ML19322B773

Reactor Coolant Flow Evaluation, Preliminary Rept

ML19322B773
Person / Time
Site:	Oconee
Issue date:	08/23/1973
From:	DUKE POWER CO.
To:
Shared Package
ML19322B771	List: ML19322B769 ML19322B773
References
NUDOCS 7912050766
Download: ML19322B773 (6)

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s O OCONEE NUCLEAR STATION UNIT 1 REACTOR COOLANT FLOW EVALUATION Preliminary Report August 23, 1973 Introduction

)conee Unit 1 was designed for a minimum primary coolant flow rate of 131.32x106 pounds per hour. A greater flow rate than the min 4="= is expected, however.

While this will afford excess DNB protection, a flow rate of 110.8% design flow has been specified by the Babcock & Wilcox Company as the upper limit to avoid core lift at the end of life.

A test was performed during the Power Escalation Sequence at the 75% full power plateau to verify that the magnitude of the primary system flow is within acceptable limits. The details of this test are delineated herein.

Evaluation The basis of the flow calculation is a calorimetric around the two steam genera-tors. Thermal-hydraulic data was monitored for an hour on July 29. 1973, properly averaged, and substitued into the heat balance equation described below to provide primary flow.

Figure 1 is a schematic of a steam generator with ita associated coolant flow loops; the dotted line represents the control volume for the derivation of the calorimetric equation. Since the energy entering the volume must leave it in some form, the following balance for the A generator can be made.

44+44-44+44++

A similar equation exists for the B steam generator. Both can be swived for ,

primary coolant system flow and are presented below. I

+

P"

=

( -

)+d (H - )+K

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C Precision thermocouples and dead-weight gages were installed on the feedwater and l steam lines to measure temperatures and pressures to calculate enthalpies.

Precision manometers were used to measure the pressure drop across the cali-brated Bailey flow nozzles for the feedwater and steam flow determination. The plant process computer was used to monitor the primary side temperatures and pressures and feedwater temperature.

Hanometer readings were taken every two minutes for the duration of the test.

Steam secondary side temperatures and feedwater pressures were recorded on a five minute interval while primary side temperatures and pressures arid feed-water temperature were monitored on a 15 second basis. The data was averaged and the flow and enthalapies were calculated.

,moso7G

t The h' eat loss term represents the surface radiation and/or convection from the surface of the piping and the steam generators. This term has minor signifi-cance but is included for completeness. Its magnitude is taken as 0.724 and 0.787 million BTU /hr for loops A and B, respectively.

Table 1 is a listing of the average values of the data collected during the test. The calculated enthalpies and flows are displayed in Table 2. The flow equation is shown below with the proper values inserted and the primary flow noted.

Wp = (1251.03 - 415.28) 4.0815 + 0.724 x 10 0 609.00 - 561.31

+ (1251.69 - 415.28) 3.9642 + 0.787 x 10 609.27 - 561.20

= 140.34 M lbm/Hr The error analysis for the above flow value is derived in Appendix A. The result of the error analysis yielded a band of 1 1.146 M lbm/Er.

Since minimum design flow is 131.32 M lbm/hr at rated power which corresponds to 130.2 M lbm/hr at 75% power, the measured flow and experimental error is 107.8 i .82 as expressed in percent.

Safety Analysis The minimum RC system flow rate shall be the FSAR basis of the 100% (131.32 x 106 lb/hr, minimum design flow at rated power) plus 2.3% excess for bypass due to removal of 44 orifice plugs. This flow rate is established as the minimum flow rate to meet the DNBR requirements stated in the FSAR. Therefore, the minimum flow shall be 134.34 x 100 lb/hr at rated power.

The maximum reactor coolant system flow rate is 110.8% of the minimum design flow rate based on fuel assembly lift limitations. This 10.8% excess flow design limit is determined by utilizing experimental evidence of fuel assembly hydraulic resistance characteristics and the maximum expected flow rate for any fuel assembly based on flow distributions from the Vessel Ebdel Flow Test. This maximum allowable flow rate is based on the more limiting end-of-life conditions.

The measured system pressure loss is lower than predicted and represents a design conservatism. Also, the modification of the reactor vessel and internals resulted in a reduction of the reactor vessel unrecoverable pressure loss. The reduction in reactor vessel pressure loss due to the internals changes is approximately 4 psi at the design flow rate. (Reference BAW-10037, Rev. 2, November 1972, " Reactor Vessel Model Flow Tests.") These two points account for the actual RC systen flow rate being above minimum design flow rate.

Therefore, the reactor coolant system flow including possible measur.ement error for Oconee 1 is within acceptable limits.

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TART.E 1. AVERAGED DATA Loop A Loop B Main Steam, Temperature, *F 590.34 590.80 Pressure, psia 911.73 912.22 Feedwater, Temperature, *F 436.47 436.25 Pressure, psia 942.61 939.00 AP, psi: Tap 1 35.64 35.25 Tap 2 35.95 33.32 Hot Leg, Temperature, *F 596.60 596.86 Pressure, psia 2122.0 2141.7 Cold Leg, Temperature, *F 560.997 560.945 Pressure, psia 2089.4 2109.1 TABLE 2. HEAT BALANCE DATA Enthalpies (BTU /lbm) Loop A Loop B Main Steam 1251.03 1251.69 Feedwater 415.28 415.28 Hot Leg 609.00 609.27 Cold Leg 561.31 561.20 Feedwater Flow (M lbm/Hr) 4.0815 3.9642 Heat Losses 0.724 0.787

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's FIGURE 1. LOOP 1 STEAM GENERATOR p HOT LEG W = Total Primary Coolant Flow i i T

= Loop i Primary Flow H..PH i

TH"L P 1 Hot Leg Temperature

/ \ STEAM g

I F P = Loop i Hot Leg Pressure T,, P,

= Loop i Hot Leg Enthalpy HEAT LOSSES FEEDWATER T = Loop i Cold Leg Temperature l l 1 i

s i F MF P = Loop i Cold Leg Pressure .

j P1y  ;

i \

H = Loop i Cold Leg Enthalpy 1- 1 COLD LEG c T P c c 1 i

T s"L P i Steam Temperature' l 1

P, = Loop i Steam Pressure H, = Loop i Steam Enthalpy Tf=LoopiFeedwaterTemperature i

Pp = Loop i Feedwater Pressure

= Loop i Feedwater Enthalpy

= Loop i Feedwater Flow l

l

[ = Loop i Heat Losses  !

APPENDIX A The basic flow equation from Figure 1 is as follows:

W p

=( -

)k +d + <d - )E+dp 80 - se 4-4

-W ,= W<x1 ,x,,....x,3 and dW p

=" eW i=1 6xi Thcrefore dW = -

+

p dk d + d

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N 0

d + dP^

n;-u; geq 64 >

_(4-4)4+efe4 dr"i . e80

,,i)

<<-u;>2 eq eq j

+ <n0- 4) 4 + e faus dr c + ensdriS i

<<-ug,2 (,,g ,4 c, de 4 - n;

. ui-u"r g - < ,g . u; - ua (eg r

4 M d4 +4 eq g>

4 ie4 ap e

<s;- 4>4 + d ( e4 dg + 64 def g - g og r 64 4 )rj <g-g> 2 ieg eg j

( -

) + 6a

+ C dT + 6a dP +

<<-<>2 i eg eq C

> <-g ..

r e d'fferentials, i dTA can be replaced by finite diff erences, A , representing the measurement toler5n,es c for each variable' substituted. The measurement tolerances are given below:

Main Steam Temperature + 0.5*F Main Steam Pressure + 1 psi Feedwater Temperature + 0.5"F

.i.

Feedwater Pressure - 1 psi Feedwater Flow + 0.5%

RC Hot Leg Temperature + 0.25'F

+

RC Hot Leg Pressure - 25 psi RC Cold Leg Temperature 0.25'F RC Cold Leg Pressure + 35 psi Ambient Heat Losses +- 50%

The heat balance data from Table 2 is substituted for the feedwater flow and enthalples. The values for the rate of change with respect to the differential are substituted for the partial derivation.

The terms of AWp are the squared, summed,and the square root taken. The terms represent the error in feedwater flow, steam temperature, steam pressure, feedwater temperature, feedwater pressure, reactor coolant hot leg temperature, reactor coolant hot leg pressure, reactor coolant cold leg temperature, reactor coolant cold leg presr ambient heat loss measurements.

l 1

l 1

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