC-AE-0163, Application for Amends to Licenses NPF-76 & NPF-80, Replacing SG Water Level Trip Setpoint Differences
References
NOC-AE-0163, NOC-AE-163, NUDOCS 9807280233
Download: ML20236T510 (16)

Text

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Nuclear Operating Company S

soah rauntyeaMecaeragsum,, ea aam matr raumn m July 22, 1998 NOC-AE-0163 )

File No.: G20.02.01 G21.02.01 l 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington,DC 20555 I l

South Texas Project Units 1 and 2 l Docket Nos. STN 50-498 and STN 50-499 l Proposed Amendment to Technical Specifications to Reflect Replacement Steam Generator Water Level Trio Setnoint Differences

Reference:

Letter from L. E. Martin to U.S. Nuclear Regulatory Commission dated May 7,1998, (ST-NOC-AE-00159) ,

The STP Nuclear Operating Company proposes to amend the South Texas Project Operating Licenses NPF-76 and NPF-80 by incorporating the attached changes into the Technical Specifications. These proposed changes are associated with the Replacement Steam Generator Project. The existing Unit 1 Westinghouse Model E steam generators are currently planned to be replaced with Westinghouse Model A94 steam generators in May,2000. The A94 design characteristics incorporate greater steam generator heat transfer area, secondary mass inventory at Hot Full Power (HFP), as well as increased reactor coolant system flow and volume. These proposed changes are necessary to reflect the steam generator water level low-low trip setpoint differences between Model E and Model A94 steam generators for Reactor Trip System and Engineered Safety Features Actuation System instrumentation. The proposed changes to the Technical Specifications included within are defined in terms of the appropriate steam generator model rather than by the South Texas Project unit in which the steam generators are installed.

1 Further proposed changes to the Technical Specifications pertaining to the A94 steam generators \ \

are currently under development. A general description of these additional proposed changes to l the Technical Specifications, as well as other planned submittals to support licensing for the A94 steam generators, is contained in the above referenced letter. o\ l l The South Texas Project has reviewed the attached proposed amendment in accordance with 10CFR50.92 and has determined that the amendment does not involve a significant hazards consideration. Additionally, it has been determined that the proposed amendment satisfies the i

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, NOC-AE-0163 Page 2 of 3 criteria of 10CFR51.22(c)(9) for categorical exclusion from the requirement for an environmental assessment. The South Texas Project Nuclear Safety Review Board has reviewed and approved the proposed changes.

The required aflidavit, along with a Safety Evaluation and No Significant Hazards Consideration Determination associated with the proposed changes, and the marked-up Technical Specification pages are included as attachments to this letter.

The South Texas Project is providing the State of Texas with a copy of this proposal in accordance with 10CFR50.91(b).

If there are any questions regarding this proposed amendment, please contact either Mr. M. A.

McBurnett at (512) 972-7206 or me at (512) 972-8787.

. H. Clo nge Vice Pr i t, Nucle gme 'ng l BJS/

Attachments: 1. Affidavit

2. Technical Specification Changes

3. Determination ofNo Significant Hazards Consideration

4. Technical Specification Marked-Up Pages j l

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, . NOC-AE-0163 Page 3 of 3 cc: '

Ellis W. Merschoff Jon C. Wood Regional Administrator, Region IV Matthews & Branscomb U. S. Nuclear Regulatory Commission One Alamo Center 611 Ryan Plaza drive, Suite 400 106 S. St. Mary's Street, Suite 700 Arlington, TX 76011-8064 San Antonio,TX 78205-3692 Thomas W. Alexion Institute of Nuclear Power Operations Project Manager, Mail Code 13H3 Records Center

. U. S. Nuclear Regulatory Commission 700 Galleria Parkway Washington, DC 20555-0001 Atlanta, GA 30339-5957 David P. Loveless Richard A. Ratliff Senior Resident Inspector Bureau of Radiation Control

, U. S. Nuclear Regulatory Commission Texas Department of Health P. O. Box 910 1100 West 49th Street Bay City, TX 77404-0910 Austin, TX 78756-3189 J. R. Newman, Esquire D. G. Tees /R. L. Balcom Morgan, Lewis & Bockius Houston Lighting & Power Co.

I800 M Street, N.W. P.O. Box 1700 Washingto 1, DC 20036-5869 Houston,TX 77251 M. T. Hardt/W. C. Gunst Central Power and Light Company City Public Service Attention: G. E. Vaughn/C. A. Johnson P. O. Box 1771 P. O. Box 289, Mail Code: N5012 San Antonio,TX 78296 Wadsworth,TX 77483 J.C.Lanier/A Ramirez U. S. Nuclear Regulatory Commission )

l City of Austin Electric Utility Department Attention: Document Control Desk i 721 Barton Springs Road Washington, DC 20555-0001 Austin,TX 78704 I ,

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NOC-AE-0163 Attachment i I

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AFFIDAVIT l

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• NOC-AE-0163 Attichment 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter )

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STP Nuclear Operating Company ) Docket Nos. 50-498

) 50-499 South Texas Project Units 1 & 2 )

AFFIDAVIT I, T. H. Cloninger, being duly sworn, hereby depose and say that I am the Vice President, Nuclear Engineering of the South Texas Project; that I am duly authorized to sign and file with the Nuclear Regulatory Commission the attached proposed amendment to the Technical Specifications; that I am familiar with the content thereof; and that the matters set forth therein are true and correct to the best of my knowledge and belief, Y1 T.H. Cl inger Vice P esid t, Nuclear ngineering  ;

STATE OF TEXAS ) (

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COUNTY OF MATAGORDA )

Subscribed and sworn to before me, a Notary Public in and for the State of Texas, this M day of CL & ,1998.

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. g% UNDARflTENBERRY

. NowyPutsc,SWeof Toms Notary Public in and for the /

%,,p/ h [ State of Texas

• NOC-AE-0163 Attachment 2 Page 1 of 7 ATTACHMENT 2 TECHNICAL SPECIFICATION CL NGES 1 l

NOC-AE-016 '

Attachment 2 Page 2 of 7 BACKGROUND The South Texas Project is planning on replacing the original Westinghouse Model E Steam Generators with Westinghouse A94 Steam Generators. Unit I steam generator replacement is scheduled to occur at the end of Cycle 9, currently planned for the spring of the year 2000. Unit 2 replacement is scheduled to occur at the end of Cycle 9, currently planned for the year 2002.

j Thus, the South Texas Project will be operating with two different models of steam generators

! for a period of time.

The design and performance differences of the replacement steam generators (Model A94), as compared with existing Model E steam generators, have required evaluation or re-analysis of

! design basis accidents whose inputs are based on any of following:

L e Reactor Coolant System (RCS) volume; e - Reactor Coolant System flow resistance;

.. Steam Generator (SG) vohune;

• Steam Generator metal mass; e ~ Steam Generator heat transfer surface area; or

. . Steam Generator heat transfer coefficient.

l Steam generator water level instrument trip setpoints have been affected by operational considerations that have been incorporated into analytical assumptions. As a result, the steam

, generator water level low-low trip setpoint for the Reactor Trip System (RTS) and the Engineered Safety Features Actuation System (ESFAS) actuation will change with the new steam generators. Technical Specification changes reflecting this new low-low trip setpoint are proposed to reflect these actuation changes.

The differences are labeled in the proposed Technical Specifications to reflect the model of steam generator rather than the unit in which the steam generators are installed. Therefore, these changes in the Technical Specifications are applicable to both units at the South Texas Project.

DESCRIPTION OF PROPOSED CHANGES The proposed Technical Specifications requiring change due to replacement of the existing l Model E steam generators with new A94 steam generators and the resulting water level low-low trip setpoint change are as follows:

1. Reactor Trip System Instrumentation Trip Setpoints - Steam Generator Water Level Low-Low (Table 2.2-1, Function 13), and

2. Engineered Safety Features Actuation System Instrumentation Trip Setpoints - Auxiliary Feedwater Actuation on Steam Generator Water Level Low-Low (Table 3.3-4, Function 6.d).

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b. .. NOC-AE-0163 Attachm nt2 Page 3 of 7 These proposed Technical Specification changes are permanent changes and are discussed as

. follows: 1

!~ l Disensnion The primary function of the Steam Generator Water Level Low-Low instrument trip signals 1 is to ensure that an adequate heat sink is re-established following Loss of Non-Emergency AC Power to the Plant Auxiliaries, Loss of Normal Feedwater Flow, and Feedwater Line Break events by tripping the reactor and by actuation of auxiliary feedwater. .These actions

terminate power operation of the reactor and provide for RCS cooling via the steam generators.

These functions are necessary to mhigate the effects of a decreasing water level in the steam generators, which could result in a loss of heat sink for the reactor. The steam generator low-low water level condition is due to loss of feedwater flow to one or more steam generators.

The three events that rely on the steam generator water level low-low trip signal for accident mitigation (i.e., the Loss of Non-Emergency AC Power to the Plant Auxiliaries, Loss of Normal Feedwater Flow, and Feedwater System Pipe Break events) have been reanalyzed using the Westinghouse RETRAN model (Reference 4). The analysis approach is consistent with the methodology that supports the current South Texas Project licensing bCs where the LOFTRAN computer code had previously been used.

The current steam generator (Model E) water level low-low signal trip setpoint is 33 %

narrow range span (NRS). During the developmental phases of the A94 steam generator, improvements in operational aspects of the steam generators were considered and focused on the following two areas:

. Increasing the operating margin to the steam generator water level low-low reactor trip setpoint; and

. Minimizing the potential for unnecessary Auxiliary Feedwater actuation and excessive RCS cooldown following a reactor trip.

These two considerations have led to a lowering of the steam generator water level low-low

' trip setpoint to 20 % NRS from the current 33 % NRS.

In addition to the steam generator water tevel setpoint changes, the Loss of Non-Emergency AC Power to the Plant Auxiliaries and the Loss ofNormal Feedwater Flow j analyses also required two further operating changes to obtain acceptable results. The first change is a modification to the pressurizer water level program and the second change is 1

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NOC-AE-0163 Attachment 2 Page 4 of 7 crediting automatic operation of the Class 1E safety-grade steam generator power-operated relief valves. Both of these operating changes are described in the Safety Evaluation section.

l The new steam generator water level low-low trip setpoint is determined by the largest value obtained after summing the accident specific safety analysis value and the associated instrumentation uncertainty or the operational consideration of setting the trip setpoint as l low as possible to maximize operating margin. This lowest operational limit is specified as i

20 % NRS, a value associated with the lower deck plate location. Water levels below this value will not provide operational benefits due to the accelerated level drop that would l

occur as the water level drops into the narrow portion of the downcomer. The Loss ofNon-l Emergency AC Power to the Plant Auxiliaries and the Loss of Normal Feedwater saf;ty analysis setpoint utilized is 11 % NRS. The event-specific instrument uncertainty fc r these events is 8.9 % NRS, which totals 19.9 %. The Feedwater Line Break safety analyr is setpoint utilized is 0.0 % NRS. The Feedwater Line Break event-specific instrument

! uncertainty due to adverse containment conditions is 17.7 % NRS, which totals 17.'/ %.

l The selected steam generator water level low-low trip setpoint of 20 % NRS bounds both

! cases and is consistent with the operational limit of 20 % NRS discussed above.

Values associated with the A94 steam generator water level low-low trip setpoint for " Total i

Allowance (TA)", "Z" (where Z is the statistical summation of errors assumed in the safety analysis excluding those associated with the sensor and rack drift and the accuracy of the l measurement), " Sensor Error (S)" and " Allowable Value" have been calculated in accordance with approved methodologies (Reference 1). These values are as follows:

l Total Allowance = 20.0 l Z = 16.7 Sensor Error = 1.9 l Allowable Value 2: 18.0 % NRS l All acceptance criteria were shown to be met for these events.

l Pronosed Changes:

• On Page 2-5, for Function 13 of Table 2.2-1:

1. The existing values for Total Allowance, 'Z', Sensor Error, Trip Setpoint and Allowable Value are made applicable specifically to Model E steam

( generators.

2. New values for Total Allowance, 'Z', Sensor Error, Trip Setpoint and Allowable Value are added applicable to the A94 steam generators.

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NOC-AE-0163 Attachment 2 Page 5 of 7 l

L e On Page 3/4 3-32, for Function 6.d of Table 3.3-4:

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l. The existing values for Total Allowance, 'Z', Sensor Error, Trip Setpoint and l

Allowable Value are made applicable specifically to Model E steam generators.

l l 2. New values for Total Allowance, 'Z', Sensor Error, Trip Setpoint and j Allowable Value are added applicable to the A94 steam generators.

SAFETY EVALUATION l Significant aspects of the safety analysis performed to support this evaluation are as follows:

• The analysis relies on application of the Westinghouse RETRAN model for Non-LOCA safety analysis. A topical report (Reference 2) describing the model and qualification is l under NRC review. The Westinghouse RETRAN model replaces the Westinghouse

. LOFTRAN model for those events that have been re-analyzed.

• ' Minimum measured Reactor Coolant Flow Rate is assumed to be 403,000 gpm.

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e Credit is taken for automatic operation of the Class lE safety-grade steam generator power-operated relief valves for accident mitigation of the Loss ofNon-Emergency AC Power to the Plant Auxiliaries and Loss of Normal Feedwater Flow events. Automatic operation of the safety-grade steam generator power-operated relief valves is currently required for Small Break LOCA accident mitigation and a pending license amendment for this requirement has been submitted (Reference 3). This pending license amendment supports crediting these valves for the Loss ofNon-Emergency AC Power to the Plant Auxiliaries and Loss of Normal Feedwater events, e Full power reactor vessel average temperatures (Tavg) between 582.3 F and 593.0 F continue to be supturted. This assumption preserves the Tavg window which is currently j supported by the South Texas Project analysis of record.

e Steam generator tube plugging levels between 0 % and 10 % continue to be supported.

y e Nominal full power main feedwater temperature operations between 440 F and 390 F will be supported in the engineering evaluation. South Texas Project is currently analyzed for operation with a nominal feedwater temperatures of 440 F. Reducing the feedwater temperature to 390*F bounds the conditions of one feedwater heater out of service.

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, NOC-AE-0163 Attachment 2 Page 6 of 7

. The pressurizer water level program, which is defined in terms of!evel span (% span) as a function of power, has been modified. The modified program is given as a linear ramp between two level values given below for Tavg programs at each end of the full-power average temperature range:

Level at no-load Tavg Level at full-nower Tavg

.High Tavg Program 25 % span 57 % span Low Tavg Program 25 % span 40 % span Modification to the pressurizer water level program compensates for more RCS coolant l expansion (with A94 steam generators) during several secondary system events.

The impact of the A94 steam generator on the limiting accidents that support the steam generator water level low-low trip setpoint was evaluated. These accidents are the Loss of Non-emergency I AC Power to Plant Auxiliaries, Loss of Nominal Feedwater and Feedwater Line Break transient.

These accidents were analyzed utilizing the Westinghouse RETRAN model. The analysis approach is consistent with the methodology that supports the current South Texas Project licensing basis where the LOFTRAN computer code had previously been used. All acceptance l criteria were shown to be met with a safety analysis steam generator water level low-low trip i setpoint of 0.0 % NRS for the Feedwater Line Break and 11 % NRS for the Loss of Non-l Emergency AC Power to Plant Auxiliaries, Loss of Normal Feedwater transients. The selected l value of 20 % NRS for reactor trip and auxiliary feedwater (AFW) actuation includes required uncertainties and is bounded by accident assumptions. Allowable values were calculated using i

approved methodologies in accordance with Reference 1. Therefore, the proposed setpoint and allowable values are acceptable.

The Loss ofNon-Emergency AC Power to the Plant Auxiliaries and Loss ofNonnal Feedwater

Flow analyses required a change to the pressurizer water level program and credited automatic l operation of the safety-grade steam generator power-operated relief valves to obtain acceptable results. A change to the pressurizer water level program was necessary to preclude pressurizer l overfill for these two events. The modified pressurizer water level program has been assessed for l all other non-LOCA licensing basis events and found to either have no effect or provide a very l l~ slight benefit.  !

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IMPLEMENTATION

- This set of Technical Specification changr must be implemented following the replacement of the Unit 1 Model E steam generators with Mod-l A94 steam generators, but prior to Unit 1 entry into MODE 3 with A94 steam generators installed.

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NOC-AE-0163 I Attachmint 2 Page 7 of 7 To allow for timely implementation, the NRC is requested to review and approve these changes by November 1999. As noted previously, these Technical Specifications changes are to be made l applicoble to both units.

REFERENCES l

l 1. WCAP 11273," Westinghouse Setpoint Methodology for Protection Systems - South Texas Project Units 1 and 2," February,1993.

2. WCAP 14882-P,"RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis," June 1997.

l 3. ST-HL-AE-5689, Letter from T. H. Cloninger to NRC Document Control Desk dated August 18,1997," Proposed Amendment ofTechnical Specification 3.7.1.6, Atmospheric Steam Relief Valves".

4. NOC-AE-0156, Letter from T.H. Cloninger to NRC Document Control Desk dated May 7, i 1998," Application for Use of the Westinghouse RETRAN Methodology for the Replacement Steam Generator Project".

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NOC-AE-0163 Attachment 3 Page1of3 J

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ATTACHMENT 3 i

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION i

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NOC AE-0163 Attachment 3 Page 2 of 3 i

. 1 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The South Texas Project has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10 Code of Federal Regulations Section 50 Subsection 92 Paragraph c (10 CFR 50.92 (c)), a proposed amendment to an operating license invalves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

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

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

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

INTRODUCTION South Texas Project proposes to revise the following South Texas Project Units 1 and 2 Technical Specification (TS):

. Table 2.2-1 (functional unit 13), " Reactor Trip System Instrumentation Trip Setpoints" (Steam Generator Water Level - Low-Low);

e Table 3.3-4 (functional unit 6.d), " Engineered Safety Features Actuation System Instrumentation Trip Setpoints" (Steam Generator Water Level - Low-Low).

Installation of the A94 Replacement Steam Generators (RSG) at the South Texas Project Unit I and 2 necessitates changes to the low-low steam generator water level trip setpoint.

NO SIGNIFICANT IIAZARDS ANALYSIS

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

This proposed change includes changing the low-low steam generator water level trip setpoint. The setpoint is being changed to enhance the operational flexibility associated with the RSGs.

The minimum setpoint change proposed in this request establishes controls to ensure that an adequate heat sink is maintained by providing an adequate secondary liquid mass to

! remove primary system sensible heat and core decay heat shortly after reactor trip and initiating auxiliary feedwater flow for long-term cooling. The accidents analyzed for this

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'. NOC-AE-0163 Attachment 3 Page 3 of 3 j

1 requirement are the Loss of Non-Emergency AC Power to the Plant Auxiliaries, Loss of Normal Feedwater and Feedwater Line Break transients. These accidents were analyzed utilizing the Westinghouse RETRAN model. All acceptance criteria were shown to be met for both these events. Therefore, the proposed steam generator water level low-low trip setpoint change is demonstrated not to result in an increase in the consequences for these accidents.

The steam generator water level low-low trip setpoint is not considered a precursor to any of the analyzed accidents, and therefore, these proposed changes do not result in an increase in the probability or consequences of any accident previously analyzed.

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

The proposed setpoint change does not create any new operating conditions or modes.

The proposed change only revises the actuation setpoints for the Reactor Trip System and Engineered Safety Features Actuation System. The actions of these systems continue to be performed in accordance with existing requirements, which are sufficient to ensure plant safety is maintained.

3. The proposed change does not involve a significant reduction in a margin of safety.

The events potentially affected by the setpoint change in the steam generator water level low-low reactor trip (Table 2.2-1, Function 13) and ESFAS Auxiliary Feedwater System 1 actuation (Table 3.3-4, Function 6.d) are the Loss of Normal Feedwater and Feedwater System Pipe Break. These events were analyzed and it was demonstrated that all acceptance criteria were met for both of these events.

Based on the above evaluation, South Texas Project concludes that the proposed change to the Technical Specifications involves no significant hazards consideration.

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NOC-AE-0163 Attachment 4 Page 1 of 4 i

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TECHNICAL SPECIFICATION MARKED-UP PAGES l

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STEAM GENERATOR WATER LEVEL TRIP SETPOINT CHANGES I