ML20202D067

Responds to Util ,Providing Update to TS Bases Pages B 2-1 & B 3/4 2-4.Confirms That NRC Agrees with Proposed Changes Because Changes Present More Complete Discussion of Safety Limits for Reactor Core

ML20202D067
Person / Time
Site:	South Texas
Issue date:	01/27/1999
From:	Alexion A NRC (Affiliation Not Assigned)
To:	Cottle W HOUSTON LIGHTING & POWER CO.
References
TAC-MA4402, TAC-MA4403, NUDOCS 9902010284
Download: ML20202D067 (8)

Text

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~e 3  % UNITED STATES

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• • • • • January 27, 1999 l l

2 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station l

P. O. Box 289 Wadsworth,TX 77483

SUBJECT:

UPDATE TO TECHNICAL SPECIFICATION BASES PAGES B 2-1 AND i B 3/4 2-4, SOUTH TEXAS PROJECT, UNITS 1 AND 2 (STP) (TACS NOS. j MA4402 AND MA4403)  !

Dear Mr. Cottle:

i The purpose of the letter is to respond to STP Nuclear Operating Company's December 10, 1998, letter which provided an update to Technical Specification (TS) Bases Pages B 2-1 and B 3/4 2-4. The change adds a reference to TS Figure 2.1-2 to Bases Section 2.1.1 because i this figure provides the departure from nucleate boiling heat flux ratio (DNBR) curves i appropriate to the alternate departure from nucleate boiling (DNB) operating criteria. Also, a change was made to Bases Section 3/4.2.2 and 3/4.2.3 by referencing TS 3.2.5 rather than an l explicit reactor coolant system (RCS) flow rate since the minimum flow rate required will be different if the alternate flow rate option is implemented. The letter did not request any approval or response from the Nuclear Regulatory Commission (NRC). However, this letter is provided to confirm that NRC agrees with the proposed change.

The NRC staff agrees with both of the proposed changes because the changes present a more complete discussion of the safety limits for the reactor core, by referencing both the primary and attemate " Reactor Core Safety Limit - Four Loops in Operation" curves and referencing TS 3.2.5 in place of a specific RCS flow rate. TS 3.2.5 provides for a reduced RCS flow rate with a reduced RCS T,,. Referencing this Specification more accurately reflects the limits of the RCS flow rate parameter.

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Eu William T. Cottle Enclosed for your information, are revised TS Bases pages B 2-1 and B 3/4 2-4 that the NRC staff will use to update NRC's copy of the Bases. The revised pages contain marginallines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness. If you have any additional questions regarding this issue, please contact me at (301) 415-1326.

Sincerely, ORIGINAL SIGNED BY:

Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

Bases Pages B 2-1 and B 3/4 2-4 cc w/encls: See next page DISTRIBUTION:

. Docket File PUBLIC PD4-1 r/f EAdensam (EGA1) JHannon CHawes l TAlexion OGC ACRS I

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Document Name:G:\STPFINAL\STPA4402.WPD ,

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JHan on DATE i / II /99  ! /b/99 l / ll /99 // 23/99 //2 1 /99 COPY / YEhNO [Yh/NO YES/NO YES/NO F /NO OFFICIAL RECORDYOPY g; l

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1 William T. Cottle Enclosed for your information, are revised TS Bases pages B 2-1 and B 3/4 2-4 that the NRC staff will use to update NRC's copy of the Bases. The revised pages contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness. If you have any additional questions regarding this issue, please contact me at (301) 415-1326.

Sincerely, fW ,

Thomas W. Alexion, Project Mdnager l', ..

45W Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

Bases Pages B 2-1 and B 3/4 2-4 cc w/encis: See next page

. a Mr. William T. Cottle  :

I STP Nuclear Operating Company South Texas, Units 1 & 2 cc:

t Mr. Cornelius F. O'Keefe Jack R. Newman, Esq.  :

Senior Resident inspector Morgan, Lewis & Bockius '

U.S. Nuclear Regulatory Commission 1800 M Street, N.W. 1 P. O. Box 910 Washington, DC 20036-5869 Bay City, TX 77414 Mr. T. H. Cloninger Vice President A. Ramirez/C. M. Canady Engineering & Technical Services ,

City of Austin STP Nuclear Operating Company Electric Utility Department P. O. Box 289 l

721 Barton Springs Road Wadsworth,TX 77483 l

Austin,TX 78704 Office of the Governor Mr. M. T. Hardt ATTN: John Howard, Director Mr. W. C. Gunst Environmental and Natural City Public Service Board Resources Policy P. O. Box 1771 P. O. Box 12428 San Antoaio,TX 78296 Austin,TX 78711 Mr. G. E. Vaughn/C. A. Johnson Jon C. Wood Central Power and Light Company Matthews & Branscomb P. O. Box 289 One Alamo Center Mail Code: N5012 106 S. St. Mary's Street, Suite 700 Wadsworth,TX 74483 San Antonio,TX 78205-3692 INPO Arthur C. Tate, Director Records Center Division of Compliance & Inspection 700 Galleria Parkway Bureau of Radiation Control ,

Atlanta, GA 30339-3064 Texas Department of Health l 1100 West 49th Street Regional Administrator, Region IV Austin, TX 78756 i U.S. Nuclear Regulatory Commission i 611 Ryan Plaza Drive, Suite 400 Jim Calloway Arlington,TX 76011 Public Utility Commission of Texas Electric Industry Analysis '

D. G. Tees /R. L Balcom P. O. Box 13326 Houston Lighting & Power Co. Austin, TX 78711-3326

~ P. O. Box 1700 l Houston,TX 77251 1

l Judge, Matagorda County Matagorda County Courthouse 1700 Seventh Street

[ Bay City, TX 77414 l

l t _ _ __

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifcrm and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

l The DNB design basis is as follows: uncertainties in the WRB-1 correlation, plant 1 operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and I computer codes are considered statistically such that there is at least a 95 percent probability with a 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and ll events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertaintles, in addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 and 2.12 show the loci of points of THERMAL POWER, I Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

These curves are based on a nuclear enthalpy rise hot channel factor, F"w, and a reference cosine with a peak of 1.61 for axial power shape. An allowance is included for an increase in F"as at reduced power based on the expression:

F", = F*w [1 + PF as (1-P))  ;

where: F"w is the limit at RATED THERMAL POWER (RTP) specife in the CORE OPERATING LIMITS REPORT (COLR);

PF w is the Power Factor Multiplier for F"a specified in the COLR; and, P is the fraction of RTP.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming axial imbalance is within the limits of the f, (delta 1) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

SOUTH TEXAS - UNITS 1 & 2 B 2-1 Unit 1 - Amendment No. 6+,

2 ReviseNy 7,g,e"N[N:r Ik7/99

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SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3110 psig) of design pressure, to l demonstrate integrity prior to initial operation.

l SOUTH TEXAS - UNITS 1 & 2 B 2-2

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INplCAffD AXIAL FWX DIFFERENCE i I

FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE YERSUS THERMAL POWE SOUTH TEXAS - UNITS 1 & 2 8 3/4 2-3

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POWER DISTRIBUTION LIMITS BASES HEIf_l.UX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANbEL, FACTOR (Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3 9 are maintained; and 1

d. The axial power distribution, expressed in terms of AXlAL FLUX DIFFERENCE, is maintained within the limits.

F", will be maintained within its limits provided Conditions a. through d. above are e maintained. The combination of the RCS flow requirement (TS 3.2.5) and the requirement on i F% guarantees that the DNBR used in the safety analysis will be met. The relaxation of F",

as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When F", is measured, no additional allowances are necessary prior to comparison with the limit. A measurement error of 4% for F", has been allowed for in the determination of the design DNBR value.

Fuel rod bowing reduces the value of DNB ratio. Margin has been maintained between the DNBR value used in the safety analyses and the design limit to offset the rod bow penalty and other penalties which may apply.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-4 Unit 1 - Amendment No. 6t, Unit 2 - Amendment No. 50 Revised by letter dated: 1727/99

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