Text

Proposed Tech Specs Pages,Reflecting Replacement SG Reactor Coolant Flow Differences
	ML20247F332
Person / Time
Site:	South Texas
Issue date:	05/07/1998
From:	; HOUSTON LIGHTING & POWER CO.
To:	;
Shared Package
ML20247F331	List: ML20247F332; NOC-AE-0080, Application for Amends to Licenses NPF-76 & NPF-80 to Reflect Replacement SG Reactor Coolant Flow Differences;
References
	NUDOCS 9805190289
	Download: ML20247F332 (9)
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NOC-AE-0080 Attachment 4 Page 1 of 9 ATTACHMENT 4 l

l TECHNICAL SPECIFICATION MARKED-UP PAGES RELATED TO i REACTOR COOLANT SYSTEM FLOW CHANGES 38R52%8881!!888be P PDR

NOC-AE-0080 Attcchment 4 Page 2 of 9 l

PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS 1

The below listed Technical Specification pages are marked-up to identify the proposed changes due to RCS flow differences and are included in this attachment. (Note: Page 2-2a is a new page to be added to the Technical Specifications). '

Pages i

2-1 ,

2-2 2-2a l

2-4 B 2-1 3/4 2-11 B 3/4 2-6 i

1 l I 1

2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l l

2.1' SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, operating loop coolant temperature shall (T,y9) not exceed pressurizer the limits shown pressure, in and I Figure 2.1-1, Soc. Model E Sfe9, cenerpforS y or in Fo*gure 2. l-2 PLIC B LIT 0 M$DYS1anY2 / OY '

l ACTION:

Whenever the point defined by the combinat' ion of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be l in HOT STANDBY with the Reactor Coolant System pressure within its limit ;

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1. )

MDDES 3, 4 and 5: ;

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 1 5 minutes, and comply with the requirements of Specification 6.7.1.

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FIGURE 2.1-1 REACTOR CORE SAFETY UMIT - FOUR LOOPS IN OPERATIONj Model E Shmm Graenafoe.s SOUTH TEXAS - UNITS 1 & 2 2-2 Unit 1 - Amendment No. /,61 Unit 2 - Amendment No. 50 I

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0 0.2 0.4 0.6 0.8 1 1.2 FRACTION OF RATED THERMAL POWER Figure 2.11 [Page 1 of 1)

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- 2. for A 4 Sle% yene< dors ATETY LlMITs foye .
BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and l
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 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 11 events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.
In addition, margin has been maintained in the design by meeting safety lysis DNBR limits in oerforming safety analyses.
ThecurvesofFigure2.1-IfshowthelociofpointsofTHERMALPOWER, !
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.
i These curves are based on a nuclear enthalpy rise hot channel factor, !
and a reference cosine with a peak of 1.61 for axial power shape. An F"To,wance al is included for an increase in F"a at reduced power based on the expression:
F"a - F"'"a p + PF, WP)] l where: is the limit at RATED THERMAL POWER (RTP) specified F"'"
in die CORE OPERATING LIMITS REPORT (COLR);
PF specified in the C0 h;isand, the Power Factor Multiplier for F"a 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 3 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.
8 2-1 Unit 1 - Amendment No. 61 50VIH TEXAS - UN115 1 & 2 Unit 2 - Amendment No. 50
l.
POWER DISTRIBUTION LIMITS h/4.2.5 DNB PARAMETERS .
LIMITING CONDITION FOR OPERATION 3.2.5 The following DN8-related parameters shall be maintained within the limits following: ,
' a. Reactor Cool' ant System T , 5 598'F Pressurizer P'ressure, > 2189 psig*
b.
c. Reactor Coolant System Flow, 2 392,300 gpm**([or Model E S l APPLICABILITY: MODE 1.
3 %3po gp#h 444 deem Cf'#e #'T)
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
l SURVEILLANCE RE0VIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The provisions of Specification 4.0.4 are l not applicable for verification that RCS flow is within its limit.
4.2.5.2 The RCS flow rate indicators shall be subjected to a channel calibration at least once per 18 months.
4.2.5.3 The RCS total flow rate shall be determined by precision heat balance measurements at least once per 18 months. The provisions of Specification 4.0.4 are not applicable.
l i l
l j
Limit not applicable during either a Thermal Power ramp in excess of 5% of '
RTP per minute or a Thermal Power step in excess of 10% RTP.
Includes a 2.8% flow measurement uncertainty.
l 3/4 2-11 Dait 1 - Amendment No. 61 SOUTH TEXAS - UNITS 1 & 2 Unit 2 - Amendment No. 50 >
l POWER DISTRIBUTION LIMITS BASES 4
3/4.2.5 DNB PARAMETERS (Continued) initial FSAR assumptions and have been analytically demonstrated adequate to maintain a l
l minimum DNBR of greater than or equal to the design limit throughout each analyzed transient.
l The T.,, value of 598'F and the pressurizer pressure value of 2189 psig are analytical values. The l readings from four channels will be averaged and then adjusted to account for measurement uncertainties before comparing with the required limit. The flow requirement (002,?m cr=)
includes a measurement uncertainty of 2.8%.
l The 12-hour periodic surveillance of these parameteri through instrument readout is sefficient to ensure that the parameters are restored within their limits following load changes and ,
other expected transient operation.
l (392,300 gpm %c mode l E Steam gedernhorr ;
H 03, 000 ym [w- ATI SW geneal0er l
l l
1 SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-6 Uni:1-Amcadmen;Nn.5 1108-95 Uni: 2 - Amendmca; Nn. 50 l
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v t e Letter Sequence Request
TAC:MA0967, (Open)
Initiation Request, Request, Request, Request Acceptance... Administration Questions 18 May 1998 Answers Results Approval... Other:
MONTHYEAR1998-05-18 [Table View]

ML20247F332

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