ML20198A218

Revised TS Pages,Adding Ref to TS Figure 2.1-2 to Bases Section 2.1.1 as Figure Provides DNBR Curves Appropriate to Alternate DNB Operating Criteria

ML20198A218
Person / Time
Site:	South Texas
Issue date:	12/10/1998
From:	HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20198A214	List: ML20198A218 NOC-AE-000348, Forwards Revised TS Bases Pages Changed Per 10CFR50.59.Ref Was Added to Bases Section 2.1.1 Because Figure Provides DNBR Curves Appropriate to Alternate DNB Operating Criteria
References
NUDOCS 9812160182
Download: ML20198A218 (2)

Text

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SAFETY LIMITS BASES -

2.11 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 THEPMAL 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 uniform 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.

The DNB design basis is as follows: uncertainties in the WRB-1 correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and 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 II events. This  !

establishes a design DNBR value which must be met in plant safety analyses using values of input l parameters without uncertainties. In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor l 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, r"an , and a reference cosine with a peak of 1.61 for axial power shape. An allowance is included for an increase in r"an at reduced j power based on the expression:

r# n=rMPa a g [g, pg ay (g _p)]

where: R rPan is the limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR): i PFan is the Power Factor Multiplier for r"an 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 i(delta I) 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.M Unit 2 - Amendment No. 60 18245-97 9812160182 981210 L PDR ADOCK 05000498 p PDR .

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POWER DISTRIBU flON LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

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

r"an will be maintained within its limits provided Conditions a. through d. above are maintained. The combination of the RCS flow requirement (TS 3.2.5) and the requirement on r"an guarantees that the DNBR used in the safety analysis will be met. The relaxation of r"an as a function I of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When r"an is measured, no additional allowances are necessary prior to comparison with the limit. A measurement error of 4% for r"an 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.

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SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-4 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. -50 l

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