ML20195E421
| ML20195E421 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/07/1999 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20195E418 | List: |
| References | |
| NUDOCS 9906110091 | |
| Download: ML20195E421 (48) | |
Text
NOC-AE-000471 Page 1 of 21 ATTACHMENT 4 PROPOSED CHANGES TO THFs TECHNICAL SPECIFICATIONS L
4 9906110091 99o607 PDR ADOCK 0500o498 j'
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i-INDEX 2 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY l IMITS 2.1.1 REACTOR CORE.................................................
2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE...........................
2-1 F!GU".5 2.1 * "5 ACTOR OCRE SAFE"' Lif.i.i-FOUR LOOPC lN OPC"".TlON.
2-2 f!OURC 2.1-2 RCAOTOR OORC SAFi'. f L".i. -FOUR LOOPS 'N OPC. ^,T!ON T
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.m.m, t qn; 9 7 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS.............
2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..
2-4 BASES SECTION PAGE 2.1 SAFCTY LIMITS 2.1.1 REACTO R CORE.................................................
B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..........................
B 2-2
? ? t IMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS.............
B 2-3 l
SOUTH TEXAS-UNITS 1 & 2 ill Unit 1 - Amendment No.-97--
Unit 2-Amendment No. 84--
L-
r 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 4g_ Cort OperaDg mi44 kffort b
REACTOR CORE
- 2.1.1 The ' combination of THERMAL POWER, pressurizer pressurg and the highest operating loop coolant temperature (T ) shall not exceed the limits shown irf........,,.,,, r;gure 5
2.12 2.:.. :;1. c.g =tr..t:.::: Opc 3r.g :. ':.".: cer.-!:ter.t 2. reic:d n::::st Oce:er,t Oyr;..T. T.ew a ;ddrand b T:2...'s' Sp;".s:,,0.2.5.
Te ser4 1 M APPLICABillm MODES 1 and 2.
ACTION:
I Whenever the point defined by the combination of the highest operating loop ave' rage temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRFRSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY: MODES 1,2,3,4, AND 5.
I ACTION:
- MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1, MODES 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 5 minutes, and comply with the requirements of Specification 6.7.1.
SOUTH TEXAS-UNITS 1 & 2 2-1 Unit 1 - Amendment No.97 Unit 2-Amendment No.84
o Insert 1 2.1.1.1 In MODES 1 and 2, the departure from nucleate boiling ratio (DNBR) shall be maintained 21.17 for the WRB-1 DNB correlation.
l 2.1.1.2 In MODES I and 2, the peak fuel centerline temperature shall be.
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C5ueIR$azPE
SAPETY LIMITS BASES-2.1.1 REACTOR CORE i
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 i
transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation
. fassenaretsary, Operation above the upper ha==da y of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departune from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable perect 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.
. Ihe 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 ar.d II events. This gg y establishes a design DNBR value which must be met in plant safety analyses using values ofinput parameters without uncertainties. In addition, margin has been maintained in the design by meeting 2
safety analysis DNBR limits in performing safety analyses.
y
'Ihe curves of Pigures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor
[
lant System pressure and average temperature below which the calculated DNBR is no less than t[h desi R value or the average enthalpy at the vessel exit is less than the enthalpy of saturat uid.
'Ihese cury based on a nuclear enthalpy rise hot channel factor, r"an,and
- crence cosine with a peak of 1.61 for power shape. An allowance is included for an inc in r"an atreduced powerbksed on the expressio F"an=F"4sll+PF 4
-P)]
e where: r an is thelimit at RATED POWER (RTP) specifiedin the CORE OPERATING LIMITS (CO PFan is the Power orMultiplierfor r"an ified in the COLR; and, Pis the f of RTP.
'Ibese ' ' ' g heat flux conditions are higher than those calculated for the of allcontrol l
rods thdrawn to the maximum allowable control rod insertion assuming axial i
=1= ace is
' the limits of the f (delta I) function of the Overtemperature trip. When the axial po mhalanca is not within the tolerance, the axial power imbalance effect on the Overtoi+4-deka 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
- ! C" ^~' -
==rmusac
.somm=
iz _.
r i
Insert 2
'The reactor. core Safety Limits are established to preclude violstion of the following fuel design criteria:
- a. = There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
- b. There' must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.
The reactor core Safety Limits are used to define the various Reactor Protection
- System (RPS) functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences
- (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot
{
leg is less than or equal to the saturation enthalpy and that the core exit quality is y
within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that, for variations in the Thermal Power, RCS Pressure, RCS average temperature, RCS flow rate, and AI, the reactor core Safety Limits will be satisfied during steady state operation, normal operational transients, and AOOs.
e.
LIMITING SAFETY SYSTEM SETTINGS-i
(
BASES' t
Intermediate'and Source Rance', Neutron Flux The' Intermediate and Source Range, Neutron Flux trips provide core i.
protection during reactor.startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical l
condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channelsJ The Source Range channels will
. initiate' a Reactor trip at about 10s counts per second unless manually blocked
. hen P-6 becomes-active. The Source Range channels are automatically blocked w
'above P-10.
The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurize'r High and Low Pressure trips. The Setpoint is auto-matically varied with:
(1) coolant temperature to' correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial Ac Core power distribution, this Reactor trip limit is always below the core Safety Operm+
Limit as shown if Figm e 2.1 1.
If axial peaks are greater than design, as
- gg$,epst, indicated by the difference between. top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1R and as spec _i fied in 44 (_, ort - 6 emM
.R4cet.
~ '
l-im M5 I
- Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions,. limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from.the core.to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceede.d. The Overpower AT trip provides
. protection to mitigate the consequences of various size steam breaks as
' reported in WCAP-9226, " Reactor Core Response to Excessive-Secondary Steam
~
. Releases."
SOUTH TEXAS - UNITS 1 & 2 B 2-5 I
i L.
. LIMITING SAFETY SYSTEM SETTINGS
. BASES l ~
Pressurizer Pressure-L
. In each of the pressurizer pressure channels, there'are two independent bistables, each with its own tri l
trip thus limiting the pressure.p setting to provide for a High and Low Pressure range in which reactor operation is permitted.
cE2 The Low.Setpoint trip protects >against low pressure which could lead to DNB by tripping the reactor. in the event of a loss of reactor coolant pressure.
On decreasing power, the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately.10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power,. automatically. reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer
. relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The Pressurizer High Water Level trip is provided to-prevent water relief through the pressurizer. safety valves. On decreasing power, the Pressurizer High Water' Level trip is-automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with.a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, auto-
.matica11y reinstated by P.-7.
Reactor Coolant Flow The Low Reactor. Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of
' RATED THERMAL' POWER or a turbine. impulse chamber pressure at approximately 10%
.of full power equivalent), an. automatic Reactor trip will occur if the flow in more than.one loop drops below approximately 92% of nominal full loop flow.
Above P-8 (a. power level of approximately 40% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below-approximately 92% of' nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7, an automatic Reactor trip will. occur on low reactor
- coolant flow in more than one loop, and below P-7 the trip function is auto-l matica11y blocked. Tk M\\u.g, for- \\oop d65%f\\ -flow is %L AAA h ehe K W M M at W @ % Lh MM d65Q
- steam Generator Water
-QOw Q SSKm e d >
D N 6 Ana lv si.s.
The Steam Generator Water Level Low-Low trip protects the reactor from i.
' loss of heat sink in the event.of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater.
The-specified Setpoint provides 4
J'
. allowances for starting delays of the Auxiliary Feedwater System.
1
' SOUTH TEXAS - UNITS 1 &'2-B 2-6 u
3 i
POWER DISTRIBUTION LIMITS 3/4.2.5 DNS PARAMETERS 4 LIMITING CONDITION FOR OPERATION i
3.2.5 The following DNS-relakd parameters shall be maintained within the limits following:
Reactor Coolant System T s 699V OC MI+ A# QdcifNd ^ +k 0 #d-a.
O craRA3 *'4J OFW l
(= ; ""!'F nr..tr::d PTS T;;;' O.2.0.c),
t
'. 4hc b'mi4 As specified to A Core Qem+ing Qwmsl O ess'saPressurizer Pressure, > 100 p%q mo 1 imids R ef or4'
)
b.
c.
Peactor Coolant System Flow,2 399;896gpm
(; a000,000 sp.T.".40. redos.d ROO T, ol 3.2.5 )
APPLICABILITY: MODE 1.
ACTION:
3,.
With any of the abovo parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILI.ANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least i
once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.~ The provisions of Specification 4.0.4 are not applicable for verification that RCS flowis 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.
NOTE SR 4.2.5.3 is required at beginning-of-cycle with reactor power 290% RTP.
'4.2.5.3 The RCS total flow rate shall be deterihined by precision heat balance or elbow tap AP l
measurements at least once per 13 months. The provisions of Specification 4.0.4 are not applicable.
" U..;' i,vi @ d..ir.g e 2.es e h.cr.e: Ic;;.-.r.p b e:zzx cf 5% cf '
n T P p u sr.lr.ete a e O e.Tr.e: Ic;;e;ep b e--:eaa ei 10% RTF.
' " h i d r : 2.0*',T. r ;. n : x.-. = : = c.t '.t i.
SOUTH TEXAS - UNITS 1 & 2
-3/4 2-11 Unit 1 - Amendment No. 0+;97,W
. Unit 2 - Amendment No. 00,0t*5 -
POWER DISTRIBUTION LIMITS
-BASES HEAT FLUX HOT CHANNEL FACTOR and HUCLEAR ENTHALPY RISE HOT CHANNEL FACIOR (Continued)
When'an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, Fxy(Z), is measured periodir. ally to provide
-assurance that the Hot channel Factor, F (Z), remains within its limit. The 9
F limit for RATED THERMAL POWER (F RTP) as provided in the Core Operating L itsReport(COLR)perSpecification6.9.1.6wasdeterminedfromexpected power control manuevers over the full range of burnup conditions in the core.
)
3/4.2.4' QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion. satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and periodically during power operation. -
The limit of 1.02,'at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for' operation with a tilt condition greater than 1.02 is-provided to allow identification and correction of c dropped or i
misaligned control rod. In the event such action does not correct the tilt, is reinsteted by reducing the maximum allowed the margin for uncertainty on Fq
-power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one e'xcore the moveable incore detectors are used to confirm that detector is inoperable the normalized symmetric power distribit on is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
-3/4.2.5' DNB PARAMETERS
- T A/S E P J 3
The U-"< an the 1)NB-related parameters assure that each af +ha a=_...
1 are maintained within tie nor-i n:_'
. L L ;rfe!ana of operation assumed in-2+ha +-=ha anu accident analyses. The limits are' consistent wiin L i
- SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 Unit 1 - Amendment No. &
Unit 2 - Amendinent No. &
Insert 3 In MODE 1, the limits on pressurizer pressure, RC3 coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced.
I coolant flow or other DNB limiting transient. In all other MODES, the power level-is low enough that the DNB is not a concern.
1 L
The values presented in the COLR are indicated values and include measurement i
L uncertainties. The value for pressurizer' pressure is averaged using plant j
L
' computer /QDPS readings from a minimum of at least 3 channels. The value for -
RCS coolant average temperature is averaged using control board readings from a l
minimum of at least 3 channels.' The value for RCS flow rate is the average from a-i minimum of at least 2 flow transmitters per RCS loop using plant computer /QDPS points.
The value for RCS flow rate presented in Technical Specification 3.2.5 is the thermal design reactor coolant system flow rate used in the analysis approved by the Nuclear Regulatory Commission in Amendments 97 and 84 on September 29, 1998. This flow rate is an analytical limit consistent with 10% plugging of the steam generator tubes and Departure from Nucleate Boiling requirements.
I I-l l
l
F POWER DISTRIBUTION LIMITS
-BASES i
3/4.2.5 DNB PARAMETERS (Continued) -
' 'tial FSAR assumptions and have been analytically demonstrated adequate to maintain a
-mim NBR of greater than or equal to the' design limit throup,hout each analyzed tent.
The T.vg valu 98 F and the pmssurizer pressure value of 2189 psig are an al values.
The readings from at ree channels will be averaged and then a to account for l
measurement uncertainties be mparing with the require it. The flow requirement (392,300 gpm) includes a measuremen rtainty o Technical Specification 3 rovides for aL ate minimum measured Reactor Coolant System flow limit c
' ent with plugging up to 10% o generator tubes and Departure from Nuc oiling requirements. When using the alternate um flow limit, the Tavs i uced to 595 F for Reactor Coolant System flow no less than 3 pm.
l Se and constant values for OPAT and OTAT are also revised accordingly when this altemate mode of operation is entered.
The RCS flow measurement uncertainty of 2.8% bounds the precision heat balance and the elbow tap Ap measurement methods. The elbow tap Ap measurement uncertainty presumes that elbow tap Ap measurements are obtained from either QDPS or the plant process computer. Based on instrument uncertainty assumptions, RCS flow measurements using either the precision heat balance or the elbow tap Ap measurement methods are to be performed at greater than or equal to 90% RTP at the beginning of a new fuel cycle. The elbow tap Ap RCS flow measurement methodology is described in ST-EAE-5707, " Proposed Amendment to Technical Specification Table 2.2-1 and 3/4.2.5 for Reactor Coolant System Flow Monitoring -
Revised,1" dated August 6,1997, and in ST-EAE-5752, " Amended Response to Request for
' Additional Information on the Propose:i Elbow Tap Technical Specification Change (Table 2.2-1
- and Section 3/4.2.5), " dated September 18,1997.
The 12-hour periodic surveillance of these parameters through instrument readout is
' sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation, i
I l
l SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-6 Unit 1 - Amendment No. 61,97,108 Unit 2 - Amendment No. 50, Si,95 3429-98
-7730-99 1
ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of' Resource Manage-ment, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy i
to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle, or any part of a reload cycle for the following:
NNI Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3/4.1.1.3, q,t".
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,
(.B'.
Control Bank Insertion Limits for Specification 3/4.1.3.6,
(, 4'.
Axial' Flux Difference limits and target band for Specification 3/4.2.1, 7 fr. Heat Fg) Hot Channel Factor, K(Z), Power Factor Multiplier, and (F for Specification 3/4.2.2, and y
S.6'.
Nuclear Enthalpy Rise Hot Channel Facter, and Power Factor g
Multiplier for Specification 3/4.2.3.
The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.
6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1.
WCAP 9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July,1985 (H Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6 -
Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor >K)3.2.S - ONB2.w hk+y LimcisParsmE+2.2)- Lim afdy S
Systets SLAhsss and ers.
1.A.
WCAP 12942-PsA, " SAFETY EVALUATION SUPPORTING A MORE
' NEGATIVE E0L MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNITS 1 AND 2."
SERr (o 3
SOUTH TEXAS - UNITS 1 & 2 6-21 Unit 1 - Amendment No. 4s-G, % W l
Unit 2 - Amendment No. +r M % W L
)
Insert 4 1.
Safetp limits for thermal power, pressurizer pressure, and the highest operating loop coolant temperature (Toy,) for Specification 2.1, 2.
. Limiting Safety System Settings for Reactor Coolant Flow-Low Loop design flow,'Overtemperature AT, and Overpower AT setpoint parameter values for 1
Specification 2.2,
)
i I
Insert 5 l
9.
DNB related parameters for Reactor Coolant System T,yg and Pressurizer Pressure, for Specification 3/4.2.5.
Insert 6 1.B.
WCAP-8745-P-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary Class 2)
J l
- k. _
NOC-AE-000471 Page 1 of 14 i
l l
l
.NITACHMENT5 l
UNIT 1 CYCLE 9 CORE OPERATING LIMITS REPORT (TYPICAL)
L
SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNIT 1 CYCLE 9 CORE OPERATING LIMITS REPORT (TYPICAL) i i
I South Texas Project i
Unit 1 Cycle 9 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 9 has been prepared in l
accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-appmved methodologies specified in Technical Specification 6.9.1.6.
The Technical Specifications affected by this report are:
1) 2.1 SAFETY LIMITS 2) 2.2 LIMITING SAFETY SYSTEM SETTINGS 3) 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT LIMITS
)
4) 3/4.1.3.5 SHUTDOWN ROD INSERTION LIMITS 5) 3/4.1.3.6 CONTROL ROD INSERTION LIMITS 6) 3/4.2.1 AFD LIMITS 7) 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR 8) 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 9)-
3/4.2.5 DNB PARAMETERS 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented below.
2.1 SAFETY LIMITS (Specification 2.1) 2.1.1. The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,vg) shall not exceed the limits shown in Figure 1, or in Figure 2 when operating under alternate operating criteria consistent with reduced Reactor Coolant System flow as addressed in Technical Specification 3.2.5.
2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2) 2.2.1 The Over Temperature AT and Over Pressure AT setpoint parameter values are listed below:
Core Operating Limits Report Page 1 of 12 5/10/99 L
m South Texas Project Unit ! Cycle 9 Overtemperature AT Setpoint Parameter Values T ; measured reactor vessel AT lead / lag time constant, Ti 2 8 sec i
T ; measured reactor vessel AT lead / lag time constant, T s 3 see 2
2 T ; measured reactor vessel AT lag time constant, T = 0 see 3
3 T4; measured reactor vessel average temperature lead / lag time constant, T4.:t 28 sec
)
l T ; measured reactor vessel average temperature lead / lag time constant, T5 s 4 see 5
T6; measured reactor vessel average temperature lag time constant, T = 0 sec 6
K ; Overtemperature AT reactor trip setpoint, Ki s 1.14; or, Ki s 1.13 for alternate i
operation with reduced RCS flow K ;Overtemperature AT reactor trip setpoint T,yg coefficient, K 2 0.028/ F 2
2 K ;Oveitemperature AT reactor trip setpoint pressure coefficient K3 2 0.00143/psig 3
T' ; Nominal full power T,vg, T's 593.0 F (or 590. O F for alternate operation with reduced RCS flow)
P' ; Nominal pressurizer pressure, P' 2 2235psig f (AI) = is a function of the indicated difference between top and bottom detectors i
of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that; i
(1) For gi - qu between -70% and +8%, or +6% for alternate operation with reduced RCS flow, fi(AI) = 0, where qi and qu are pement RATED THERMAL POWER in the top and bottom halves of the core respectively, and gi + qu is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each pement that the magnitude of gi-qu exceeds -70%, the AT Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of gi - q3 exceeds +8%, or +6%
for alternate operation with reduced RCS flow, the AT Trip Setpoint shall be automatically reduced by 2.65% of its value at RATED THERMALPOWER.
Core Operating Limits Report Page 2 of 12 5/10/99 L
l South Texas Project l
Unit 1 Cycle 9 Overpower AT Setooint Parameter Values l
T ; measured reactor vessel AT lead / lag time constant, ti 2 8 see i
T ; measured reactor vessel ATlead/ lag time constant, T s 3 sec 2
2 T ; measured reactor vessel AT lag time constant, T = 0 see 3
3 T ; measured reactor vessel average temperature lag time constant, T6 = 0 see 6
T7 ; Time constant utilized in the rate-lag compensator for T vg, T z 10 see 7
L ; Overpower AT reactor trip setpoint, Ls 1.08; (or Ls 1.07 for altemate operation with reduced RCS flow)
K ; Overpower AT rextor trip setpoint T vg rate / lag coefficient, K 2 0.02/ F for 5
5 increasing average temperature, and K 2 0 for decreasing avenige temperature 5
K6 ; Overpower AT reactor trip setpoint T vg heatup coefficient K 2 0.002/ F for T 6
> T" and, K6 = 0 for T $ T" T" ; Nominal full power T vg, T's 593.0 F (or 590. O F for alternate operation with reduced RCS flow) i f (A1) = 0 for all(AI) 2 2.2.2 The loop design flow for Reactor Coolant Flow-lmw is 95,400 gpm (or 92,500 gpm for attemate operation with reduced RCS flow).
2.3 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3):
2.1.1 The BOL, ARO,_ MTC shall be less positive than the limits shown in ;
Figure 3.
d 2.1.2 The EOL, ARO, HFP, MTC shall be less negative than -6.12 x 10 Ak/k/ F.
d 2.1.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -5.22 x 10 Ak/k/ F (300 ppm Surveillance Limit).
where: BOL stands for Beginning-of-Cycle Life EOL stands for End-of-Cycle Life ARO stands for All Rods Out HFP stands for Hot Full Power (100% RATED THERMAL POWER)
HFP vessel average temperature is 589 F Core Operating Limits Report Page 3 of 12 5/10/99
I i
l South Texas Project Unit t Cycle 9 J
l 2.4 ROD INSERTION LIMITS (Specification 3.1.3.5 and 3.1.3.6):
2.2.1 All banks shall have the same Full Out Position (FOP) of at least 250 steps withdrawn but not exceeding 259 steps withdrawn.
2.2.2 The Control Banks shall be limited in physical insertion as specified in Figure 4.
2.2.3 - Individual Shutdown bank rods are fully withdrawn when the Bank Demand
- Indication is at the FOP and the Rod Group Height Limiting Condition for Operation is satisfied (T.S. 3.13.1).
2.5 AXIAL FLUX DIFFERENCE (Specification 3.2.1):
2.5.1 AFD limits as requimd by Technical Specification 3.2.1 are determined by j
CAOC Operations with an AFD target band of +3, -12%.
2.5.2 The AFD shall be maintained within the ACCEPTABLE OPERATION
. portion of Figum 5, as required by Technical Specifications.
)
2.6
' HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):
i 2.6.1 Fo"" = 2.55.
2.6.2 K(Z)is provided in Figure 6.
2.6.3 The Fxy limits for RATED THERMAL POWER (Fxy"") within specific com planes shall be:
2.6.3.1 Less than or equal to 2.102 for all core planes containing Bank "D" control rods, and
.2.6.3.2 Less than or equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.
2.6.3.3 PFxy = 0.2.
These Fxy limits were used to confirm that the heat fiux hot channel factor Fn(Z) will be limited by Technical Specification 3.2.2 assuming the most-limiting axial-power distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying-variations in the axial xenon and power distributions, as described in WCAP-8385. Therefore, these Fxy limits pmvide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 10 CFR 50.46.
For Unit.1 Cycle 9, the L(Z) penalty is not applied (i.e., L(Z) = 1.0 for all core elevations).
Core Operating Limits Report Page 4 of 12 5/10/99 Lu
South Texas Project Unit 1 Cycle 9 l
2.7 ENTHALPY RISE HOT CII ANNEL FACTOR (Specification 3.2.3) l Standard Fuel VANTAGE SH / VANTAGE + / RFA Fuel 2.7.1 FmRW = g49 p RW = g $$7 2.7.2 PFm = 0.3 PFm = 0.3 2.8 DNH PARAMETERES (Specification 3.2.5) 2.8.1 The following DNB-related parameters shall be maintained within the following limits:
- a. Reactor Coolant System T,yg, s 595 F (or s 593 F with reduced RCS flow of Technical Specification 3.2.5.c)
- b. Pressurizer Pressure, > 2219 psig*
- c. Minimum Measured Reactor Coolant System Flow 2 392,300 gpm** (or 2 380,500gpm** with reduced RCS T.vg of Technical Specificaiton 3.2.5.a)
Limit not applicable during either a Thermal Power ramp in excess of 5% of l
RTP per minute or a Thermal Power step in excess of 10% RTP.
- Includes a 2.8% flow measurement uncertainty.
3.0 REFERENCES
l 3.1 Letter from R. A. Wiley (Westinghouse) to Dave Hoppes (STPNOC), " Unit 1 Cycle 9 Core Operating Limits Report,"
99TG-G-0041 (ST-UB-NOC-1919),
April 1999.
3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. I and 2.
i i
i 1
Core Operating Limits Report Page 5 of 12 5/10/99 k.
SouthTexas Project Unit I Cycle 9 Mgumi Reactor Com Safety Limit-Four Loops in OperaGon l li l I J
J l l 1
4 l
l l
i l.
l pmsessq l f
UNACCE!PTABLE i
Ai a
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540 0-0.2 0.4 0,6 0.8 1
1.2 FRACTION OF RATED THERMALPOWER i
h @ating Limits Report Page 6 of 12 5/10/99
i SouthTexas Project Unit 1 Cycle 9 I4 gum 2 Reactor Com Safety Umit -Four Imops in Operation (Alternate) 680-
)
i noine x s-tuw.< ouw 660 %N N5 Lo a
v NN%
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a FracdonofRatedThermalPower e %dng Wu D Page 7 of 12 5/10/99 m
South Texas Project Unit I Cycle 9 Mgum 3 MTCversus Power Level 7.0 UNACCEPTABLEOPERATION 6.0 l ll 5.0
~
4,o 3.0 2.0 h
ACCEPTABLE OPERATION 1.0 0.0
.-1.0
-2.0
-3.00.0-20.0 40.0 60.0 80.0 100.0 Relative Power (%)
Core Operating Limits Report Page 8 of 12 5/10/99 l
E
South Texas Project Unit 1 Cycle 9 Figure 4 Contrel Rod Insertion Limits
- versus Power level pusg:tumapovens, ps.tsg:122sw o%
puss tatmapoveden ps.tsg: tat me,o %
ausettomapoveder pu,tsg:tto a.,o asp 260 puse 17sm,ovedae (rr.tset17a.,o,ede, pasa:ttsmapovede, e -
pe,tsg:1ts espo,ede, e
(ts,tstitsmenovede, s
pu,ese: tis a,o,ed.,
240 l
e i
' BankB
/
r g
j r
200
- (n, sag s
/
(1M174 :
qg
)
A' 160 f
/
BankC s
140
/
e 120 i
s 100
/
h s
60
/
/
BankD s
60 : (teg
/
40.
s l
0 Cts,0) 0 0
10 20 30 40 50 60 70 80 90 100 RELATIVEPOWBt(IQ
. Control Bank A is aheady withdrawn to Full Out Position. Fully withdrawn region shall be the condition where shutdown and contzul banks are at a position within the interval of 2: 250 and :s 259 steps withdrawn, inclusive.
Core Operating Limits Report Page 9 of12 5/10/99 L
Sosth Texas Project Unit 1 Cycle 9 Mgure 5
' AFD Limits versus Rated 'Ihermal Power 9
120 100 1-11,1i01 111,901 unAccePunsa
/
\\
oncezmsa OPERATION OPERATION
(
\\
~
~
80
/
\\
u
~
8
/
\\
~
g 3
) 60
/
\\
/
T r
ACCEPTABLE I,
Y OPERATION MiM
' '31,501 j
40
~
20 1
I I
g
-50
-40
-30
-20
-10 0
10 20 30 40 50 Flux Difference (AI) %
Crie Operating Limits Report Page10of 12 5/1099
South Texas Project Unit ! Cycic9 ligure 6 K(Z) - Normali=d Fo(Z) versus Core Height 1.2 E5 x
~
af hc.s 1
e OA Core Elevation FQ' K(Z) 0.0ft 2.55 1.0 S 'N 7.0 ft -
2.55 1.0 g
14.0 ft 2.359 0.925 g
@o.2 e
14 8 0.0 7A COREHSGHT(FT)
I Core Operating Limits Report Page!Iof12 5/10/99 L
o South Texas Project Unit i Cycle 9 Table 1 i
Unrodded Fxy for Each Core Height
- i Core Height Unrodded Core Height Umudded Gt.)
Fxy Wt.)
Fxy 14.000 3.4%
6.800 1.887 13.800 3.158 6.600 1.894 13.600 2.812 6.400 1.907 13.400 2.461 6.200 1.915 13.200 2.191 6.000 1.920 13.000 2.021 5.800 1.926 12.800.
1.983 5.600 1.935 12.600 1.%7 5.400 1.946 12.400 1.968 5.200 1.956 12.200 1.965 5.000 1.%3 12.000 1.%9 4.800 1.965 11.800 1.977 4.600 1.964 11.600 1.992 4.400 1.962 11.400 2.011 4.200 1.%7 11.100 2.033 4.000 1.978 11.000 2.044 3.800 1.961 10.800 2.051 3.600 1.947 10.600 2.055 3.400 1.924 10.400 2.054 3.200 1.907 10.200 2.046 3.000 1.895 10.000 2.033 2.800 1.884 9.800 2.017 2.600 1.875 9.600 2.001 2.400 1.862 9.400 1.987 2.200 1.848 9.200 1.976 2.000 1.832 9.000 1.969 1.800 1.817 8.800 1.964 1.600 1.803 8.600 1.960 1.400 1.810 8.400 1.934 1.200 1.813 8.200 1.922 1.000 1.827 8.000 1.909 0.800 1.958 7.800 1.901 0.600 2.195 7.600 1.893 0.400 2.487 7.400 1.887 0.200 2.777 7.200 1.884 0.000 3.066 7.000 1.884 1
l For Unit 1 Cycle 9. the 1(Z) penalty is not applied (i.e 1/Z) 1.0 for all core elevations).
j Core Operating Limits Repw.
Page120f12 5/10/99 1
i e
NOC-AE-000471 Page 1 of 14 ATTACHMENT 6 UNIT 2 CYCLE 7
]
CORE OPERATING LIMITS REPORT
-i (TYPICAL) i l
l i
j l
(
SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNIT 2 CYCLE 7 CORE OPERATING LIMITS REPORT (TYPICAL) e
South Texas Project Unit 2 Cycle 7 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 2 Cycle 7 has been prepared in accordance with the requimments of Technical Specification 6.9.1.6. The core operating
~
limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.6.
The Technical Specifications affected by this report are:
.1) 2.1 SAFETY LIMITS' 2).
2.2 LIMITING SAFETT SYSTEM SETTINGS 3) 3/4.1.1.3_
MODERATOR TEMPERATURE COEFFICIENT LIMITS 4) 3/4.1.3.5
-SHUTDOWN ROD INSERTION LIMITS 5) 3/4.1.3.6.
. CONTROL ROD INSERTION LIMITS 6) 3/4.2.1 AFD LIMITS 7) 3/4.2.2 HEAT FLUX HOT CHANNELFACTOR 8) 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 9) 3/4.2.5 DNB PARAMETERS i
2.0 OPERATING LIMITS I
The cycle-specific parameter hmits for the specifications listed in Section 1.0 are presented below.
2.1 SAFETY LIM'ITS (Specification 2.1) 2.1.1 The combination of THERMAL POWER, pressudzer pmssure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 1, or in Figure 2 when operating under alternate operating cdteria consistent with reduced Reactor Coolant System flow as addressed in Technical Specification 3.2.5.
2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2) i 2.2.1 The Over Temperature AT and Over Pressure AT setpoint parameter values am listed below:
f; l'
^
[
Core Operating Limits Report Page 1 of 12 5/26/99 l.
k:.[
FT F
South Texts Project Unit 2 Cycle 7
)
Overtemperature AT Setooint Par===ter Values j
. T ; measured reactor vessel AT lead / lag time constant, t 2 8 sec I
i t
T ; measured reactor vessel AT lead / lag time constant, T 5 3 see 2
2 T3; measured reactor vessel AT lag time constant, T s 0 see 3
T4; measured reactor vessel average temperature lead / lag time constant, t42 28 sec T ; measured reactor vessel average temperature lead / lag time constant, T s 4 see 5
5 t ; measured reactor vessel average temperature lag time constant, T s 0 sec 6
6 K ;Overtemperature AT tractor trip setpoint, K s 1.14'; or, K s 1.13 for alternate i
i i
operation with reduced RCS flow i
K2;Overtemperature AT reactor trip setpoint T,v, coefficient, K 2 0.028/ F 2
l K3 ;Overtemperature AT reactor trip setpoint piessure coefficient K3 2 0.00143/psig l
T' ; Nominal full power T.vg, T's 593.0 F (or 590. 0 *F for attemate operation with reduced RCS flow) a
' P' ; Nominal pressurizer pressure, P' 2 2235psig I
f (AI) = is a function of the indicated difference between top and bottom detectors i
of the power-range neutron ion chambers; with gains to be selected based.
on measured instrument response during plant startup tests such that;
'(1) For gi - qu between -70% and +8%, or +6% for alternate operation with reduced RCS flow, fi(AI) = 0, where qi and qu are percent
]
RATED THERMAL POWER in the top and bottom halves of the 4
core respectively, and gi + qu is total THERMAL POWER in percent
)
of RATED THERMAL POWER;-
J (2) ; For each percent that the magnitude of qi-qu exceeds -70%, the AT Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THF.RMAL POWER; and (3) For each percent that the magnitude of gi-quexceeds +8%, or +6% for n
attemate operation with reduced RCS flow, the AT Trip Setpoint shall be automatically reduced by 2.65% ofits value at RATED THERMAL POWER.
Core Operating Limits Report Page 2 of 12
$/26/99 U
South Texas Project Unit 2 Cycle 7 Overpower AT Setooint Parameter Values T ; measured reactor vessel AT lead / lag time constant, T 2 8 see i
i T ; measured reactor vessel AT lead / lag time constant, T s 3 sec 2
2 T ; measured reactor vessel AT lag time constant, T = 0 see 3
3 T6; measured reactor vessel average temperature lag time constant, T6 = 0 see T ; Time constant utilized in the rate-lag compensator for T vg, T7 a: 10 sec 7
L ; Overpower AT reactor trip setpoint, Ls 1.08; (or L s 1.07 for alternate operation with reduced RCS flow)
K ; Overpower AT mactor trip setpoint T.vg rate / lag coefficient, K 2 0.02/ F for 5
5 increasing average temperatum, and K 2 0 for decreasing average temperature 5
& ; Overpower AT reactor trip setpoint T vg heatup coefficient k 2 0.002/ F for T
> T" and, & = 0 for T $ T" T"; Nominal full power T.vs, T's 593.0 F (or 590. O F for alternate operation with reduced RCS flow) f2(AI) = 0 for all (AI) 2.2.2 The Loop design flow for Reactor Coolant Flow-Low is 95,400 gpm (or 92,500 gpm for alternate operation with reduced RCS flow).
2.3 MODERATOR TEMPERATURE COEITICIENT (Specification 3.1.1.3):
2.1.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure 3.
q 4
2.1.2 The EOL, ARO, HFP, MTC shall be less negative than -6.12 x 10 Ak/k/ F.
4 2.1.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -5.22 x 10 3
Ak/k/ F (300 ppm Surveillance Limit).
where: BOL stands for Beginning-of-Cycle Life EOL stands for End-of-Cycle Life ARO stands for All Roda Out U
HFP stands for Hot Full Power (100% RATED THERMAL POWER)
HFP vessel average temperature is 589 *F Core Operating Limits Report Page ? of 12 5/26/99
e-i South Texas Project i
Unit 2 Cycle 7 2.4 ROD INSERTION LIMITS (Specification 3.1.3.5 and 3.1.3.6):
2.2.1 All banks shall have the same Full Out Position (FOP) of at least 250 steps withdrawn but not exceeding 259 step:: withdrawn.
2.2.2 The Control Banks shall be limited in physical insertion as specified in Figure 4.
2.2.3 Individual Shutdown bank rods are fully withdrawn when the Bank Demand Indication is at the FOP and the Rod Group Height Limiting Condition for Operation is satisfied (T.S. 3.1.3.1).
2.5 AXIAL FLUX DIFFERENCE (Specification 3.2.1):
2.5.1 AFD limits as required by Technical Specification 3.2.1 are determined by CAOC Operations with an AFD target band of +3,-12%.
2.5.2 The AFD shall be maintained within the ACCEPTABLE OPERATION ponion of Figure 5, as required by Technical Specifications.
2.6 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):
2.6.1 FqRU' = 2.55.
2.6.2 K(Z)is providedin Figure 6.
2.6.3 The Fxy limits for RATED THERMAL POWER (Fxy""') within specific core planes shall be:
2.6.3.1 I2ss than or equal to 2.102 for all cote planes containing Bank "D" control rods, and 2.6.3.2 Less than or equal to the appropriate core height-dependent value from Table I for all unrodded com planes.
2.6.3.3 PFxy = 0.2.
These F,y limits were used to confirm that the heat flux hot channel factor Fo(Z) will be limited by Technical Specification 3.2.2 assuming the most-limiting axial power distributions expected to result for the insenion and removal of Control Banks C and D during operation, including the accompanying variations in the a.tial xenon and power distributions, as described in WCAP-8385. Therefore, these Fxy limits pmvide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 10 CFR 50.46.
For Unit 2 Cycle 7, the I4Z) penalty is not applied (i.e, L(Z) = 1.0 for all core elevations).
Core Operating Limits Report Page 4 of 12 5/26/99 I_ut
South Texas Project Unit 2 Cycle 7 2.7 E_NTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):
Standard Fuel VANTAGE 5H / VANTAGE + / RFA Fuel 2.7.1 Fai"" = 1.49 Fai"* = 1.557 2.7.2 PFai = 0.3 PFai = 0.3 2.8 DNH PARAMETERES (Specification 3.2.5) 2.8.1 The following DNB-related parametens shall be maintained within the following limits:
J
- a. Reactor Coolant System T,vg, s 595 F (or s 593 F with reduced RCS flow of Technical Specification 3.2.5.c)
- b. Pressurizer Pmssure, > 2219 psig*
- c. Minimum Measured Reactor Coolant System Flow 2 392,300 gpm** (or 2 380,500gpm** with reduced RCS T,y, of Technical Specificaiton 3.2.5.a)
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.
1
3.0 REFERENCES
3.1 letter from R. A. Wiley (Westinghouse) to Dave Hoppes (STPNOC), " Unit 1 Cycle 9 Core Operating Limits Repon,"
99TG-G-0041 (ST-UB-NOC-1919),
April 1999.
3.2 NUREG-1346, Technical Specifications, South Texas Pmject Unit Nos. I and 2.
Core Operating Limits Report Page 5 of12 5/26/99
SouthTexas Project Unit 2 Cycle 7 Mgure l-Reactor Core Safety Lhnit - Four Loops in Opemtion 4
e
- g i ; i g (
1
-gg
,j, 680 i
1 I l
l l
UNAOCEPTABLE
+ = of l
i t
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540 0
0.2 0.4 0.6 0.8 1
1.2 FRACTION OF RATEDTHERMALPOWER I
'i 5/26/99 j
cae operating umits Report Page 6 of 12 1
l I
SouthTexas Project Unit 2 Cycle 7 Figum 2 Reactor Com Safety Lhnit-Four Loops in Operation (Altemate) i 1
\\
680 -
claimde)
A s-tu,w ot.wg C' WNhks u
w%
(uw.s wo w
w NX v
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580 g g y, anan
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560,
1.2 0
02 0A 04 OA 1
4 FractionofhatedThenna1 Power l
l l
1 5/2 W Core Wng Umits Report Page 7 of12 1
I
SouthTexas Project i
Unit 2 Cycle 7 Mgure 3 MTC versusPowerlevd 7.0 6.0-l 5.0 4
4,o 3.0.
{d 2.0 h
ACCEPTABLEOPERATION 0.0
.-1.0
-2.0
-3.00.0 20.0 40.0 60.0 80.0 100.0 Relative Power (%)
Core Operating Limits Report Page 8 of 12 5/2 & 99
SouthTexas Project Unit 2 Cycle 7 Mgure 4 Control Rod Insertion Lhnits' versus Power I4 vel fpower.sepse orars an (12,259): 122 step overlap (23,258): 121 step overlap 260 r
(22,256): 119 step overlap
,/
j i
(21,254): 117 step overlap
,e
.r r
r 240 tsawer.voele oversan r
7
/
(79,259): its step overlap r
i isaur s i
,r (79,258): 121 step overlap
/
(78,256): 119 5 top overlap j
220
/
/
(77,254): 117 step overlap
)
f j
200 40,202 l
,/
j
/
\\
\\
l l100,174
}-
180 Y
4 one
)
e 160 j
/
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~
)
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l140 I
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)
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/
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/
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l
)
f
/
)
g 100 r
j
)
/
lhMt1 D I j
f j
s0 r
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. r' ^
0 ss 50 e
r
.r r
40 r
2
/
r 20 l
r
/
129,0 1 0
o 10 to to ao 50 so To 80 90 100 nurmelvs sonst (W Control Bank A is aheady withdrawn to Full Out Position.
Core %tingIJmits Report Page 9 of12 5/2G99
SouthTexas Project Unit 2 Cycic7 Mgure 5 AFD Limihi venus Rated 'Ibennal Power 120
- =
~
100 l-11,901 111,90l'
~"" ~
uNAccePTa nE
/
\\
uxAccEmsuE OPERATION
/
)
OPERATION p
80
/
\\
ul
/
)
i k
/
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T
)g '*
/
\\
[
ACCEPTABLE
\\
.f OPERATION g
F31,50!
!31,50!
40 20 1
0
-50
-40
-30
-20
-10 0
10 20 30 40 50 Flux Difference (AI) %
Core Operating Limits Report Page10of12 5/26/99
l South Texas Project Unit 2 Cycle 7 Figure 6 K(Z)- NomII=d Fo(Z) versus Core Height I
12-E1 er k02 l
f I
e OA i
g K(Z 1.0)
Core Elevation FQ 2.55
@ 'd
~
7.0 ft 2.55 1.0 1
0.0ft 14.0 ft 2.359 0.925 c2 0
72 14D 0.0
. CORE HEIGHT (PT)
Y l
l a
Core Operating Limits Report Page11of12 5/26/99 J
1 i
South Texas Project Unit 2 Cycle 7 Table 1 Unrodded Fry for Each CoreIIelght*
Core Height Unrodded Core Height Unrodded mt.)
-ly__
mt.)
F,y_
14.000 3.443 6.800 1.905 13.800 3.125 6.600 1.907 13.600 2.793 6.400 1.909 13.400 2.453 6.200 1.911 13.200 2.193 6.000 1.909 13.000 2.032 5.800 1.905 12.800 1.987 5.600 1.904 12.600 1.975 5.400 1.907 12.400 1.083 5.200 1.916 12.200 1.985 5.000 1.923 12.000 1.994 4.800 1.929 11.800 2.007 4.600 1.926 11.600 2.029 4.400 1.923 11.400 2.024 4.200 1.924 11.200 2.000 4.000 1.927 11.000 1.949 3.800 1.937 10.800-1.918 3.600 1.951 10.600 1.899 3.400 1.945 10.400 1.892 3.200 1.929 10.200 1.889 3.000 1.908 10.000 1.886 2.800 1.886 9.800 1.883 2.600 1.862 i
9.600 1.878 2.400 1.844 9.400 1.872 2.200 1.831 9.200 1.866 2.000 1.816 9.000 1.861-1.800 1.805 8.800 1.857 1.600 1.788 8.600 1.855 1.400 1.786 8.400 1.854 1.200 1.777 8.200 1.857 1.000 1.779 8.000 1.862 0.800 1.896 7.800 1.872 0.600 2.113 7.600 1.884 0.400 2381 7.400 1.895 0.200 2.643 7.200 1.901 0.000 2.893 7.000 1.903 For Unit 2 Cycle 7, the MZ) penalty is not applied (i.e 14Z) = 1.0 for all core elevations).
Core Ope sting Limits Report Page12of 12 5/2099
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