ML20196F352

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Non-proprietary WCAP-14788, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power)
ML20196F352
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 04/30/1999
From: Ciocca C, Moomau W
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20137U998 List:
References
WCAP-14788, NUDOCS 9906290188
Download: ML20196F352 (67)


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,

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14788 Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power) i April 1999 C. F. Ciocca W.H. Moomau WESTINGHOUSE ELECTRIC COMPANY LLC 4350 Nonhern Pike Monroeville, Pennsylvania 15146-2886 C 1999 Westinghouse Electric Company LLC All Rights Reserved r/

9906290188 990622 PDR ADOCK 05000266 Kh v. P PDR

TABLE OF CONTENTS Section Title Page I.

-Introduction 1

II.

Methodology 3

III.

Instrumentation Uncertainties 7

IV.

Results/ Conclusions 31 References 32 i

LIST OF TABLES Table Number Title Page 1

Pressurizer Pressure Control System Uncertainty 8

2 Tavg Rod Control System Uncertainty 10 3

Flow Calorimetric Instrumentation Uncertainties 16 4

Flow Calorimetric Sensitivities 17 5

Calorimetric RCS Flow Measurement Uncertai..ty 18 6

Loop RCS Flow Uncertainty 21 7

Power Calorimetric Instrumentation Uncertainties 27 8

Power Calorimetric Sensitivities 28 9

Secondary Side Power Calorimetric Measurement Uncertainty 29 ii

LIST OF FIGURES Figure Number

. Title Page 1.

Calorimetric RCS Flow Measurement 33 2

Calorimetric Power Measurement 34 i

iii

This page left intentionally blank.

iv

1 WESTINGHOUSE REVISED THERMAL DESIGN PROCEDURE INSTRUMENT UNCERTAINTY METHODOLOGY s

(FUEL UPGRADE & UPRATE TO 1656 MWT - NSSS POWER) j L

INTRODUCTION i

An upgrade to the fuel product at uprated conditions of 1656 Mwt - NSSS power has been requested by Wisconsin Electric for both Units 1&2 at Point Beach Nuclear Power Station. The fuel product to satisfy the intended requirements is the Westinghouse 14 X 14 PERFORMANCE

+ 422 fuel assembly. To utilize this new fuel assembly at uprated conditions, a new accident analysis will be required in addition to recalculating and revising the Instrument Uncertainty Methodology. This report supersedes "ITDP Instrument Uncertainty Report" dated June 7,1984 (84WE*-G-044).

Four operating parameter uncenainties are used in the uncertainty analysis of the Revised Thermal Design Procedure (RTDP). These parameters are Presrurizer Pressure, Primary Coolant Temperature (Tavs), Reactor Power, and Reactor Coolant System Flow. They are frequently monitored and several are used for control purposes. Reactor power is monitored by the performance of a secondary side heat balance (power calorimetric) at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. RCS flow is monitored by the performance of a calorimetric flow measurement at the beginning of each cycle. The RCS Cold Leg loop flow indicators are evaluated against the calorimetric flow measurement. Pressurizer pressure is a controlled parameter and the uncertainty reflects the control system. T b a controlled parameter via the temperature input to the rod control ayg system, and the uncertainty reflects this control system. The RTDP(l)is used to predict the plant's DNBR design limit. The RTDP methodology considers the uncertainties in the system operating plant parameters, fuel fabrication and nuclear and thermal parameters and includes the use of various DNB correlations. Use of the RTDP methodology requires that variances in the plant operating parameters are justified. The purpose of the following evaluation is to define the

]

specific Point Beach Units 1 & 2 Nuclear Plant instrument uncertainties for the four primary i

system operating parameters which are used to predict the plant safety analysis DNBR design limit via the RTDP, and to determine the starting points of certain plant parameters in some of the accident analyses.

4 Westinghouse has been involved with the development of several techniques to treat instrumentation uncenainties. An early version (for D. C. Cook 2 and Trojan) used the methodology outlined in WCAP-8567 " Improved Thennal Design Procedure", (2,3,4) which is based on the conservative assumption that the uncenainties can be described with uniform probability distributions. Another approach is based on the more realistic assumption that the uncenainties can be described with random, normal, two sided probability distributions.(5)

This approach is used to substantiate the acceptability of the protection system setpoints for many Westinghouse plants, e.g., D. C. Cook 2(6), V. C. Summer, Wolf Creek, Millstone Unit 3 and others. The second approach is now utilized for the determination of all instrumentation uncertainties for the RTDP parameters and protection functions.

I

The determination ofpressure, temperature, power and RCS flow uncertainties are applicable for the Point Beach Plant Units 1 & 2 for power levels up to 1656 Mwt - NSSS power, for 18 month fuel cycles + 25% per the plant Technical Specifications, and for a full power Tavg window of 558.1 to 574.0 F.

2

1 I

H.

METHODOLOGY The methodology used to combine the error components for a channelis the square root of the sum of the squares of those groups of components which are statistically independent. Those errors that are dependent are combined arithmetically into independent groups, which are then

)

systematically combined. The uncertainties used are considered to be random, two sided distributions. The sum of both sides is equal to the range for that parameter, e.g., Rack Drift is typically [

]+a,c, the range for this parameter is [

]+a,c. This technique has been utilized before as noted above, and has been endorsed by the NRC stafy(7,8,9,10) and various industry standards (ll,12),

The relationships between the error components and the channel instrument error allowance are variations of the basic Westinghouse Setpoint Methodology (13) and are defined as follows:

1.

For precision parameter indication using Special Test Equipment or a digital voltmeter (DVM) at the input to the racks; CSA = {(SMTE + SCA)2 + (SPE)2 + (STE)2 + (SMTE + SD)2 + (SRA)2

+ (RDOUT)2}1/2 + BIAS Eq. I 2.

For parameter indication utilizing the plant process computer; CSA = ((SMTE + SCA)2 + (SPE)2 + (STE)2 + (SMTE + SD)2 + (SRA)2

+ (RMTE + RCA)2 + (RTE)2 + (RMTE + RD)2 + (RMTE + A/D)2} 1/2

+ BIAS Eq.2 3.

For parameters with closed-loop automatic control systems, the calculation takes credit for [

]*. There is a functional dependency between the transmitters / racks and the automatic control system / indicator when an uncenainty in the transmitters / racks is common to the automatic control system and the indicator. That is, an uncertainty in the high direction in the transmitter / racks will result in a high uncertainty in the automatic control system and the indicator. To account for the functional dependency, a square root function is used for the transmitter / racks / reference signal, and a square root function is used for the controller / indicators; 3

2 CSA = ((PMA (random) + (PEA)2

+ (SMTE + SCA)2 + (SPE)2 + (STE)2 + (SMTE + SD)2 + (SR

+ (RMTE + RCA)2 + (RTE)2 + (RMTE + RD)2 + (REF)2} 1/2

+ {(CMTE + CA)2 + (RMTE + RCA)2m + (RDOUT)2,}1/2

+ BIAS Eq.3 where:

CSA Channel Statistical Allowance

=

PMA ProcessMeasurement Accuracy

=

PEA PrimaryElement Accuracy

=

SRA Sensor Reference Accuracy

=

SCA Sensor Calibration Accuracy

=

SMTE Sensor Measurement and Test Equipment Accuracy

=

SPE

. Sensor Pressure Effects

=

STE Sensor Temperature Effects

=

SD

+

Sensor Drift

=

RCA Rack Calibration Accuracy

=

RMTE Rack Measurement and Test Equipment Accuracy

=

RTE Rack Temperature Effects

=

RD Rack Drift

=

RDOUT Readout Device Accuracy

=

CA Controller Allowance

=

CMTE Controller Measurement and Test Equipment Acuracy

=

A/D Analog to Digital Conversion Accuracy

=

REF Reference signal for automatic control sytem

=

IND Indicator.

=

PMA and PEA terms are not included in equations 1 and 2 since the equations ar instrumentation uncertainties only. PMA and PEA terms are included in the dete control system uncertainties.

The parameters above are defined in references 5 and 12 and are based on SAM 20.1, 1973(14).

However, for ease in understanding they are paraphrased below:

PMA

- non-instrument related measurement errors, e.g., temperature stratification of a fluid in a pipe.

PEA

- errors due to a metering device, e.g., elbow, venturi, orifice.

SRA

- reference (calibration) accuracy for a sensor / transmitter.

SCA

- calibration tolerance for a sensor / transmitter.

SMTE

- measurement and test equipment used to calibrate a sensor / transmitter.

SPE

- change in input-output relationship due to a change in static pressure 4

fo a differential pressure (d/p) cell.

STE

- change in input-output relationship due to a change in ambient temperature for a sensor or transmitter.

SD

- change in input-output relationship over a period of time at reference conditions for a sensor or transmitter.

RCA

- calibration accuracy for all rack modules in loop or channel assuming the loop or channel is string calibrated, or tuned, to this accuracy.

RMTE - measurement and test equipment used to calibrate rack modules.

)

RTE

- change in input-output relationship due to a change in ambient temperature for the rack modules.

RD

- change in input-output relationship over a period of time at reference conditions for the rack modules.

RDOUT - the measurement accuracy of a special test local gauge, digital voltmeter or multimeter on it's most accurate applicable range for the parameter measured, or 1/2 the smallest increment on an indicator (readability).

CA

- allowance of the controller rack module (s) that performs the comparison and calculates the difference between the controlled parameter and the reference signal.

CMTE - measurement and test equipment used to calibrate the controller rack

)

module (s) that perform (s) the comparison between the controlled parameter I

and the reference signal.

A/D

- allowance for conversion accuracy of an analog signal to a digital signal for process computer use.

REF

- the reference signal uncertainty for a closed-loop automatic control system.

IND

- indicator accuracies are used for these uncertainty calculations. Control board indicators are typically used.

BIAS

- a one directional uncertainty for a sensor / transmitter or a process parameter with a known magnitude.

A more detailed explanation of the Westinghouse methodology noting the interaction of several parameters is provided in references 6 and 13.

5

k This page is left intentionally blank.

6

III.

INSTRUMENTATION UNCERTAINTIES The instrumentation uncertainties will be discussed first for the two parameters which are controlled by automatic systems, Pressurizer Pressure, and Tavg (through automatic rod control).

1.

PRESSURI7FR PRESSURE Pressurizer pressure is controlled by a closed-loop automatic control system that compares the measured vapor space pressure to a reference value. This uncertainty calculation takes credit for the closed-loop control system design where [

]

The control channel uncertainties for the automatic control system include allowances for the pressure transmitters, the process racks, and the control system reference setpoint. The pressurizer pressure control system reference setpoint is generated by the setting of a variable potentiometer on the Main Control Board manual / automatic station. The reference setpoint (Pref) is adjusted

)

and verified by the plant operators with the control board indicators. This uncertainty calculation also includes the control board indicators for veri 6 cation of the automatic control system performance.

)

On Table 1, the electronics uncertainty for this function is [

] with a [

]**'

bias corresponding to [

]*** with a [

]*** bias for the average of 3 control board indicators. In addition to the control system uncertainty, an allowance is made for pressure overshoot or undershoot due to the interaction and thermal inertia of the heaters and spray, An allowance of[

]# is made for this effect. The total control system uncertainty including indication is [ _

]*** with a [

] bias which results in a standard deviation of

[

] for a normal, two sided probability distribution.

1 7

TABLE 1 PRESSURIZER PRESSURE CONTROL SYSTEM UNCERTAINTY All Valuesin % Span

- +a,C REF

=

PMA

=

PEA

=

SRA

=

SCA

=

SMTE

=

STE

=

SD

=

BIAS

=

RCA

=

RMTE

=

RTE

{

=

RD

=

RCA

=

g RMTE

=

g RDOUT

=

g CA

=

CMTE

=

RANGE = 1700 - 2500 psig, SPAN = 800 psi CHANNELS P-429, -430, -431 & -449 ELECTRONICS UNCERTAIhTY =

+a.c

{

PLUS t

ELECTRONICS UNCERTAINTY =

PLUS i

CONTROLLER UNCERTAINTY

=

  • 15 psi setting tolerance around 2235 psig 8

l l

2.

IaYg T,yg is controlled by a system that compares the high T,yg rom the loops with a reference f

derived from the First Stage Tuitine Impulse Chamber Pressure. T,yg is the average of the narrow range T and Tc alues. The high loop T,yg s then used for rod control. Allowances H

v i

l are made (as noted on Table 2) for the RTDs, transmitter and the process racks / indicators and

)

controller. The CSA for this function is dependent on the type of RTD, pressure transmitter, and the location of the RTDs, i.e., in the Hot and Cold Leg bypass manifolds. Based on one Tu and one Tc RTD per channel to calculate Tavg and with the RTDs located in the hot and cold leg bypass manifolds, the CSA for the electronics is [

]+a c. Assuming a normal, two sided probability distribution results in an electronics standard deviation (si) of[

]+a,c.

However, this does not include the deadband of 1.5 F for automatic control. The T ava controller accuracy is the combination of the instrumentation accuracy and the deadband. The probability distribution for the deadband has been determined to be [

{

s

).+a,c The variance for the deadband uncertainty is then:

)

(s2)2=[

]+a.c - [

)+a c i

where [

]+a,c, Combining the variance for instrumentation and deadband results in a controller variance of:

i (st)2-(sg)2+(5)2-[

]+a.c 2

The controller sT = [

]+a.C for a total random uncertainty of[

]+a,c, An additional bias of[

]+a,e for Teoig streaming (in terms of Tavg) based on a conservative

[

]+a.cTcoia treaming uncertainty is included in Table 2. An additional bias of s

[

]+a.c for R/E linearization (in terms of Tavg) is ir.cluded in Table 2. Therefore, the total uncertainty of the controller with the additional biases is [

]+* c random and

[

]+a.c bias.

i 9

TABLE 2 TAVG ROD CONTROL SYSTEM UNCERTAINTY

% T,yg Span

% Turbine Pressure Span (P-485,-486)

+a.c PMAg

=

PMA2

=

SRA

=

SCA

=

SMTE

=

STE

=

SD

=

BIAS:

=

BIAS 2

=

R/E

=

R/E_MTE

=

RCA

=

RMTE

=

RTE

=

RD

=

RCA i

=

m RMTE

=

m RDOUT

=

m CA

=

CMTE

=

  1. Hot Leg RTDs = 1/ Channel
  1. Cold Leg RTDs = 1/ Channel Tavg span 100 "F (520-620 F)

=

R/E span 150 F (500-650 F for Hot and Cold Leg)

=

Turbine pressure span =

650 psi (0-650 psig)

~~

~

+a.C ELECTRONICS CSA

=

ELECTRONICS SIGMA

=

CONTROLLER SIGMA

=

CONTROLLER CSA

=

CONTROLLER BIAS

=

Note A: Module TM-401BB = 0.5% span-Note C: Mo'dule TM-401EE = 0.5% span

~

Module TM-401D = 0.5% span l

Module TM-40lH = 0.5% span j

Note B: Module PM-485A = 0.5% span Module TM-40lM = 0.25% span Module TM-40lP = 0.82% span Module TM-40lN = 1.63% span

}

Modules for Loop Al similar for A2, B1 & B2.

Module TM-40ll = 0.5% span I

10 L

3.

- RCS FLOW Calorimetric RCS Flow Measurement Uncertainty (Using LEFM on the Feedwater Header)

RTDP and Point Beach's Technical Specifications require an RCS flow measurement with a high degree of accuracy. A total RCS flow measurement every fuel cycle,18 months, is performed to verify RCS flow and to normalize the RCS flow instrument channels. Interim surveillances performed with the process computer ensure that the RCS flow is maintained within the assumed safety analysis values, i.e., Minimum Measured Flow (MMF). The 18 month RCS flow surveillance is satisfied by a secondary side power-based calorimetric RCS flow measurement.

l The calorimetric flow measurement is performed at the beginning of a cycle near full power operation.

Eighteen month instrument drift is used in this uncertainty analysis for hot and cold leg RTDs, and for feedwater pressure, steam pressure and pressurizer pressure transmitters.

l A Leading Edge Flow Meter (LEFM) installed on the Feedwater header is used to determine total Feedwater flow. Feedwater temperature indication by the LEFM is compared to individual loop feedwater temperatures which are then adjusted if necessary.

The flow measurement is performed by determining the Steam Generator thermal output (corrected for the RCP heat input and the loop's share of primary system heat losses) and the enthalpy rise (Delta-h) of the primary coolant. Assuming that the primary and secondary sides are

)

in equilibrium, the RCS total vessel flow is the sum of the individual primary loop flows, i.e.,

WRCS " Ti 1 (W )i.

Eq.4 L

The individual primary loop volumetric flows are determined by correcting the thermal output of the Steam Generator for Steam Generator blowdown (if not secured), subtracting the RCP heat addition, adding the loop's share of the primary side system losses, dividing by the primary side enthalpy rise and multiplying by the Cold Leg specific volume. The equation for this calculation is:

W = (AMOsa. _QP + (OJNM(Vc) t (hn - he)

Eq. 5 where; W

Loop Flow (gpm)

=

i 3

A Constant conversion factor 0.1247 gpm/(ft /hr)

=

Q Steam Generator therral output (BTU /hr)

=

SG Q

RCP heat addition (BTU /hr)

=

p Q

Primary system net heat losses (BTU /hr)

=

11

V S ecific volume of the Cold Leg at TC(

}

P C

h.

Number of primary side loops N

=

Hot Leg enthalpy (BTUMb)

=

H h

Cold Leg enthalpy NTU/lb).

=

C The thermal output of the Steam Generator is determined by a secondary side calorimetric measurement, which is defined as:

Qso = (h, - hr)Wr Eq.6 where;.

h, Steam enthalpy (BTU /lb)

=

h Feedwater enthalpy (BTU /lb)

=

p Feedwater flow (LEFM feedwater header flow divided by W

=

7

  1. loops)(lb/hr).

The Steam enthalpy is based on the measurement of Steam Generator outlet Steam pressure assuming saturated conditions. The Feedwater enthalpy is based on the measurement of Feedwater temperature and nominal Feedwater pressure. The Feedwater flow is determined by LEFM measurements.

(

RCP heat addition is determined by calculation, based on the best estimate of coolant flow, pump head, and pump hydraulic efficiency.

The primary system net heat losses are determined by calculation, considering the following system heat inputs (+) and heat losses (-):

~ Charging flow (+)

Letdown flow (-)

Sealinjection flow (+)

RCP thermal barrier cooler heat removal (-)

Pressurizer spray flow (-)

. Pressurizer surge line flow (+)

Component insulation heat losses (-)

Component support heat losses (-)

CRDM heat losses (-).

A single calculated sum for 100% RTP operation is used for these losses or heat inputs.

The Hot Leg and Cold Leg enthalpies are based on the measurement of the Hot Leg temperature, Cold Leg temperature and the nominal Pressurizer pressure. The Cold Leg specific volume is based on measurement of the Cold Leg temperature and nominal Pressurizer pressure.

12

The RCS flow measurement is thus based on the following plant measurements:

Steamline pressure (P )

s Feedwater temperature (Tr)

Feedwater pressure (Pr)

Feedwater flow from LEFM Hot Leg temperature (T )

H Cold Leg temperature (T )

C Pressurizer pressure (P )

p Steam Generator blowdown flow (if not secured) and on the following calculated values:

Feedwater density (pr)

Feedwater enthalpy (hr)

Steam enthalpy (h )

s Moisture carryover (impacts h )

s Primary system net heat losses (QL)

RCP heat addition (Qp) l Hot Leg enthalpy (h )

H l

Cold Leg enthalpy (h )-

C These measurements and calculations are presented schematically in Figure 1. The derivation of the measurement and flow uncertainties on Table 5 are noted below.

Secondary Side The secondary side uncertainties are in four principal areas, Feedwater flow, Feedwater enthalpy, Steam enthalpy and net pump heat addition. These areas are specifically identified on Table 5.

For the measurement of Feedwater flow, the LEFM is located on the feedwater header and provides a total flow. The accuracy to which the total flow is determined is based on calculations performed by the manufacture of the LEFM.

Using the NBS/NRC Steam Tables it is possible to determine the sensitivities of various parameters to changes in Feedwater te.nperature and pressure. Table 3 notes the instrument i

uncertainties for the hardware used to perform the measurements. Table 4 lists the various sensitivities. As can be seen on Table 5, Feedwater temperature uncertainties have an impact on i

Feedwater density and Feedwater enthalpy. Feedwater pressure uncertainties impact Feedwater density and Feedwater enthalpy.

l 13

Using the NBS/NRC Steam Tables, it is possible to determine the sensitivity of Steam enthalpy to changes in Steam pressure and Steam quality. Table 3 notes the uncertainty in Steam pressure and Table 4 provides the sensitivity. For Steam quality, the Steam Tables were used to determine the sensitivity at a moisture content of[

]+a.c. This value is noted on Table 4.

The net pump heat addition uncertainty is derived from the combination of the primary system net heat losses and pump heat addition and are summarized for a two loop plant as follows:

System heat losses

- 2.0 MWt Component conduction and convection losses

- 1.4 MWt Pump heat adder

+ 9.4_MWt i

Net Heat input to RCS

+ 6.0 MWt The uncertainty on system heat losses, which is essentially all due to charging and letdown flows, has been estimated to be [

]+a.c of the calculated value. Since direct measurements are not possible, the uncertainty on component conduction and convection losses has been assumed to be

[

]+a.c of the calculated value. Reactor coolant pump hydraulics are known to a relatively high confidence level, supported by system hydraulics tests performed at Prairie Island Unit 2 and by input power measurements from several other plants. Therefore, the uncertainty for the pump heat addition is estimated to be [

]+a.c f the best estimate value. Considering these o

parameters as one quantity, which is designated the net pump heat addition uncenainty, the combined uncertainties are less than [

]+a.c of the total, which is [

]+a.c ofcore power.

Primary Side l

l The primary side uncertainties are in three principal areas, hot leg enthalpy, cold leg enthalpy and I

cold leg specific volume. These are specifically noted on Table 5. Three primary side parameters l

are actually measured, Tg, Tc and Pressurizer pressure. Hot Leg enthalpy is influenced by T,H Pressurizer pressure and Hot Leg temperature streaming. The uncertainties for the instmmentation are noted on Table 3 and the sensitivities are provided on Table 4. The hot leg l

streaming is split into random and systematic components. For Point Beach Units 1 & 2 where I

the RTDs are located in bypass manifolds, the hot leg temperature streaming uncertainty components are [

]+a.c random and [

]+a.c ystematic.

s The cold leg enthalpy and specific volume uncertainties are impacted by T and Pressurizer C

pressure. Table 3 notes the T instrument uncertainty and Table 4 provides the sensitivities.

C l

l Parameter dependent effects are identified on Table 5. Westinghouse has determined the l

dependent sets in the calculation and the direction ofinteraction, i.e., whether components in a dependent set are additive or subtractive with respect to a conservative calculation of RCS flow.

l The same work was performed for the instrument bias values. As a result, the calculation 1

14

explicitly accounts for dependent effects and biases with credit taken for sign (or direction of impact).

I Using Table 5, the 2 loop uncertainty equation (with biases) is as follows:

+a.c i

)

l l

l l

. Based on the number ofloops; number, type and measurement method of RTDs, and the vessel Delta-T, the flow uncertaintyis:

  1. ofloops flow uncertainty (% flow) 2

[

]+a,c I

l l

i i

15 L.

TABLE 3 FLOW CALORIMETRIC INSTRUhENTATION UNCERTAINTIES

% SPAN FW TEMP FW PRES FW FLOW STM PRESS TH TC PRZ PRESS

+a.c LEFM

=

SRA

=

SCA

=

SMTE

=

SPE

=

STE

=

SD

=

BIAS

=

R/E

=

RMTE

=

RTE j

=

RD

=

b RDOUT

=

f CSA

=

  1. OF INSTRUMENTS USED 1/ Loop 1/ Loop 1

1/ Loop 2/ Loop 2/ Loop 4

  • F psi

% Flow psi

'F

  • F psi INST SPAN =

(1) 1500W 100W 1400W 150 )

150 )

800*

0 0

+a.c INST UNC.

(RANDOM) =

INST UNC.

(BIAS)

=

800-700-590.2-526.0-NOMINAL =

440.7 'F 875 psia 100 % Flow 775 psia 605.5 'F 542.5 'F 2250 psia (1) Special test instamentation TE-3111 and an Resistance Thermometer bridge are used for this measurement.

(2) Pressure (P-2245) is measured with a digital voltmeter at the input of the process instrumentation.

(3) Flow (F-3110) is measured with an LEFM on the feedwater header.

(4) Pressure (P-468, -469, -478, -479, -482 -483) is measured with a digital voltmeter at the input of the process instrumentation.

(5) Temperature is measured with a digital voltmeter at the output of the R/E process instrumentation modules.

(6) Pressure (P-429, -430 -431, -449) is measured with a digital voltmeter at the input of the process instmmentation.

I6

l TABLE 4 FLOW CALORIMETRIC SENSITIVITIES FEEDWATER FLOW

+a c LEFM-

=

DENSITY TEMPERATURE

=

PRESSURE

=

FEEDWATER ENTHALPY TEMPERATURE

=

PRESSURE

=

+

h,

=

h

=

g Dh (SG)

=

STEAM ENTHALPY PRESSURE

=

MOISTURE

=

HOT LEG ENTHALPY TEMPERATURE

=

PRESSURE

=

h

=

g h

c Dh (VESS)

=

COLD LEG ENTHALPY TEMPERATURE

=

PRESSURE

=

COLD LEG SPECIFIC VOLUME TEMPERATURE

=

PRESSURE

=

17 I.

TABLE 5 CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTY COMPONENT INSTRUMENT UNCERTAINTY FLOW UNCERTAINTY

+3.C FEEDWATER FLOW LEFM DENSITY TEMPERATURE PRESSURE FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION HOT LEG ENTHALPY TEMPERATURE STREAMING, RANDOM STREAMING, SYSTEMATIC PRESSURE COLD LEG ENTHALPY TEMPERATURE PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE

  • " +, ++ INDICi.TES SETS OF DEPENDENT PARAMETERS l

18

l 1

l l

TABLE 5 (CONTINUED) l CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTY COMPONENT FLOW UNCERTAINTY

+a.c BIAS VALUES FEEDWATER PRESSURE DENSITY ENTHALPY STEAM PRESSURE ENTHALPY PRESSURIZER PRESSURE ENTHALPY - HOT LEG ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG COLD LEG ENTHALPY TEMPERATURE COLD LEG SPECIFIC VOLUME TEMPERATURE FLOW BIAS TOTAL VALUE

+3,C 2 LOOP UNCERTAINTY (WITHOUT BIAS VALUES) 2 LOOP UNCERTAINTY (WITH BIAS VALUES) 19

{

i l

i

Loop RCS Flow Uncertainty (Using Plant Computer)

As noted earlier, the calorimetric RCS flow measurement is used as the reference for normalizing the loop RCS flow measurement from the cold leg elbow tap transmitters. Since the cold leg elbow tap transmitters feed the plant computer, it is a simple matter to perform an RCS flow surveillance. Table 6 notes the instrument uncertainties for determining flow by using the loop RCS flow channels and the plant computer, assuming one loop RCS flow channel per reactor coolant loop. The d/p transmitter uncertainties are converted to percent flow using the following conversion factor:

% flow = (d/p uncenainty)(1/2)(FLOWmax / FLOWnominal)2 where FLOWmax is the maximum value of the loop RCS flow channel. The loop.RCS flow uncertainty is then combined with the calorimetric RCS flow measurement uncertainty. This combination of uncertainties results in the following total flow uncertainty:

  1. ofloops flow uncertainty ( % flow )

2 1 2.3 The corresponding value used in RTDP is:

  1. ofloops standard deviation ( % flow )

+a,c 2

20

[

l TABLE 6 LOOP RCS FLOW UNCERTAINTY PLANT COMPUTER l

INSTRUMENT UNCERTAINTIES 1 LOOP RCS FLOW CHANNEL PER REACTOR COOLANT LOOP (F-411, -412, -413, -414, -415, -416)

% d/p SPAN

% Flow

+a.c PMA

=

PEA

=

SRA

=

SCA

=

SMTE

=

SPE

=

STE

=

SD

=

4 BIAS

=

RCA

=

RMTE

=

RTE

=

RD

=

A/D

=

A/D_MTE=

FLOW CALORIMETRIC BIAS

=

FLOW CALORIMETRIC

=

INSTRUMENT SPAN

=

SINGLE LOOP ELBOW TAP FLOW UNCERTAINTY

+a.c

=

2 LOOP RCS FLOW UNCERTAINTY (WITHOUT BIAS VALUES)

=

2 LOOP RCS FLOW UNCERTAINTY (WITH BIAS VALUES) 2.3

=

Note A: Module FM-411, -412, -413, -414, -415, -416 = 0.5% span

  • Zero values due to normalization to calorimetric RCS flow measurement l

21 l

l

This page left intentionally blank.

22

4.

REACTOR POWER The plant performs a primary / secondary side heat balance at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when power is above 15% Rated Thermal Power. This heat balance is used to verify that the plant is operating within the limits of the Operating License and to adjust the Power Range Neutron Flux channels when the difference between the Power Range Neutron Flux channels and the heat balance is greater than required by the plant Technical Specifications.

Assuming that the primary and secondary sides are in equilibrium; the core power is determined by summing the thermal output of the steam generators, correcting the total secondary power for Steam Generator blowdown (if not secured), subtracting the RCP heat addition, adding the primary side system losses, and dividing by the core Btu /hr at rated full power. The equation for this calculation is:

RP = L_N{Qso Qe + (Oi/N)l l(100)

Eq.7 i

H i

where; RP

= Core power ( % RTP )

N

= Number of primary side loops QSG = Steam generator thermal output (BTU / hr ) as defined in Eq. 6 Qp

= RCP heat addition (BTU / hr ) as defined in Eq. 5 Qt

= Primary system net heat losses (BTU / hr ) as defined in Eq. 5 H

= Rated core power (BTU / hr).

For the purposes of this uncertainty analysis (and based on H noted above) it is assumed that the plant is at 100% RTP when the measurement is taken. Measurements performed at lower power levels will result in different uncertainty values. However, operation at lower power levels results in increased margin to DNB far in excess of any margin losses due to increased measurement uncertainty.

The feedwater flow in equation 6 is determined by multiple measurements and the following calculation:

Wr= (K)(FJ{(p r)(d/p)}

  • Eq. 8 where:

Wr

= Feedwater loop flow (Ib/hr)

F

= Feedwater venturi flow coefficient F,

= Feedwater venturi correction for thermal expansion 3

pr

= Feedwater density (lb/ft )

d/p

= Feedwater venturi pressure drop (inches H O).

2 23 e

The feedwater venturi flow coefficient is the product of a number of constants inc dimensions of the venturi and calibration tests performed by the vendor. The th correction is based on the coefficient of expansion of the venturi material and the differe between feedwater temperature and calibration temperature. Feedwater density is b measurement of feedwater temperature and feedwater pressure. The venturi pressure obtained from the output of the differential pressure transmitter connected to the venturi.

The power measurement is thus based on the following plant measurements:

Steamline pressure (P )

Feedwater temperature (Tr)

Feedwater pressure (Pr)

Feedwater venturi differential pressure (d/p)

Steam generator blowdown (if not secured);

and on the following calculated values:

Feedwater venturi flow coefficients (K)

Feedwater venturi thermal expansion correction (F.)

Feedwater density (pt) l Feedwater enthalpy (hr)

Steam enthalpy (h.)

l Moisture carryover (impacts h.)

l Primary system net heat losses (Qt.)

RCP heat addition (Q,)

Secondary Side The secondary side power calorimetric equations and effects are the same as those note calorimetric RCS flow measurement (secondary side portion), equation 6. The measurem calculations are presented schematically on Figure 2.

For the measurement offeedwater flow, each feedwater venturi is calibrated by the vendo hydraulics laboratory under controlled conditions to an accuracy of[

]***

The calibration data which substantiates this accuracy is provided to the plant by the vendor. A additional uncertainty factor of[

] is included for installation effects, resulting in a conservative overall flow coefficient (K) uncertainty of[

]. Since the calculated steam generator thermal output is proportional to feedwater flow, the flow coefficient u is expressed as [

]**'. It should be noted that no allowance is made for feedwater venturi fouling. The effect of fouling results in an indicated power higher than actual, whic conservative.

24

The uncertainty applied to the feedwater venturi thermal expansion correc' m (F.) is based on the uncertainties of the measured feedwater temperature and the coefficient of thermal expansion for the venturi material,304 stainless steel. For this material, a change of 1.0 'F in the nominal feedwater temperature range changes F. by [

]* and the steam generator thermal output by the same amount.

Based on data introduced into the ASME Code, the uncertainty in F. for 304 stainless steel is 5

%. This results in an additional uncertainty of(

]' in power.

Using the NBSNRC Steam Tables it is possible to determine the sensitivities of various parameters to changes in feedwater temperature and pressure. Table 7 notes the instrument uncertainties for the hardware used to perfcrm the measurements. Table 8 lists the various sensitivities. As can be seen on Table 8, feedwater temperature uncertainties have an impact on venturi F., feedwater density and feedwater enthalpy. Feedwater pressure uncertainties impact feedwater density and feedwater enthalpy.

Feedwater venturi d/p uncertainties are convened to % feedwater flow using the following conversion factor:

% flow = (d/p uncertainty)(1/2)(FLOWmax/FLOWnominal)2,

The feedwater flow transmitter span (FLOWmax) is 120.0% of nominal flow.

Since it is necessary to make this determination daily, the plant computer is used for the

. calorimetric power measurement. As noted in Table 9, Westinghouse has determined the dependent sets in the calculation and the direction ofinteraction. This is the same as that performed for the calorimetric RCS flow measurement, but applicable only to power. The same was performed for the bias values.

Using the power uncenainty values noted on Table 9, the 2 loop uncenainty (with bias values) equation is as follows:

+a.c 25 e

L

' Based on the number ofloops and the instrument uncenainties for the four parameters, the uncenainty for the secondary side power calorimetric measurement is:

  1. ofloops power uncertainty (% RTP)

+a,e 2

l 26

TABLE 7 POWER CALORIMETRIC INSTRUMENTATION UNCERTAINTIES

(% SPAN)

FW TEMP FW PRES FW D/P STM PRESS

+a,c SRA

=

SCA

=

SMTE =

SPE

=

STE

=

SD

=

BIAS =

RCA

=

RMTE =

RTE

=

RD

=

A/D

=

CSA

=

  1. OF INSTRUMENTS USED 1/ Loop 1/ Loop 1/ Loop 1/ Loop "F

psi

% d/p psi INST SPAN = 150 1600 120% Flow 1400

+a.c INST UNC.

(RANDOM)

=

INST UNC.

(BIAS)

=

800-700-NOMINAL

= 440.7 F 875 psia 100 % Flow 775 psia 1

(a) Included in RCA Feedwater temperature measurement is from channels T-2104 and -2105 Feedwater pressure measurement is from channels P-2289 and -2290 Feedwater flow measurement is from channels F-466, -467, -476 and -477 Steam pressure measurement is from channels P-468, -469, ~478, -479,

-482 and -483.

27

TABLE 8 POWER CALORIMETRIC SENSITIVITIES FEEDWATER FLOW.

+a.c F.

TEMPERATURE

=

MATERIAL

=

DENSITY TEMPERATURE

=

PRESSURE-

=

DELTA P

=

FEEDWATER ENTHALPY TEMPERATURE

=

PRESSURE

=

h,

=

h

=

f Dh (SG)

=

STEAM ENTHALPY PRESSURE

=

MOISTURE

=

i 28

i TABLE 9 SECONDARY SIDE POWER CALORIMETRIC MEASUREMENT UNCERTAINTY COMPONENT INSTRUMENT UNCERTAINTY POWER UNCERTAINTY

_ +a.c FEEDWATER FLOW VENTURI THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION i

BIAS VALUES FEEDWATER DELTA P FEEDWATER PRESSURE DENSITY ENTHALPY STEAM PRESSURE ENTHALPY POWER BIAS TOTAL VALUE SINGLE LOOP UNCERTAINTY (WITHOUT BIAS) 2 LOOP UNCERTAINTY (WITHOUT BIAS) 2 LOOP UNCERTAINTY (WITH BIAS VALUES)

  • *
  • INDICATES SETS OF DEPENDENT PARAMETERS 29

i i

i t

Tiss page left intentionally blank.

l l

30

IV.

RESULTS/ CONCLUSIONS The preceding sections provide the methodology to account for pressure, temperature, power and RCS flow uncertainties for the RTDP analysis. The uncertainty calculations have been performed for Point Beach Units 1 & 2 with the plant specific instrumentation and calibration procedures.

The follow'mg table summarizes the results and the uncenainties that are used in the Point Beach 1 & 2 safety analysis.

l Parameter Calculated Uncertainty -

Uncertainty Used in Safety Analysis Pressurizer Pressure

  • 31.1 psi (random) 50.0 psi (random) 0.0 psi (bias)

Tavg

  • 4.4 T (random)
  • 6.0 T (random)

-1.

as)

(includes bias)

Power

  • l.9% RTP (random) 2.0% RTP (random)

RCS Flow (plant computer)

  • 2.08% flow (random) 2.4% flow (random)

+0.26% flow (bias)

(includes bias)

(calorimetric measurement) 1.9% flow (random)

+0.26% flow (bias) i i

31 g

E REFERENCES 1.

Westinghouse WCAP-11397-P-A, " Revised Thermal Design Procedure", dated April 1989.

2.

Westinghouse letter NS-CE-1583, C. Eicheldinger to J. F. Stolz, NRC, dated 10/25/77.

3.

Westinghouse letter NS-PLC-5111, T. M. Anderson to E. Case, NRC, dated 5/30/78.

4 Westinghouse letter NS-TMA-1837, T. M. Anderson to S. Varga, NRC, dated 6/23/78.

5.

Westinghouse letter NS-EPR-2577, E. P. Rahe Jr. to C. H. Berlinger, NRC, dated 3/31/82.

6.

Westinghouse Letter NS-TMA-1835, T. M. Anderson to E. Case, NRC, dated 6/22/78.

7.

NRC letter, S. A. Varga to J. Dolan, Indiana and Michigan Electric Company, dated 2/12/81.

8.

NUREG-0717 Supplement No. 4, Safety Evaluation Reeort related to the operation of Virgil C. Summer Nuclear Station Unit No.1, Docket 5%5, August,1982.

9.

Regulatory Guide 1.105 Rev. 2, " Instrument Setpoints for Safety-Related Systems",

dated 2/86.

10.

NUREG/CR-3659 (PNL-4973), "A Mathematical Model for Assessing the Uncertainties of Instrumentatior. Measurements for Power and Flow of PWR Reactors", 2/85.

I 1.

ANSI /ANS Standard 58.4-1979, " Criteria for Technical Specifications for Nuclear Power Stations".

12.

ISA Standard S67.04,1994, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants" 13.

Tuley, C. R., Williams T. P., "The Significance of Verifying the SAMA PMC 20.1-1973 De6ned Reference Accuracy for the Westinghouse Setpoint Methodology",

Instrumentation, Controls, and Automation in the Power Industry, June 1992, Vol.35, pp. 497-508.

14.

Scientific Apparatus Manufacturers Association, Standard PMC 20.1,1973, " Process Measurement and Control Terminology" 15.

WCAP-11397-P-A, " Revised Thermal Design Procedure", dated April 1989.

32

v 1

SECONDARY SIDE PRIMARY SIDE l.

P, P

T

% Flow T

1 T

g g

H C

l l

F h,

h pg h,

h g

e W

Ah v

g g

l OSG

+

O#

O L

p

+

Mass

  • L I

Volumetre L

- calculated I

Otis Loop RCS VOLUMETRIC FLOW Figure 1 Calorimetric RCS Flow Measurement (Using LEFM) 33

W SECONDARY SIDE P,

P T

Ap g

g F

h, h

pg F,

K f

Wi I

OSG I

Other Loop i

+

+

k I

9p CORE POWER 0 -c.icui.iea 0

- '"'="=d Figure 2 Calorimetric Power Measurement (Using Feedwater Venturi) 34 i

F

]

NPL 99-0369 - Marked up Technical Specification Changes 1

)

l - Marked up Technical Specifications for implementation of the 422V+ fuel assemblies at Point Beach Nuclear Plant Units 1 and 2 Included in this attachment are the marked-up Technical Specifications indicating the proposed changes necessary for incorporation of the 422V+ fuel assemblies at PBNP. These changes are discussed in detail in Attachment 1 of this submittal. Also included in this attachment is a clean copy of the Technical Specification pages with the changes incorporated.

l l

l 1

l 15.2.0-SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE Applicability:

Applies to the limiting combinations of thermal power, reactor coolant system pressure, and coolant temperature during operation.

Objective:

fi k 0" p[(6. Z..l-2 ^ 5 "fT ca To maintain the integrity of the fuel cladding.

Specification:

1.

The combination of thermal power level, cool t pressure, and cool temperature shall not exceed the limits shown in Figure 15.2.1-1 for Units 1 and 2 The safety limitis exceeded if the point defined by the combination of reactor coolant system average j

temperature and power level is at any time above the appropriate pressure line.

Basis:

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 restiicting fuel operation to within the nucleate boiling regime where the helt 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 excess cladding temperature 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.

Figure 15.2.1-1 applies to Unit 2 following U2R22 and to Unit I following U1R24. Prior to UIR24, Figure 15.2.1-2 applies to Unit 1.

Unit 1 - Am= Ament No 173 15.2.1-1 uly 1,1997 Unit 2-AmendmentNo. 7

This relation has been developed to predict the DNB flux and the location of DNB The local DNS heat for axially uniform and non-uniform heat flux distributions.

flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, is indicative of the margin to g yjypl a0k Of A DNB design basis is as follows:

there must be at least a 95 percent proba-bility at a 95 percent confidence level that DNB will not occur during steady state operation, normal operational transients, and anticipa't'ed transients and is b an appropriate margin to DNB for all operating conditions Ag rmbW igure 15.2.1-1 @ 15.2.1-2_ are\\l applicable ja y

s lQf&

o a core of 14 x 14 The curves 0FA.ffhe curves also apply to the reinsertion of previously-depleted 14 x 14 hndard fuel assemblies into an 0FA coredhe use of these assemblies is s.

The WRB-1 correlation is used to justified by a cycle-specific reload analyst Uncertainties in p1 int parameters and DNB correlation generate these curves.

obtain a DNBR uncertainty factor.

predictions are statistically convoluted t This DNBR uncertainty factor establishes value of design limit DNBR. This value of design limit DNBR is shown to met in plant safety analyses, using g

values of input parameters considered t their nominal. values.

(enb x

D" yhDIb Il[o f "- }

QQ l&)l c>&Y O Y 0

)

cm 4 4n 1ad a sw blieG v

O Qctober 27, 1993 Unit 1 - Amendment No.

142 15.2.1-2 Unit 2 - Amendment No.

146

l I

Figure 15.2.1-1

  • q POINT BEACH NUCLEAR PLANT UNITS 1 AND
  • Q REACTOR CORE SAFETY LIMITS i

670

\\

660 -N l

l 650

~

2425 psia 640 q

}

50 nu
  1. 8630 e

3 ee

$620 h

2000 psia O

z m 610 h

b 600 ms pew j

590 580

~

570 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

1.1 1.2 Core Power (fraction of 1518.5 MWt)

This figure applies to Unit 2 following U2R22 and to Unit 1 following U1R24. Prior to U1R24, Figure 15.2.1-2 applies to Unit 1.

is L

Qg Unit 1 - Amendment No 173 Unit 2 - Amendment No.177

'Ili.s 4'icpuR yp(ies 46 coa te_Qad c vi, %

% t%h:n& % f C58 & UpfhU bFAgath aw bere s

f d t.Q.Wi~h M l

Figure 15.2.1-2*

REACTOR CORE SAFETY LIMITS O

POINT BEACH UNIT 1 660-658-

,,,, 33, Ede-2250 SIA 650-000 51A e

620-p 0

775 PSIA G10-688-590-5BeB.

.I 2

5 4

.5

.s

.7

.0 9

1 1.1 1.2 POWtt treection or nominsti

's figure applies to Unit 1 prior to UlR24. Following UIR24, Figure 15.2.1-1 applies to Unit 1.

O' July 1,199

't 1 - Amendment No.173 Unit 2 - Amendment No.177

LMAY Figure 15.2.1-2 +

I Point Beach Nuclear Plants Units 1 and 2 Reactor Core Safety Limit i

Two Loops in Operation 680 i

l

- 2425 psia 660 - -

- 2250 psia Unweeptable Operation E

i 6,40 - -

2000 psia g

2 620 -

1775 psia b

c.s

~$

Ue 600 --

0 s

h 580 -

Acceptable Operation s s.o -

i

.'8

[

tI3 14 540

.'2

.'4

.6 0

Fraction'of Rated Power T% Z:y ypl;cc Ao rer almd L wh ay.cu,b s

N 4'22 V+-lubOaS% Llies, %d O A ed Wil app l;pc F

or a f aA0 cc et f 4n o i jur0 a sw b w

15.2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION V\\

V Aeolicability:

Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, flow, pressurizer level, and permissives related to reactor protection.

Obiective:

To provide for automatic protective action in the event that the principal process variables approach a safety limit.,

Specification:

1.

Protective instrumentation for reactor trip settings shall be as follows:

A.

Startup protection (1)

High flux, source range - within span of source range instrumentation.

(2)

High flux, intermediate range - 540% of rated power.

(3)

High flux, power range'(low setpoint) - s25% of rated power.

O B.

Core limit protection V

(1)

High flux, power range (high setpoint) - $108% of rated power.

s (2)

High pressurizer pressur 385 psig for operation at 2250 psia primary system pressure s2210 psig for operation at 2000 psia primary system pressurej uh (w 42&f a%W blicS These values apply to Unit 2 following U2R22 and to Unit I following UIR24. Prior to UlR24, the high pressurizer pressure reactor trip setpoint for Unit 1 is s2385 psig.

A y

Unit 1 - Amendment No.173 15.2.3 1 July 1,1997

. Unit 2 - Amendment No.177

(3)

Low pressurizer pressure * - 21905 psig for operation at 2250 psia primary system pressure l

21800 psig for operation at 2000 psia

'I primary system pressure

%(un j

(4 Overtemperature 1

I AT (1 + t 3S)

%c

.s A T. (Ki - K2 (T(1 + t 4S) - T')(1 + T 21 + t ' S ) +

1 S

where alues are applicable to operation at both 2000 psia and 2250 sia unless otherwis 'ndicated)

AT indicated AT at rated power, F

=

o T

average temperature, "F

=

T' 5

2.9"F*

  • P pr squrizer pressure, psig

=

2235Ksig (2250 psia operation only)

P'

=

P' 1985 ph (2000 psia operation only)**

=

K 5

1.19 (225-sia operation only) i K

s 1.14 (2000 ia operation only)*

  • i K2 0.025 (2250 p 'a operation only)

=

K2 0.022 (2000 psi operation only

=

K3 0.0013 (2250 psia eration ly)

=

K3 0.001 (2000 psia ope tion nly)**

=

25 sec

=

Ti 3 sec

=

T2 13 2 sec for Rosemont equivai nt RTD

=

0 sec for Sostman r equivalen TD

=

2 see for Rose nt or equivalent

=

T4 0 sec for Sos an or equivalent R

=

and f(AI)is an even fun on of the indicated difference etween top and bottom detectors of the power ange nuclear ion chambers; with g 'ns to be selected based on measured instrume response during plant startup tests, wh e q, and qs are the percent power in e top and bottom halves of the core respect, ely, and qt + qb s total i

core power in p cent of rated power, such that:

(a) for q, q3 within -17, +5 percent, f(AI) = 0.

(b) for ch percent that the magnitude of q - q3 exceeds +5 per nt, the AT trip s point shall be automatically reduced by an equivalent of 2.0 ercent of rated ower.

These value apply to Unit 2 following U2R22 and to Unit 1 following U1R24.

'or to

UlR24, low pressurizer pressure reactor trip setpoint for Unit 1 is 21790 psig.

These v ues apply to Unit 2 foilowing U2R22 and to Unit 1 following U1R24. Prio o U1

, the values are: T's 573.9"F, P' = 2235 psig, K s 1.30, K = 0.0200, and K =

i 2

3 0.

0791.

O nit 1 - Arnendment No.173 15.2.3-2 July 1,1997

. Unit 2 - Amendment No.177

V)\\$ It S '^

"R

__ Z (c) for each percent that the magnitude of q, - q, exceeds -17 percent, the ATp setpoint

)

shall be automatically reduced by an equivalent of 2.0 percent of rated power.

1 (5

Overpower AT (1 + x3S) s AT.[L-K5( *S +1)(1+ t.S)T-L[T(1 T S)- T'))

T 4

3 I

where (values applicable to operation at both 200 sia and 2250 psia) in 'cated AT at rated power, *F AT,

=

=

aver e temperature, 'F T

T s

572.9

  • K, s

1.09 of ted power

  • 0.026,2 fo ' creasing T K

=

5 0.0 for dec ing T

=

0.00123 for T T K.

=

0.0 for T < T

=

10 sec

=

T3 2 see for Rose nt r equivalent RTD

=

T3

('~')

0 see for Sos or uivalent RTD

'J 2 sec for R emont or uivalent RTD

=

T4 0 sec for stman or eq alent RTD l

(6)

Undervoltag - 2 3120 V.

(7)

Indicated r tor coolant flow per loo -

290 pere t of normalindicated loop w

(8)

Reacto coolant pump motor breaker ope (a)

Low frequency set point 255.0 HZ l

(b)

Low voltage set point 2 3120 V.

These values apply to Unit 2 following U2R22 and to Unit 1 follo U1R24. Prior to U1R24, the values for Unit I are: T s 573.9'F and K s 1.089 of rate ower.

4

_./

Unit 1 - Amendment No.189 15.2.3-3 April 1999 Unit 2 - Amendment No.194

]

I WSE BT

.2C (3) i l Low pressurizer pressure 2

- 21905'psig for operation at 2250 psia primary system pressure

- 21800 psig for operation at 2000 psia primary system pressure and cores not containine 422V+ fuel assemblies These value: apply ic Uni: 2 fc!!cwing U2R22 and :o Uni: 1 fc!!cv ing UIR21 Pricr :c-U1R24, the !cw pressurizer pressure rene:cr trip se:pcin for Uni: 1 i: 21790 psig.

TS 15.2.3.1.B(4)

Overtemperature 1

1 AT (1 + r3S) s ATo (Ki-K2(T(1+ r4S) - T')(1 + r21 + r S ) +

S where (values are applicable to operation at both 2000 psia and 2250 psia unless otherwise indicated) indicated AT at rated power, F ATo

=

T average temperature, F

=

I s

569.0 F (for cores containing 422V+ fuel assemblies)

T' s

572.9 F**(for cores not containine 422V+ fuel assemblies) pressurizer pressure, psig P

=

2235 psig (for 2250 psia operation edy)

(,

P'

=

1985 psig (for 2000 psia operation and cores not containine 422V+ fuel P'

=

assemblies edy)**-

Ki 5

1.16 (for 2250 osia operation and cores containine 422V+ fuel assemblies)

Ki s

1.19 (for 2250 psia operation and cores not containing 422V+ fuel assemblies Ody)

Ki 5

1.14 (foJ 2000 psia operation and cores not containine 422V+ fuel assemblies enlyF 0.0149 (for 2250 osia operation and cores containing 422V+ fuel assemblics)

K; t

=-

0.025 (for 2250 psia operation and cores not containing 422V+ fuel assemblies K2

=

Only) 0.022 (for 2000 psia operation and cores not containing 422V+ fuel assemblies K2

=

edy)**~

0.00072 (for 2250 psia operation and cores containing 422V+ fuel assemblies)

K3

=

0.0013 (for 2250 psia operation and cores not containing 422V+ fuel assemblies K3

=

only) 0.001 (for 2000 psia operation and cores not containine 422V+ fuel assemblies K3

=

edyF 25 see ti

=

T2

=

3 sec 2 sec for Rosemont or equivalent RTD T3

=

et 2-b J

0 sec for Sostman or equivalent RTD

=

2 sec for Rosemont or equivalent RTD T4

=

O sec for Sostman or equivalent RTD

=-

and f(AI)is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where qt and qu are the percent power in the top and bottom halves of the core respectively, and g + qu is total core power in percent of rated power, such that:

(a) for gi-qd within -17, +5 percent, f(AI) = 0 for cores not containing 422V+ fuel assemblies: for a, - as within -12. +5 percent. f(AI) = 0 for cores containing 422V+ fuel assemblies.

(b) for each percent that the magnitudv of gi-qd exceeds +5 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power for cores not containine 422V+ fuel assemblies and reduced by an equivalent of 2.12 percent of rated power for cores containing 422V+ fuel assemblies.

(c)

' for cores not containing 422V+ fuel assemblies. for each percent that the magnitude of gi

- quexceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power: for cores containing 422V+ fuel assemblies. for each nercent that the magnitude of a,- asexceeds -12 percent the AT trin setooint shall be automatically reduced by an equivalent of 2.0 percent of rated power.

4 These values apply :o Uni: 2 fc!!cwing U2R22 and :c Uni: I following UlR21 Prior 10 U1R21, the value; are: T' C 573.9 F, P' = 2235 psig, L c 1.30, L = 0.0200, and L-=

0.000791.

TS 15.2.3.1.B(5)

Overpower 1

1 1

AT (1 + t3S) s ATo[K4-K5( tsS )(l+T S)T-K6(T(1+ T S)-

T S +1 5

4 4

where (values are applicable to operation at b,th 2000 psia and 2250 psia) indicated AT at rated power, *F.

ATo

=

average temperatur, F T

=

I 569.0 F (for cores containine 422V+ fuel assemblies)

T' s

572.9 Fr(for cores not containinc 422V+ fuel assemblies)

L 1.10 of rated power (for cores containing 422V+ fuel assemblies)

K4 s

1.09 of rated powert(for cores not containing 422V+ fuel assemblies) 0.0262 for increasing T K5

=

0.0 for decreasing T

=

l&

5 0.00103 for T '2 T' (for cores containine 422V+ fuel assemblies)

~

N hd ' 2 (-

7 0.00123 for T 2 T'(for cores not containine 422V& fuel assemblies)

K6

=

=.

0.0 for T < T' 10 sec T5

=

2 see for Rosemont or equivalent RTD T3

=

0 see for Sostman or equivalent RTD 2 see for Rosemont or equivalent RTD T4

-=

0 see for Sostman or equivalent RTD (6)

Undervoltage - 23120V_-

(7)

- Indicated reactor coolant flow per loop 290 percent of normal indicated loop flow (8)

Reactor coolant pump motor breaker open (a)

Low frequency set point 255.0 HZ (b)

- Low voltage set point 23120V These values apply :c Unit 2 fc!!cv.ing U2R22 and :c Uni: I fo!!cwing U!R2-1 Prior-te U1R2 ?, me value; for Uni: 1 are: T'd 573.9 F and L f 1.089 cf rated power.

-s

.l C..

Other reactor trips:

(1)

High pressurizer water level - s95% of span (2)

Low-low steam generator watei' level -

220% of narrow range instrument span

% of narrow range instrument span (Unit 1)*

{

\\

(3)

Steam-Feedwater Flow Mismatch Trip - s1.0 x 10' lb/hr (4)

Turbine Trip (Not a protection circuit)

(5)

Safety Injection Sepal (6)

Manual Trip O

\\.

This setting limit applies to Unit i until the narrow rige lower tap is changed to the lower sition consistent with Unit 2.

' Unit 1 - Amendment No. 73 15.2.3-3 a July 1,1997 Unit 2 - Amendment No.177

B. asis The source range high flux reactor trip prevents a startup accident from suberitical conditions from proceeding into the power range. Any setpoint within its range would prevent an excursion from proceeding to the point at which significant thermal power is generated.").

The high flux low power reactor trip provides redundant protection in the power range for a power excursion beginning from low power. This tnp insures that a more restrictive trip point is used for this case than for an excursion beginning from near full power.")

The overpower nuclear flux reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure circuitry. The prescribed setpoint, with allowance for errors, is consistent with the trip point assumed in the accident analysis.0) g*)

The overpower AT reactor trip prevents power density anywhere in the core from exceeding 08'/ of design power density, and includes corrections for change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The specified setpoints meet this requirement and include allowance for instrument errors.*

O The overtemperature AT reactor trip provides core protection against DNB for all combinations of pressui' e, power, coolant temperature, and axial power distribution, provided only that (1) the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds)m, and (2) pressure is within the range between the high and low pressure reactor trips. With normal axial A

9 Unit 1 - Amendment No.123 15.2.3-5 July 31,1989 N

Unit 2 - Amendment No.126 November 1,1989

i power distribution, the reactor tdp limit, with allowance for errors (2) is always below the core safet p

Q limit as shown on Figures 15.2.1-1 and 15.2.1-2. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is ad **

""' ***c'"'*["5'";u c,a;apnn so>

f an+r The overpower, overtemperature and pressurizer pressure system setpoiats include the effect of reduced system pressure operation (including the effects of fuel densification).nThe setpoints will not.

exceed the core safety limits as shown in Figures 15.2.1g and 15.2.1-2Y p e Ayu o%a %~n)

~

The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclear overpower and overpower AT trips.

The high and low pressure reactor trips limit the pressure range in which reactor operation is l

pemdtted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The lowW pressurizer pressure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident O The low flow reactor tdp protects the core against DNB in the event of eith measured flow in the loops or a sudden loss ofpower to one or both reactor coolant pumps. The setpoint specified is consistent with the value used in the accident' analysis

  • The low loop flow t

signal is caused by a condition ofless than 90 percent flow as measured by the loop flow instrumentation. The loss of power signal is caused by the reactor coolant pump breaker opening

% ghed q p $224 + h da Aob hC ftdn. tad ] M [ r*C9 8 ^L cpsh %

NS CA d i ynv+ fua msk k ag83d d zzso psk.

O Unit 1 - Amendment No.

15.2.3-6 July 1,1997 Unit 2 - Amendment No; 177

as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch. The significant feature of the breaker trip is the frequency setpoint,55.0 HZ, which g

assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. The m and transient overshoot in level specifled setpoint allows adequate operating instrument error before the reactor trips.

The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.W Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have *1% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. The difference between the nominal and maximum allowed value (or minimum allowed value) is to account for "as measured" rack drift efTects.

Sustained operation is not permitted with only one reactor coolant pump. If a pump is lost while operating below 50 percent power, an orderly shutdown is allowed. The power-to-flow ratio will be maintained equal to or less than unity, which ensures that the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase above the maximum enthalpy rise which g occurs during full power and full flow operation.

References W FS 14 W FSAR 14.3.1 M FSAR 3.2.1 J m FS age 14-5

  • FSAR 7.2,7.7 M FSAR 14.1.11 l@ J m,i.io 4 l

l i

l O

Unit 1 - Amendment No. 89 l

15.2.3-7 April 23,1999 i

Unit 2 - Amendment No.194

kp(4 (A Whe h

/

g G.

OPERATIO L I IMITATIONS i

The following D related parameters shall be maintained within th imits shown during Rated Power opera 'on:

1.

T,,, shall be main ' ed 2557'F and $573.9*F.

2.

~ Reactor Coolant Syste (RCS) pressurizer pr ure shall be maintained:

22205 psig during operati n at 2250 ps' or 21955 psig during operatio at 2000 sia.

3.

Reactor Coolant System raw asured Total Flow Rate shall be maintained 2181,800 gpm.

Basis:

The reactor coolant syst total flow rate of181,800 m is based on an assumed measurement uncertainty of 2.1 per at over thermal design flow (1,000 gpm). The raw measured flow is based upon the use of no alized elbow tap differential pres which is calibrated against a precision flow calorimetric t the beginning of each cycle.

O Unit 1 - Amendment No.173 15.3.1-19 Unit 2 - Amendment No.177 July 1,1997

,./,

Y p/

~

)

I G. Operational Limitations The following DNB related parameters shall be maintained within the limits shown during rated power operation:

1. T shall be maintained 2558.1 F and $574.0 F for cores containing 422V+ fuel assemblies..

m Tog shall be maintained 2557 F and $573.9 F for cores not containine 422V+ fuel assemblies.

' 2. Reactor Coolant System (RCS) pressurizer pressure shall be maintained:

22205 psig during operation at 2250 psia, or l

21955 psig during operation at 2000 psia for cores not containing 422V+ fuel assemblies.

4

3. Reactor Coolant System raw measured Total Flow Rate shall be maintained 2182.400 enm for cores containing 422V+ fuel assemblies, or 2181,800 gpm for cores not containing 422V+ fuel assemblies.

TS 15.3.1 Basis:

The reactor coolant system total flow rate of 182.400 gnm for cores containing 422V+ fuel assemblies is based on an assumed measurement uncertainty of 2.4 percent over thermal design flow (178.000 epm). The reactor coolant system total flow rate of 181,800 gpm for cores not containine 422V+ fuel assemblies is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm). The raw measured flow is based upon the use of nonnalized elbow tap differential pressure which is calibrated against a precision flow

,' calorimetric at the beginning of each cycle.

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  • d U U N

l

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1 AND e

b.

Within two hours fully withdraw the shutdown banks.

l l

c.

If the above actions and associated completion times are not met. be in hot l

shutdown within the following six hours.

2.

When the reactor is critical, the control banks shall be inserted no further than the limits shown by the lines on Figure 15.3.10-1. If this condition is not met, l

perform the following actions:

Within one hour verify that the shutdown margin exceeds the applicable a.

value as shown in Figure 15.3.10-2; DE within one hour restore the shutdown margin by boration; AND b.

Within two hours restore the control banks to within limits.

c.

If the above actions and associated completion times are not met, be in hot shutdown within the following six hours.

lE.

POWER DISTRIBUTION LIMITS l

1.

Hot Channel Factors lVeO l a.

The hot channel factors defined in the basis shall meet the following gQ for 472V+ b-limits:

Fg(Z) S ( '5 } x K(Z) for P > 0.5 VWh 8h O

P p(g)66.20*$b:-)

F (Z) s 5.00 x K (Z) for P S 0.5 g

[64 '- ('@ X Ll + o.3(I-h fm < l.70 x [l + 0.3 (1-P)]

Where P is the fraction of full power at which the core is operating, K(Z) is the function in Figure 15.3.10-3pnd Z is the core height location of Fq.

Lor-Ryu. W.E to-%,Qp%b(.e 3 b.

If F (Z) exceeds the limit of Specification 15.3.10.E.1.a. within fifteen q

minutes reduce thermal power until F (Z) limits are satisfied; o

(1)

After thermal power has been reduced in accordance with Specification 15.3.10.E.1.b, perform the following actions:

.O

%)

Unit 1 - Amendment No.171 Unit 2 - Amemdment No.175 15.3.10-5 January 16,1997

Power Distribution fM hC m

I During power operation, the global power distribution is limited by TS 15.3.10.E.2, " Axial Flux Difference," and TS 15.3.10.E.3, " Quadrant Power Tilt," which are directly and continuously measured process variables. These specifications, along with TS 15.3.10.D, " Bank Insertion Limits." maintain the core limits on power distributions on a continuous basis.

-41 The purpose of the limits on the values of F (Z), the height dependent heat flux hot channel 9

factor, is to limit the local peak power density. The value of Fn(Z) varies along the axial height (Z) of the core.

Fn(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, F (Z) is a measure of the peak fuel pellet power within the reactor core.

n Fn(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution. F (Z) is measured periodically using the incore detector system. These 9

measurements are generally taken with the core at or near steady state conditions.

The purpose of the limits on F"3n, the nuclear enthalpy rise hot channel factor, is to ensure that the fuel design criteria are not exceded and the accident analysis assumptions remain valid. The design limits on local and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that

(

local conditions in the fuel rods and coolant channels do not challenge core integrity at any i

Iqcation during either normal operation or a postulated accident analyzed in the safety analyses.

F AH, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral oflinear N

power along a fuel rod to the average fuel rod power. Imposed limits pertain to the maximum F"3a in the core, that is the fuel rod with the highest integrated power. It should be noted that AH s based on an integral and is used as such in the DNB calculations. Local heat flux is N

i F

obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal power shape at the point of maximum heat flux is not necessarily directly related to F"3a.

F"an is sensitive to fuel loading pattems, bank insertion, and fuel burnup. F"3s typically increases with control bank insertion and typically decreases with fuel bumup.

F"3a is not directly measurable but is inferred from a power distribution map obtained with the movable incore detector system. Specifically, the results of the three dimensional power distribution map are analyzed by a computer to determine F"an. This factor is calculated at least l

monthly. However, during power operation, the global power distribution is monitored by TS l

15.3.10.E.2, " Axial Flux Difference," and TS 15.3.10.E.3, " Quadrant Power Tilt." which address l

directly and continuously measured process variables.

I Unit 1 - Amendment No.

l hary 16,19D Unit 2 - Amemdment No.175 15.3.10-14 J

m bMut As a result of the increased peaking factors allowed by the new 422V+ fuel, a new column was added to TS 15.3.10.E.1.a. The full power F"an peaking factor design limit (radial peaking factor) for 422V+ fuel will increase to 1.77 from the 1.70 value for the OFA fuel. The maximum Fo(Z) peaking factor limit (total peaking factor) for 422V+ fuel will increase to 2.60 from the i

N 2.50 value for the OFA fuel. The OFA fuel design will retain the current F n nd Fo(Z) a peaking factors of 1.70 and 2.50, respectively. In addition, the K(Z) envelope for the new 422V+ fuel was modified and a new TS figure 15.3.10-3a was developed and insened in the Technical Specifications. The K(Z) envelope in TS Figure 15.3.10-3 remains for the OFA fuel.

i I

i l

l i

f^

& f6 f d Y VI f '

y opy p

l It has been determined that, provided the following conditions are observed, the hot channel Q

factor limits will be met:

l.

Control rods in a single bank move together with no individual rod insertion differing by more than 24 steps from the bank demand position, when the bank demand position is between 30 steps and 215 steps. A misalignment of 36 steps is allowed when the bank l

position is less than or equal to 30 steps, or, when the bank position is greater than or equal to 215 steps, due to the small worth and consequential effects of an individual rod

. misalignment.

2.

Control rod banks are sequenced with overlapping banks as described in Figure 15.3.10-1.

l3.

Control bank insertion limits are not violated.

4.

Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits, are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

O The perm ed relaxation of F AH allows radial power shape changes with rod insertion to the N

insertion lim' It has been determined that provided the above four conditions are observed, qthese hot c el factor limits are met. In Specification 15.3.10.E.1. Fn is arbitrarily limited for p s 0.5.

(

q gg,3,(o, ko.

upper bound velo of 2.5 es the normalized peaking factor axial dependence of Figure 15.3.10-onsi nt with the Technical Specifications on power distribution control as given in Sectio _15.3. 0[was used in the large and small break LOCA analyses. The envelope was determined basedpn allowable power density distributions at full power restricted to axial l flux difference (AI) values consistent with those in Specification 15.3.10.E.2.

The results of the analyses based on this upper bound envelope indicate a peak clad temperature ofless than the 2200 F limit. When an Fn measurement is taken, both experimental error and l manufacturing tolerance must be taken into account. Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance. In the design limit of F"3s,

. there is eight percent allowance for uncertainties which means that normal operation of the core s 1.70/1.0% The logic behind the large[r un is expected to result in a design F"3s l case is as follows:

Q 4

(a)

Normal perturbations in the radial power shape (i.e., rod misalignment) affect F"3s, in most cases without necessarily affecting Fq.

Unit 1 - Amendment No.

- s Unit 2 - Amemdment No. 75 15.3.10-15 Janua,ry16,1997

t FIGURE 15.3.10-3 POINT BEACH UNITS 1 AND 2 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE pop _ oM Aub (APGRA4Eh 0FA FUEL-1.2 (0.0.1.0)l (6.0.1.0)1 3,

(12.0. 92) l

.9

.6 t

.7

.6

.4

.3

=

.2

.1 00 1

2 3

4 S

e 7

a 9

to 11 12 CORE BIGrf (FT)

/~~T eY N

Unit 1 - Amendment No.120 May 8,1989 Unit 2 - Amendment No.123

' November 1,1989 /

i FIGURE 15.3.10-3A POINT BEACII UNITS 1 AND 2 IIOT CIIANNEL FACTOR NORMALIZED OPERATING ENVELOPE FOR 422V+ FUEL 1.2 1.1 (0.0,1.0)

(6.0,1.0)

(12.0,1.0) 1 0.9 0.8 0.7 k

g 0.6 0.5 0.4 0.3 O.2 0.1 0

0 1

2 3

4 5

6 7

8 9

10 11 12 COREIIEIGIIT (Fr)

O 15.5.3 REACTOR 1

Applicability Applies to the reactor core, Reactor Coolant System, and Emergency Core Cooling Systems.

\\

Objective To define those design features which are essential in providing for safe system operation.

Specifications A.

Reactor Core ytM 1.

General o

The uranium fuel is in the fo of slightly enriched uranium dioxide pellets.

The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.

The reactor core is made up'of 121 fuel assemblies.

Each II) fuel assembly nominally contai s 179 fuel rods Where safety limits are not violated, limit ed substitutions of fuel rods by filler i

v l

rods consisting of Zirealoy 4 or stainless steel, or by vacancies, may j

be made to replace damaged fuel rods if justified by cycle specific reload analysis.

g.Q ggs/+ 3/ bred lUJN DFA m gg C

2.

Core

) or G/

A reactor core.is a core loadin atterncontaininganycombinationoff.

14x14 0FA and 14x14 upgraded 0F fuelassemblies/Thecoremayalso

~

ontain previously depleted 14x14 standard fuel assemblies. The use of s

reviously depleted 14x14 standard)fue1~assembliDs w111 be justified by acyclespecificreloadanalysis.I

-f ftLL O

Unit 1 - Amendment No.120 15.5.3-1 May 8, 1989 November 1, 1989 Unit 2 - Amendment No. 123

I 15.5.4 FUEL STORAGE if m Anplicability Applies to the capacity and storage arrays of new and spent fuel.

~

Obiective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.

Specification l.

The new fuel storage and spent fuel pool structures are designed to w'thstand the anticipated earthquake loadings as Class I structures. The spent fuel pov has a stainless steel liner to ensure against loss of water.

2.

The new and spent fuel storage racks are designed so that it is impossible to store assemblies in other than the prescribed storage locations. The fuel is stored vertically in an array with sufficient center-to-center distance between assemblies to assure K <0.95 d

with the storage pool filled with unborated water and with the fuel loading in the Q. M O

assemblies limited to 5.0 w/o U-235,with or without axial blanket loadings. Each [s O

?**

assembly with a fuel loading greater than 4.6 w/o U-235 must contain Integral Fuel)

Bumable Absorber (IFBA) rods in accordance with Figure 15.5.

ve a reference l mfinite multiplication factor, K., less than or equal to 1.49364, which includes a 1% AK Q

reactivity bias.r) An inspection area shall allow rotation of fuel assemblies for visual inspection, but shall not be used for storage.

3.

The s nt fuel storage pool shall be filled with borated water at a concentration of at least N

ppm boron whenever there are spent fuel assemblies in the storage pool.

I 4.

Except for the two storage locations adjacent to the designated slot for the spent fuel "5

storage rack neutron absorbing material surveillance specimen irradiation,ppent fuel assembly storage locations immediately adjacent to the spent fuel pool perimeter or divider walls shall not be occupied by fuel assemblies which have been suberitical for Fe [)ud asmbun -iW *ar%'d;ck *'d'I less than one year.

12sx A f.O we iptpaat (A2K -d a Waim a() 32 I BA rd 5 cm ulitia $\\\\ aud laN. ad % Velt sky e e00 s p

V Unit 1 - Amendment 179 15.5.4-1 September 4,1997 Unit 2 - Amendment 183

1 i

NPL 99-0369 AttachnL. 5 - Westinghouse RTDP j

Part A of Attachment 5 " Westinghouse Revised Thermal Design Procedure (RTDP)

Instrument Uncertainty Methodology for WE Point Beach Units 1 and 2 (Fuel Upgrade and Uprate to 1656 Mwt NSSS Power)" - WCAP-14787 (Proprietary) l l

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