ML20072H322
| ML20072H322 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/21/1983 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| 923, NUDOCS 8303290393 | |
| Download: ML20072H322 (17) | |
Text
r-g,\\~
TOLEDO
%mm EDISON ArcaAno P. CROUSE Docket No. 50-346 Vce fhtsdent Nucles License No. NPF-3 "S 25S522' Serial No. 923 March 21, 1983 Mr. John F. Stolz, Director Nuclear Reactor Regulation Operating Reactor Branch No. 4 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Stolz:
Toledo Edison acknowledges receipt of your January 26, 1983 letter (Log No. 1197) requesting additional information for the proposed change to Appendix A, Technical Specification 3.2.5 - Flow / Thermal Power Setpoint Change - for the Davis-Besse Nuclear Power Station, Unit 1.
Please find enclosed the requested information.
Very truly yours, RCP:JAE Enclosure i
dh e/4 Q0' THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652 8303290393 830321 PDR ADOCK 05000346 P
& 9A3 Qu:stion 1.
Your " Safety Evaluation" in support of thn proposed change on Technical Specification 3.2.5 indicates the reactor coolant measurement uncertainty of 2.5%.
Provide a detailed descrip-tion on how the RC flow measurement uncertainty is obtained, including a detailed breakdown of measurement components and uncertainty associated with each component.
Response
The attached report entitled " Determination of Total RC Flowrate and its Accuracy for Davis-Besse 1" provides a detailed description of how the RC flow measurement uncertainty is obtained.
A detailed breakdown of measurement components and the uncertainty associated with each component is included.
This report was transmitted via the attached letter from Lowell E. Roe to John F. Stolz, dated May 26, 1978.
The flow measurement accuracy deter-mined in the report is 2.2%.
The value of 2.5% which is indicated in the
" Safety Evaluation" mentioned above is conservative with respect to the calculated value.
Qdestion 2.
The same Safety Evaluation states that B&W has performed calcu-lations to determina the DNBR margin gain for the proposed RC flow and power tradeoff.
Is the B&W analysis done specifically for Davis Besse? Provide the B&W analysis report.
Question 3.
Figure 1 of your submittal gives a relationship between MDNBR, calculated with BAW-2 correlations, and the thermal power reduc-tion factor N.
Also, the proposed Technical Specification change uses N=2(%).
Are the figure and the associated analysis based on current Davis Besse fuel design and loading? Do the analysis and the N=2 bound all fuel loading and fuel design for the future cycles?
Response
INTRODUCTION Section 3.2.5 of the Davis Besse Technical Specifications (Reference 1) sets acceptance criteria on the DNB related parameters of RCS pressure, hot leg flow temperature, and RCS flow rate.
If any of these parameters exceed the prescribed limits, the parameter must be restored to within its limits within two hours or the reactor must be reduced to 5% of rated thermal oower within the next four hours (rated thermal power = 2772 MWt).
The minimum reactor flow rate specified in Reference 1 is 396880 GPM with 4 RC pumps operating (includes a 2.5% flow rate uncertainty).
This minimum flow criterion is based upon the 110% of design flow assumed for DNB analyses (design flow = 88000 GPM/ pump).
The minimum RC flow rates measured at DB-1 are typically in the range of 111 to 112% of design flow (including the 2.5%
flow uncertainty reduction).
Toledo Edison desires to modify the action requiring a reduction in reactor thermal power to 5% of rated power in the event measured flow is outside its allowable value.
This modification will require a reduction of the maximum allowable thermal power by 2% for every 1% the RC flow rate is below the Tech. Spec. 3.2.5 limit (which would be unchanged).
This Tech. Spec. modification is reconinended only to cover small changes in measured steady state flow, on the order of si to 2 percent, and not large changes in measured flow rate which obviously would indicate more serious concerns.
This report describes the criteria, methods, results, and conclusions of the analyses supporting the proposed Tech. Spec. change.
, CRITERIA The purpose of Tech. Spec. 3.2.5 is to ensure that each of the DNB-related parameters is maintained within the normal steady state envelope that was assumed as an initial condition in transient and accident analysis.
Ul timately then the purpose of Tech. Spec. 3.2.5 is to ensure that the DNBR safety limit is not violated during steady state or moderate frequency transient conditions.
Since any changes to the transient initial (time = 0) DNBR (i.e., steady state DNBR) will tend to be carried through to the transient minimum DNBR, then the appropriate basis for judging the proposed Tech. Spec. change is the effect of the proposed combinations of reactor power and RC flow rate on the steady state minimum DNBR.
The datums for these comparisons should be the minimum DNBR's at 08-1 design over power conditions for 4 RC pump and 3 RC pump opera-tion.
The minimum DNBR's to be compared to the design over power cases should be calculated at sets of conditions defined by the following formulas:
1.
For 4 RC Pump Operation:
RC Flow Rate = (110 - N)% of design 4 pump flow Reactor Power Level = (112 - 2N)% of rated power 2.
For 3 RC Pump Operation:
RC Flow Rate = (110 - N)% of design 3 pump flow Reactor Power Level = (90 - 2N)% of rated power where N = arbitrary integer (1, 2, 3, etc.)
If the minimum DNBR's increase with increasing values of N, then the trade-off of power for flow (2 for 1) will have been shown to be conservative in tenns of minimum DNBR, thus demonstrating that the proposed Tech. Spec. change
~
is conservative.
. METHODS t
The methods employed in the analyses supporting the proposed Tech. Spec.
changeareasdescribedinChapter4.4oftheDB-kFSAR(Reference 2).
Briefly,
there were two steady state DNBR analyses performed specifically for Davis Besse:
one with 4 RC pump operation and one with 3 RC pump operation.
These analyses conservatively bound the current Davis Besse fuel design and all expected loadings.
No fuel design changes that would impact the results of these analyses are planned for Davis Besse.
Maximum design conditions were assumed for both DNBR analyses.
Maximum design conditions assume that the most conservative nuclear, thermal, and mechanical conditions exist simultaneously in a particular subchannel.
The maximum design conditions at DB-1 are represented by the following assumptions:
1.
A maximum fuel pin radial - local power factor (Fah) of 1.714.
2.
A symetric cosine axial flux shape with a max./ avg. peak (F ) of 1.5.
7 3.
The limiting fuel assembly is assumed to receive only 95% of the average fuel assembly flow.
4.
The limiting fuel assembly is assumed to have a reduced peripheral flow area because'of adjacent fuel assembly proximity.
5.
A maximum core pressure error of -65 psi is assessed against the nominal RCS pressure.
6.
A maximum RC inlet temperature error of +2 F is assessed against the nominal inlet temperature based on heat balance.
7.
Three engineering hot channel factors are applied to the subchannel types with maximum Fah values to account for as-built variations of key parameters:
A.
The subchannel flow area is reduced by a factor (FA) of.97 or
.98 depending on subchannel type.
B.
The fuel pin local surface heat flux is increased by a factor (FQ") of 1.014.
C.
The fuel pin heat output (i.e., subchannel enthalpy rise) is increased by a factor (FQ) of 1.011.
- - + ~
m -
8 A maximum core bypass flow fracticn of 10.7% which includes the absence of all orifice rods.
9 A miriimum fuel stack height resulting from fuel densification.
~
The computer codes used in the two DNBR analyses are the CHATA code (Reference 3) for core flow distribution on an assembly by assembly basis and the TEMP code (Reference 4) for detailed DNBR analysis of the limiting fuel assembly.
The CHATA code calculates the bund:-) by bundle flow distribu-tion for a core given the power level, inlet temperature and pressur total core flow rate, radial power distribution, and hydraulic characteristics of the fuel assemblies, assuming a uniform pressure drop across the core.
The flow rate of the limiting fuel assembly calculated with CHATA is input to the TEMP model.
TEMP will distribute this input flow rate among the fuel assembly subchannels according to the local power distribution and the hydraulic charac-teristics of the different subchannels assuming a constant pressure drop across the fuel as:embly.
TEMP allows the transfer of energy (enthalpy) between subchannels by turbulent mixing but allows no mass interchange.
CHATA and TEMP are the steady state thermal hydraulic codes that were used for DB-1 licensing.
i
'RESULTS The analysis with 4 RC pumo operation assumed 112% of rated power and 110% of design flow as base case conditions.
These are the design over power conditions for DB-1.
The minimum DNBR of this base case is compared with the minimum DNBR's of the other 4 RC pump cases in Figure (.
The "N" parameter on the abscissa of Figure 2 relates to the power and flow conditions of each case as defined by equation in the " Criteria" section of this report and as listed below for illustration.
N_
% Rated Power
% Design Flow
- 0 112 110 2
108 108 4
104 106 6
100 104
- 100% design flow is the RC pump design flow rate of 88000 GPM/ pump.
The range of power and flow investigated was deemed adequate because of the clearly evidenced trend in Figure h, that minimum DNBR increases with increas-ing "N" value.
This trend indicates that with the proposed Tech. Spec. change is place, an RC flow rate below the Tech. Spec. 3.2.5 limit would restrict allowable reactor power to a level where greater themal margin would exist than assumed in the initial conditions for transient analysis.
This strongly implies, that greater themal margin would exist during a transient initiated from N > 0 conditions than from N = 0 conditions.
Therefore the consequences of transients analyzed in Chapter 15 of the 08-1 FSAR will remain bounding with the proposed Tech. Spec. change enacted, so long as measured flow is
~
100 percent or greater of the pump design flow rate.
The analyses with 3 RC pump operation assumed 90% of rated power and 110% of design 3 pump flow as base case conditions.
The 90% power level was arbitrarily chosen as a bounding 3 pump allowable power.
The minimum DNBR's of this base case and other 3 RC pump cases are compared in Figure 2.
The "N" parameter on the abscissa again relates power and flow as defined by equation in the " Criteria" section of this report and as listed below for illustration.
l
N_
E Rated Power
% Design 3 Pumo Flow 0
90 110 2
86 108 4
82 106 6
78 104 As with 4 RC pump operation, the minimum DNBR increases with increasing "N"
value with 3 RC pump operation.
For the same reasons as with 4 RC pump operation, therefore, the proposed Tech. Spec. change will not adversely affect the consequences of transients analyzed with 3 RC pumps operating.
e o
s m
.v.-,
CONCLUSIONS The conclusion of the two analyses supporting the proposed Tech. Spec.
3.2.5 change, and of this report, is that the proposed Tech. Spec. change is conservative in terms of minimum DNBR.
The improvement in DNBR margin over that associated with the assumed transient initial conditions leads to the further conclusion that 08-1 FSAR Chapter 15 analytical results would not be adversely affected by the proposed Tech. Spec. change.
Thus the intent of Tech. Spec. 3.2.5, to ensure that the consequences of existing transient analyses remain bounding by preserving the transient initial conditions, is clearly satisfied.
O
REFERENCES i) 05-0011-16, " Davis Besse Nuclear Power Station Unit 1 Technical Speci-fications", October 2, 1980.
2)
Davis Besse Unit 1 Final Safety Analysis Report, Docket #50-346.
3)
CHATA - Core Hydraulic and Themal Analysis, J. M. Alcorn and R. H. Wilson, BAW-10110, Babcock & Wilcox, January 1976.
4)
TEMP - Thermal Enthalpy Mixing Program, BAW-10021, Babcock & Wilcox, April 1970.
l
{
FIGURE i Minumum DNBR as a Function of Flow and Power Parameter "N" with 4 RC Pump Operation am-2-
CQ k
I N
82 -
1 m ' tm -
aN' PARAMETER DEFINED AS:
I4 Z
~
FLOU=(110-N)4 0F 4 RC PUMP DESIGN g
3 N
m_
POWER =(112-2N)X OF RATED E
L76 -
y m
i i
0 1
2 3
4 6
6 Flow and Power Parameter "N"
i 1
.~
i FIGURE 2 Minumum DNBR as a Function of Flow and Power Parameter "N" with 3 RC Pump Operation 22-
\\
tu m-1 m
i 2-g na
'N' PARAMETER DEFINED AS:
'4 A
h FLOU=(110-H)* OF 3 RC PUMP DESIGN a
a u-POWER =(90-2H)* OF RATED l
1.7 0
1 2
3 4
6 6
Flow and Power Parameter "N"
', 4.',Whsn opsrating with the reduced PC flow and powar in accordance with the.
proposed Technical Specification, are your current design saf ety analyses with respect to all anticipated operational occurrences and accidents still valid?
Response
Background:
If during power operation, measured RC flow decreases, power will be reduced according to the followina scheme set in the Technical Soecifications:
For every IX drop in measured RC flow, power will be reduced by 2%.
This rule will be applied for both three and four pump initial operation.
The FSAR analysis was performed assuming an initial condition of 102% power and 100 % design flow. With the above scheme, the plant may run at lower power and flow than what was assumed in the FSAR. The FSAR analysis has been reviewed to assess the impact of'the different initial conditions. The results of this assessment are summarized in Table I and the following paragraphs.
Results of the Assessment:
An important assumption in this review is that flow will not drop below 100%
of design flow.
At the present time Davis-Besse 1 is operating at 110% of design flow. Thus, measured flow is calibrated to 110% of design flow. This review will apply only to measured flows > 91%.
Since it is assumed that RC flow rate will be > design flow, only reduced power levels need be addressed in this review.
1 Each FSAR transient was examined to determine the impact of a lower initial power level. The results are summarized in Table 1.
Some general conclusions can be drawn from this table:
a.
For overheating transients, a lower initial power level means less heat added to the RCS and thus lower peak RCS pressure.
b.
For overcooling transients, a lower initial power level may give a delay in reactor trip. However, for most transients sensitivity studies have been performed that cover the variation in power level.
==
Conclusion:==
Based upon a review of the FSAR analysis, the proposed Tech Spec is within the current design safety analyses with respect to all ant.icipated operational occurrences and accidents.
Table 1 Transient Criteria Assessment of Impact Conclusion 15.2.1 Uncontrolled Control Thermal Power <112%
Power Operation limits have no FSAR analysis still Rod Assembly Group Withdrawal RCS Pressure < 110%
impact on startup conditions.
appl icable.
from a Subcritical Condition of design.
15.2.2 Uncontrolled Control Thermal Power <112%
For flow 3,100% of design and FSAR analysis still Rod Assembly Group Withdrawal RCS Pressure < 110%
power < 102%, the time to high applicable.
at Power of design.
flux trip will be increased.
This is addressed in the sensitivity studies on trip delay time and rod withdrawal rate.
15.2.3 Control Rod Assembly Thermal Power <l12%
Thermal power does not exceed FSAR analysis still Misal ignment RCS Pressure < 110%
initial conditions. This is applicable.
of design.
a depressurization event so peak RCS pressure is not a concern.
15.2.4 Makeup and purifica-Thermal power <112%
For flow 2.100% of design and FSAR analysis still tion System Malfunction RCS pressure < 110%
power < 102%, the time to trip appl icabl e.
i of design.
1% AK/K will be delayed. This is Shutdown margin.
addressed in the sensitivity study on dilution rates.
Power Operation has little impact on post trip shutdown margin.
15.2.5 Loss of Forced DNBR > 1.3 From FSAR Figure 15.2.5-3 FSAR analysis still Peactor Coolant Flow.
it is seen that MONBR applicable.
increases with decreasing power at which coastdown begins.
9
Table I cont.
Transient Criteria Assessment of Impact Conclusion
~
15.2.6 Startup of an Inactive Thermal Power < l12%
Case analyzed in the FSAR is FSAR analysis still Reactor Coolant Loop.
RCS pressure < 110%
for two pump operat ion, bounding of design.
presently not allowed by Tech Spec. Lower initial power level will increase margin to thermal power limit 15.2.7 Loss of External Load No fuel damage. RCS Plant runs back to 15% power.
FSAR analysis still and/or Turbine Trip.
pressure < 110% of Lower initial power will bounding design.
result in a faster runback and less added heat. Therefore, peak RCS pressure should be lower.
15.2.8 Loss of Normal No fuel damage. RCS Peak RCS pressure can be FSAR analysis is Fe ed wa t er.
pressure < 110% of controlled by the pressurizer representative.
design.
safety valves to less than 110%
for all allowable power levels.
15.2.9 Loss of all AC Ho fuel'domage. RCS Lower initial power level will FSAR analysis still Power to the Station pressure < 110% of result in less added heat to the bounding.
Auxiliaries.
design.
RCS and lower peak RCS pressure.
15.2.10 Excessive Heat No fuel damage. RCS For reduction in Feedwater FSAR analysis still Removal due to Feedwater pressure < 110% of temperature, MDNBR occurs post applicable.
System Malfunction.
- design, trip.
Initial power level should have little impact on MONBR.
For increase in Feedwater Flow, the analysis was performed at startup.
15.2.11 Excessive Load No analysis.
N/A Increase
Table I cont.
Transient criteria Assessment of Impact Conclusion 15.2.12 Anticipated Variations N/A N/A in the Reactivity of the Reactor 15.2.13 Failure of No analysis N/A Regulating Instrumentation 15.3.1 Loss of Reactor Coolant No analysis N/A from Small Ruptured Pipes or from cracks in Large Pipes which Actuates Emergency Core Cooling.
15.3.2 Unadvertant t.oading N/A N/A of a fuel Assembly into an improper position.
15.4.1 Waste Gas Decay N/A N/A Tank Rupture 15.4.2 Steam Generator Tube Doses < 10 CFR 100 Doses are independent of initial FSAR analysis still Rupture.
No additional loss of power level.
applicable.
reactor coolant boundary integrity.
15.4.3 Gontrol Rod Ejection No additional loss of Sensitivity studies are performed FSAR analysis still Accident.
reactor coolant at Zero power and full power.
applicable.
boundary integrity.
This bounds intermediate power levels. The concern in this transient is local peaking effects-which are independent of initial power level.
Table 1 cont.
Transient criteria Assessment of Impact Conclusion 15.4.4 Steam Line Break Cord shall remain The case analyzed in the FSAR is a FSAR analysis still intact. No SGTR double ended break of a 36" steam applicable.
I induced by the SLB.
line.
For this case, the reactor Doses < 10 CFR 100.
trips almost immediately on low RCS pressure. Thus, a lower initial power level will have no impact.
15.4.5 Break in Doses < 10 CFR 100 A lower initial power level FSAR analysis still instrument lines or lines should have little impact on applicable.
from primary system that isolation time.
penetrate containment.
15.4.6 LOCA Dose.S < 10 CFR 100 Lower initial power level will FSAR analysis bounding.
give less limiting clad temperature.
15.4.7 Fuel Hand)ing N/A N/A Accident
i S) Tatsoo r
t
% Ei3iSTd Docket No. 50-346 LOWELL E. roe Vice Presscent Operating License No. NPF-3 a...o
,n.a.
(4191 259 5242 May 26, 1978 Serial No. 436 Director of Nuclear Reactor Regulations Attention:
Mr. John F. Stolz, Chief Light Water Reactors Branch No.1 Division of Project Management United States Nuc. lear Regulatory Cocmission Washington, D. C.
20555
Dear Hr. Stols:
As a result of our review of the uncertainties in the determination of C-the reactor coolant flow rate at the Psvis-Besse Nuclear Power Station Unit No. 1, we have found that the value for the uncertainty given in our letter to you dated April 24, 1978 (Serial No. 428) is incorrect. Attach-ment I to this letter shows that the actual uncertainty value is 2.2%. is a " Statement Addressing Foreign Material and Deposits in Secondary System."
The uncertainties.on the reactor coolant flow rate are incorporated in the technical specification changes given in the attachment to our letter to you dated May 26, 1978 (Serial No. 439).
Tne reactor coolant flow rate uncertainty also impacts the treatment of the rod bow effect as addressed in our letter to Mr. Roger S. Boyd dated April 10, 1978 (Serial No. 426).
Yours very truly, J/
~
, r Attachment jh c/Il f
MleN THE 70LECO ECISCN CCMPANY EC: SON PLAZA 3CO MADISON AVENUE TCLECO.CHIO 6 2
32-9:.74 00 i~
Attachment 1
to Toledo Edison Company letter Dated May 26, 1978; Serial No; 436 DETERMINATION OF TOTAL RC FLOWRATE AND IT,S ACCURACY FOR DAVIS-BESSE 1 J
s
(
BY ROBERT W. WINKS PRINCIPAL ENGINEER BABC0CK & WILCOX COMPANY LYNCHBURG, VIRGINIA MAY 2S, 1978' O
e 4
M
(
a Sh & lM $h67
- 9. -
i'
32-9:.74 oc r
DETERMINATION OF TOTAL RC FLdWRATE AND ITS ACCURACY FOR' DAVIS-BESSE 1
- e
/
TABLE OF CONTENTS PAGE NUMBER
~
r Introduction 1
{
Sumary
-1 o
Sensor A'ccuracy
~
2
' Heat Balance Data
'S Calculated Total RC Flowrate 11 Error Analysis -
13 f
+
6 p
S E
4 w
,s e
ene OH l
e e
.o c
O
,s
4...
32-9174 CC o
r DETERMINATION OF TOTAL RC FLOWRATE AND ITS ACCURACY FOR DAVIS-BESSE 1 AT 100%
POWER LEVEL INTRODUCTION In a B&W nuclear power plant, Gentile flowmeters are used to measure ' Loop 1
~
and 2 reactor coolant flowrates.
These primary loop flowmeters.are not calibrated prior to installation.
Loop 1 and 2 feedwater flowrates are measured with calibrated flow meters and B&W utilizes a plant heat balance to set the calibration of the primary loop ficwmeters.
An error analysis on the equations used to determine the total reactor core flowrate (Loop 1 plus Looo 2) has revealed that the errors in reactor coolant temperatures and feedwater flowmeter differential pressure are the most significant terms in calculating accurate values of total reactor core flowrata SUM 4ARY
{
The calculated RC flowrate for Davis-Besse 1 at 1007 overish3.2%)timesthe design flowrate of 352,000 gpm.
The accuracy 1 2.2%
This was determined with RC temperature instrument string errors equal to 20.79F and feedeater flowmeter APerrorsequalto[21.25%Mdsteamtemperatureerrdsequalcoi4.2F.
~
R.eactor coolant system temperatures are measured with + 1/4% accurate pre-
. calibrated RTD's over a range of 520 to 620F.
SimilarTy, the Bailey Meter Company differqntial pressure transmitters are calibrated to + 1/2% at time of installation.
The two feedwater flowmeters are calibrated to + 1/2% prior s
g to installation.
For normal everyday conditions in the instrumentation area of the plant, B&W has determined the accuracy of all input measurements used in this error.
analysis.
(Refer to page 4.)
e s
a f
e e
/ - -
o se ei? g f
g
32-9:.74 00 ACCURACY. OF MAJOR INSTRUMENTATION f
USED FOR PLANT HEAT BALANCE CALCULATIONS
./
a
/
D' G
, 6 N@
e LOOP L
' cent LOOPL 3
OTSO OT.5G N
g J
v g
H L
.J e
g x
y g
e e
e.
j l
(
The following RTD's, eacticalibrated to + 1/4 F,and tables prepared and sent with the sensor, were used for plant Feat balance data for calculating total RC flowrate:
b No,.
Description
^
_1 Loop 1 hot leg temp. Narrow Range,.520 to 620F.
3 Loop 1 hot leg temp. Narrow Range, 520 to 620F I
/
Loop 2 hot leg temp. Narrow Range, 520 to 620F 9
Loop 2 Hot leg temo. Narrcw Range, 520 to 620F 5
Loop 1 cold leg temp. Narrow Range, 520 to 620 F 6
Loop 1 cold leg temo. Narrow Range, 520 to 620 F 11 Loop 2 cold leg temp. Narr6w Range, 520 to 620 F 12 Loop 2 cold. leg temp. Narrow Range, 520 to 620 F RTD's 13,14,15 and 16 are wicie range (50 to 650 F) sensors and were not used.
CITD's 2,4,8 and 10 are inputs to the RPS and were not used for heat balance) lC i*
3 l.
e e
32-9:74 m Wv FEEDWATER FLOWRATE MEASUREMENT ACCURACY f.
LuoP t LonP R STEAM STEAM GENE.RATOR GENERATOR O
O Y
N a.
}
}
le a
geo,,
FLOWMETER
.v7
(
l L0oP 2 FLOWMETER
(
-C FROM
=
FEED WATER HEATERS LOOP 1 LOOP Z.
FW TEMPERATURE. '
1:w TEMPEPATURE Each feedwater flowmeter is calibrated with 455 F water (rated flow) and the flow coefficient for each set of taps (two on each flowmeter) is supplied by t
l the vendor.
The required accuracy is + 1/2%.
Each AP transmitter is calibrated to the ennen specified by the measured flow coefficient within an accuracy off 0.25%.)
The accuracias for the different, parameters are shown.in the Table on the next page.
The accuragy for each parameter was conservatively calculated by summing the string errors.
The environmental errors from changes in camperature and the errors from the pomputer were included in the values shown in the Table.
C
- 4 32-9:L74 y ACCURACY OF PRIMARY AND SECONDARY SIDE MEASUREMENTS
~
USED FOR CALCULATION OF TOTAL RC FLOWRATE MEASUREMENT ACCURACY PARAMETER ACCURACY P.
SPAN UNITS
+ 0.79 '
520 to 620F
/ +0.79 F RC hot leg temp.
C29c 3!-3b5 T 4
~
g g fd RPS M ds RC cold leg temp.
1 0.79 /
520 to 620F v 10.79 F Steam temp.
1 0.60 s 0 to 700F, 14.2 F 11.13..
O to 600F 16.8 F Feedwater temp.
. s Feedwater pressure O to 1500 psig 115 psi
(
Steam pressure 11.89%
0 to 1200 psig i 23 psi RCpres$ure 10.77%/
0 to 2500 psig 1 19 psi Feedwater Flow
- "'8 " O to 960 inches
+ 12. inches (Std. H O) 2
+ 1.046 O to 910 inches
+ 9.5 inches RC Flowrate
~
~
(Std.H0)
~
2 52 2ao ns e
y
~
a 8
32-9.74 00 p
COMPARIS0N OF HOT LEG TEMPERATURES (Tg)
FOR DAVIS-BESSE 1 AT 100% POWER ON
~
APRIL 5, 1978 LOOP 1 LOOP LOOP 1 LOOP 1 T 720 ({F)
TIME T719 (OF)
T721 ( F)
T722 ( F) 14:39 605.6 605.8 605.9 606.1 14:44 606.0-606.1 606.2 606.1 14:50 606.0 606.2 606.1 606.2 14:54 605.9.
606.1 606.4 606.4 14:59 605.6 606.1 -
605.9 605.9 15:04 605.8 606.1 606.1 606.1 15:09 605.9 605.9 605.9 606.0 15:14 605.9 606.1 606.2 606.4 Midpoinc 605.8' 606.0 606.15 605.15 Span
+ 0.2
+ 0.2
+ 0.25
+ 0.25 (during minute Average Ta
= 60G.03 F C
~
LOOP 2 LOOP 2 LOOP 2 LOOP 2 TIME T729 ( F)
_T730 (op)
T728 ( F)
T731 ( F) 14:39 604.9' 605.1 604.3 604.9 14:44 605.3 605.3 604.8' 605.0 14:50 605.4 605.4 604.9.
605.2 l
14:54 605.1 605.3 604'.4 605.2 1
14:59 605.1 605.4.-
604.6 604.8 15:04 605.1 605.1 604.4 604.8 15:02 605.0 604.9 604.6 604.7 15:14 605.2 605.4 604.7 605.2 l
l Midpoine 605.15 605.15 604.6 604.95 t
Span 1 0.25 1 0.25
+ 0.3 1 0.25 (during minute Average Ta
= 604.96 F g)T W TO c
^
^
n m
3.2-9:.74- 0:
COMPS tISON OF COLD LEG TEMPERATURES ('n:). FOR DAVIS-BESSE I AT 100% POWER ON APRIL 5, 1978 TIME RCP 1-1 TEMP ( F)
RCP 1-2 TEMP ( F) 14:39 559.2.
558.6 14:44 559.1 558.4 14:50 558.9 558.3 14:54 558.9 558.3 14:59 558.8 558.1 15:04 559.0 558.4 15:09 558.9.
558.3 15:14 559.2' 558.4 Midpoint 559.0 558.35 Span
+ 0.2
+ 0.25 (during 35 minut L6op l Avg. Tc
= 558.7 F s
C TIME RCP 2-1 TEMP (*F)
RCP 2-2 TEMProF) 14:39 559.2 559.1 14:44 558.7 559.4 14:50 558.7 559.3 14:54 558.7 558.7 14:59 558.7 e
558.6 15:04 558.3 558.8 15:09' 558.4 558.8 15:14 559.1 558.9 Midpoint 558.75 559.'O Span
+ 0.45
+ 0.40 (during 35 mi'nute Loop 2 Avg. Tc
= 558.9 F p
6-8 e
32-9:'74 C0 COMPARISON OF STEAM TEMPERATURES (Ts) AND PRESSURES (Ps)
FOR DAVIS-BESSE 1 AT 100% POWER ON APRIL 5,1978 LOOP 1
. LOOP 2 TIME STEAM TEMP ( F)
STEAM TEMP ( F) 14:39 595.6 596.2 14:44 595.6 596.7 14:50 595.5 596.7 14:54 595.5 596,4 14:59 595.4 596.6 15:04 595.5 596.4 15:09 595.5.
596.4 15:14 595.6 596.4 Midpoint 595.5 596.45 Span j; 0.1
+ 0.25 (during 35 minutes s
F LOOP 1 LOOP 2 TIME STEAM PRESS. (osis)
STEAM PRESS. (este) 14:39 905.8 881.2 14:44 907.8 879.4 14:50 905.2 882.0 14:54 905.4 881.1 14:59 -
906.1 885.9 15:04 905.1
.,e 881.1 15:09 903.7 879.9 15:14 905.1 884.7 Midpoint 905.75 882.65 Span
+ 2.05
+ 3.25 (during 35 minutes)
J C:
7-e
-~
x-1174 CC CO EJARISON OF FEEDWATER TEMPERATURES (T ) AND F
PRESSURES (P ) FOR DAVIS-BESSE 1 AT 100% POWER F
g ON APRIL 5, 1978 LOOP 1 FEEDWATER LOOP 2 FEEDWATER TIME TEMP ( F)
TEMP ( F) 14:39 459.8 459.8 14:44 459.9 459.9 14:50 460.1 460.2 14:54 459.8 460.1 14:59 460.2 460.2 15:04 459.9 459.9 15:09 460.0 459.9 15:14 459.9 460.0 Midpoint 460.0 460.0 Span t0.2
$0.2 (during 35 minutes)
Tg = 460.0 F for Loops 1 an4 2 C
LOOP 1 FEEDWATER LOOP 2 FEEDWATER TIME PRESSURE (psig)
PRESSURE (psig) 14:39 942.4 955.9 14:44 942.8 956.1 14:50 944.6 957.2 14:54 943.2 956.7 14:59 944.9 958.8 15:04 943.2 956.1 15:09 943.2 956.7 15:14 943.2 957.0 Midpoint 943.65 957.35 Span 2:1.25 21.45 (during 35 ninutes:
W I-c a
6 32-9:.74 00-r COMPARISON OF MEASURED FEEDWATER FLOWRATES (W )
F FOR DAVIf-BESSE 1 AT 100% POWER ON APRIL 5, 1978 LOOP l FEEDWATER FLOWRATE LOOP 2' FEEDWATER FLOWRATE TIME (MP PH )
(MPPH) 14:39 5.806 5.784 14:44 5.787 5.753 14:50 5.841 5.811 14: 54 5.812 5.780 14:59 5.808 5.787 15:04 5.820 5.814 15:09 5.798 5.762 15:14 5.822 5.785 Average value:
5.812
' i 5.785 Midpoint:
5.814 5.784 C
Span:
1 0.027 1 0.030 (during 35 minutes h
I
.c f
.m l
~
1
'C
_g_
. O
~
PRIMARY SIDE ENTHALPY CALCULATION:
NOMENCLATURE e
HH = Reactor Coolant hot leg enthalpy HC = Reactor Coolant cold leg enthalpy HS = Enthalpy of steam at steam generator outlet HF = Enthalpy of feedwater to the steam generator 6 H = Change in enthalpy, an added subscript: 'pri' or 'see' implies the primary-and secondary loops respectively.
LOOP 1:
HH @ 606.0 F and 2159 psia = 622.41 Btu /lb HC @ 558.7 F and 2220 psia = 558.14 Bru/lb AHpri = 64.27 Btu /lb LOOP 2:
HH @ 605.0 F and 2159 psia = 620.94 Beu/,1b HC @ 558.9 F and 2220 psia = 558.39 Beu/lb AH pri = 62.55 Btu /lb SECONDARY SIDE ENTHALPY CALCULATION:
LOOP 1:
HS @ 595'.5 F and 920 psia = 1254.63 Bru/lb HF G 460 F and 958 psia = 441.73 Bru/lb AHsec = 812.90 Beu/lb l
LOOP 2:
HS @ 596.5 F and 897 psia = 1258.03 Bru/lb HF @ 460 F dnd 972 psia = 441.74 Beu/lb AH 816.29 Beu/lb sec 2 d
i' a
e
32-9:;74. OC
~
e cALCU LATED TOTAL RC FLOK) RATE s_..
I r-pt LWCALIBRATED PRIMARY LOCP PRF.SSURE f"LOWM ETE R O(
O o
D t10T LEG TEMP.
LocP 2
/
Loop 1 2
I N' Dj D
G CCLD LE6 TEMP.
j,
{
PRT.Mimy LeoP' ccMFIGURATlON For h
- eniive, arimag - sec.a,4 og syshm,
W Mi) PRIM R Rca. Md &Qgg,pq NRtt pgg
= 956 + Qp.AdrAT209 Los.s E.s l
h We i = R_e.a.er,r-wt FAo (Re p bp (
N W
e.2 = R-ee.& M*.t %o Ran, p bg 2 R
Qp 3 = He df h m _. %. c s e u P g s 6
C
- ~15.3 y 10 Bra An'
),
r
)
34-9:' 74 CC c
i i
c______L__,
U#
- t.ccP1 l
O CCRE -
l OTSG.
I I
I_ _ _.,_ _,_ 7 _ J I
I I
I i
~
I SECOND ARY L O OP. tom FIG U RATIO N
~
Qge::(Np A H se.c ),p +(Wj= A'Hsec) Loce 2.
6 RADIATT0td =.
O 54 xio BM fw l
Lo sses SuJostZh ARL oJoOut bh waa (AH )g g,% bbggy engg i
6
--- (Wp6H ec)ggpp Wsec)toer2.* *P h 3
" W a (6hdPuMAny + W c2.(bH.)pp_.,uAgy =hFO c h ooP2 S
2 R
R C
c
-,qqswo su wF 6 Loort w
O
32-9.~.74 00 r
b 1
- 74 76Xlo as r~
cv N ei(6H )\\
-hwF yN.ac /toopi
~L 2.
R t FFatARy i
- 1T7'XI gg__
w Wgez ChA)peAgy =(WAHyc),g2 F
5olvinsfw hlpg4 amMWe2. :
6' (Wp AHs<c) coop, - 3P38 Xio 8h,W W
//1 H T
(
11 PRIMARY caut (WAH ec)me2 - 37'3# Xl0 Ek P
3 Wg1 I
/AH2 I
(.
JPP2 MAP-y io l
b r~ q 4 (bhdfgi 64 2T Bh %
="
6.'2 s r E r w a (6H2) Par (6W ec)eep,
= B12 30 Bh11b 3
816 '1s 3Nb
~
[6Hoec)t. cop 2.
=.
(
m Ms d',hTi trd.
cueo t % Ucetu.e.o g n.,
6 l(o ;,y.
W
=
f ' S l 2. X 10
(--
FLeePI 6
A ha T
we ep2 5 7&s mo Is
)d.- 4.L (4 UU g
c o 2'u" 5'0l2XlO
'00 Q
=
I
~
t (64-a7) s
~ 1 93 XIo Ib Iw
=
N e'2. _ f6 785Xio (816 23) - 37 '32 X lo' 3% h~
R (64 ss) e
'/9 90 )qo C
=
16 9 C#at Re p ><xt = vJp.e 7. = IAr
+kVac.2 act 6
f 7'83 Xto
\\b l br/ w l@M m'
=
% w qq p n x.t..a c pn m(> %JwhPL a.o b cT C4 k) t R
Wpyr C[")
=
g3.gos x Tcw w.
\\
c p
J cMh c, q s.16 /J.j g3 af_.sgg,3 p g
(
0 2..n o psia.
RtT (g =1
=
3.9 84 Xio' d'*
- lN i
t'
& lg f0, W
= Hb W th J
C U P
=
3 52, 000 {"
D dodA'% -
M O; g
o
= \\\\3.Q
32-9E74 OC t
ERROR ANALYSIS OF CALCULATED RC FLOWRATE f
The basic equation for Loop 1 or 2 RC flowrate is:
f
~
RC F.
- sec, where C = p%j 6
Q
~ 9' rad. losses = 74.76 x 10 tu/hr p ups w (Io If one assumes ac extreme case of a O10%rgorinC,thenCwouldbeequalto 7 Btu /hr rather than 74.76 x 10 Btu /hr.
8.2236 x 10 Then total RC Flowrate would become:
6 pet S-811X 106 W
=-
g g lq. So - 3T3g y j o feb4lW G
g G4 27 Ib14w 5 785 Xl X Ts 16
- 29 - 3T37 Xlo' Bh &
Wgez c
=
'MMO 6 Q Ss Ab{4vy M R2 y Mtt. = \\H] 1x[h L6 &
M %
Change in total RC flowrate is g, l 3 6
For this small change we shall assume C is a constant and is 74.76 x 10 Btu /hr.
If it is in error, its influence on the final value is insignificant.
G
=
a m
d l
l e
l
a.j:.
32-9174 0o r'
ERROR ANALYSIS OF CALCULATED RC FLOWS
. Tor either Loop A or B the ecuation for RC flow is:
Wgc =
VV (I4-ggli F
3 Note:
Introducing. a coE
( g pc.}
stant in the ex-pression will no' T
T change the follot Cje.oF p (Ts, P j-p (/,=, Pgj T
ing worx.
3
\\
i.
- p (Tu; O-p(L PJ 4
N,
- Ngg Y Wge g,c Determinetheexpressionfortheerrorinh for a normal distribution of small' errors in -iach of the beasured parameters R
C
~~~'"~~~"
Where[isafunctionofT and P,; an
~
E=g/g *'7. h f C, the flow coefficient is assumea to 1.
y
- * ~ ~ ~ -
be a,,const, ant.
._..'._dW[= /pwp [ [gw;d P
P
' : ~
}(
-) \\
)
~
MP h d T= + 4-J Pw,.
Bw1 3p
- ..... d a =
'~
t
-=
u bY d~ p F
DP y
) Wp C
AP'
) \\dy
=
c, 9
- d. P Z
{
D4.P 2
\\
C
~
16
32-9:.' 74 OC
~ '
The value for each enthalpy in the first equation given on the previous page was determined from temperature and pressure measurements.
The value of W.,
w s obtained from the feedwater temperature, and from feedwater differential pressure measurements.
In the first equation given on the previous page, the only parameters that depend on a coc=on measurement are H.,and W.
Specifically, H.,and W both depend on T.
Consequently,intheerroranalfsis,acorr)1ation# term 15 daveloped to account p
fsr the dependence of Hp and W on T.
y 7
The correlation term is:
h YCC.
YF
~
_[fE o
AC y
b h ?. F...... y,- y,'? c-).-fr--
\\.'N.E.
50 r-...,..
S - NF A
' - -g uj, -- s p, - ge l-Yk ff ?
l.
A P 'Yd p.-
0.WQ.fc]. WS d-T, q
eryekD4r')
y
eq)]'TeJ-kidgg.
- ^
J
=
wp-c.gpse 7
, f NS-NF
=
\\
By-He NS ~ $2
~~b._Yi'Y f.h$f \\
l--/
)./f E
. =,l.g...yq-3 7-Ha - ke
- (_
"e-He.
. Y t.
Y ~'
t
(
The correlarion term becomes:
l E
0
-rJ/g..Hef.-.b- (-.k.) 12.)\\._fHe-He.g. 2 7e.)._
~
\\..J.
. z -
x.- ch..R.fHs-M.k) dPh.dNF T,.
(He - He Tu 9 te. j) Te J - - -- -- - ---
~
- 17
S 9~174 m uu T
For the Bailey Meter Company - supplied feedwater flowmeters which are calibrated prior to shipment to the site, we have the following typical equation:
b C-- -f---
=
F-d !
- =:A.'Ci)?.CsCdEsi d, *. cj.._._.R.
0 = _ A..d A = conversica factor approach factor (dependent on the beta ratio)
C e
Fa = cffect of thermal expansion of flowmeter throat diameter d
measured throat diamter - inches
=
o and C is the flow coefficient derived from the calibration d
data for each set of pressure taps.
of 1
bY O
,- A 0..., A.
+
f 2
7 2
2 5& C -l. g.4S, + 0.dl_ PgA.fa_,.
f i
2 C
3 Typically d
= 9.121 inches measured to + 0.001 inches
/do g
0.0ll%,wiiich is negligible thus 6d Also, ar or near full power flow, the test data values of C are 0.9909,,
d 0.9916, o.9928 and 0.9878.
The uncertainty li; the above values is negligible
. as compared to other uncertainties.
Based on the above description, C will be assumed to be constant for this error analysis.
Theexpressio,nfor pmay then be written as:
ty Qg p..
[
2, sP -
-}
F-2 y y.
" " ~..
--.~-
y p y Jf.
gg)F
-f'2 C:- A R S P 3 4-f 2-E" P --Q z;p s P --- ---
Since T determines feedwater enthalpy as well as feedwater F
- flowrate, an error in T.
can result in non-(
independent errors in n a n d" -W y
p, i
e 18..
bd.-417% CC Contin ui ng wi th the develoomen t o,f,,d_W I
RC Yx. = f X, by 2
l"
"' d.H, = I~(2 TxQ d.y;.' G /~ ~/h _,p 'j _
l 2.s u <
8C k'-~-{--S~
~
a general expression is
(-gg.//,_]
developed for the accuracy of either loop 1 or 2 RC 'flowrate y
7>9ur,G),fe)idac-g.g
- 4. fc) 10ec..gll j k.Ugg =
d Yac.)
7--- }-2-y,
- 9 y,
.z.-
- j..
y z. _.
..YAC.
g
.f U2C.}
9 y,,
g.gc..-.}.....}..-
6 4
-i?AP (//s-Up)~_ eTf ~&Hp g.y,.?~ ~g
^
~
^
(-Hir-Hi)* - D -Teni) Tr.-j 3
Calculate coefficient C for the feedwater flowmeter:
- W -
[ lbr
- C l2
,.AA inCSet-r/d.oebr-p sur
(
From Bailey Meter Co. Calibration sheets 7 0 xto % s/hr C=
7
=
o.3154 x /0 T SI.4*)*95G*S 3
[p at 455 F and 1065 psia - 51.49 lbs/f t and AP (full scale)= 956.5 inches of standard water.
Calcu, late typical value of aP at full power:
3p (Wpf J. _
- 5. 812. sto
- [_ L 6 62. iru. ass
=
\\T/f
' O.0 61s y no g 5 1. 1 4.
3 gat 460Fand1050 psia =,51.26lbs/ft r-
- 0
q,...
.3d-9:.74 oo Calculate Hg-Hp (HHHC)2 Use Loop 2 values to obtain 816.29 (62.55)2
= 0.2086.
Evaluate 39 for variation around 460 F:
3T p 3
g at 457.4 F = 51.335 lbs/ft 3
jo at 462.2 F = 51.099 lbs/ft 3B 0.236
- 0.0492 lb
=
=
-4.8 ft"-F p
Evaluate 3 H-for variation around 460 F:
~
BT '
p h
a t 462F = 443. 99 Beu/lb p
h at 458F = 439.48 Btu /lb p
C
-3 H p 1.128 BTU /lb-F
=
ST p Using the above values, the complete correlation term (P) may be calculated as:
0 3
P = -(3.154) 2 -x 10 x 0.C62 x 10 x 0.2086 :: (-4.92 x 10-2)
(1.128 (dT,) 2)
Or 9
P = 7.624 x 10 x (dT )2 g
.em l
C
=
- 20 i
32-9:.74 OC l
Evaluate the following terms:
AN IN I R C IN 3N RC RC R C_,
RC 3W 2H 2H 2H N
p 3
p H
C 3W 3
W (H3-H)
NS-NF RC p
F
=
A 3W Stl l-Hg.- HC HH-HC p
p H'
W 2W 3
N NS-NF p
p B
RC F
3H 3H Hg-HC HH-HC S
3 2W 3
W Hg-Wp p
p H
-W RC F
_=-B 2H 2H HH,HC J-HH-NC p
p i,
(
3W 3
NF (H3 - H )'
H)
-N
-W (H3-RC,
p p
E RC
=
-0
=
3H IN (Hg-H)
INH-N)
HH-HC H
H C
C H
RC ',
3 NF (H3
.H )
+W
&W p
RC
=0 3H 3H (H
-H)
NH-HC C
C g
C e
G p
e e
M e
4
~ '
a l
L... ::.
32-9:' 74 C0
'A = 812.90
+
816.29 64.27 62.55 12.85
=
r-B = 5.812 X 106 + 5.785 X 10 6 4
2 9.146 X 10
__ l b
=
64.27 62.55 BTU-br 2
D = 73.18 X 100 = 75.16 X 106 6
2 1.170 X 10 lb
=
64.27 62.55 BTU-hr 2
Rewriting the expression for d WRC 2 (d W )2,3 (dH )2,g (dH )2+D (dH )2+0 (dH )
2 2
2 2
dWRC =
A
+P p
3 p
g c
oA 2
h dWRC
- dT
+A daP p
p 2
2 z
2 8 < aH 2 '. 82 eaH 3
+ e rah, 2+82 rah, 3
s s
(W e1 ee s
s,
- t21, e1, l3 ees 5
1ae
+D2 g33 drg2. g g gg 2
9 H
dP
+O2ggg dT
+D 2H 2
g H
C C
C dP
+
\\W J
kW
{x j
W c
.... n. 3, _. p. _ e.
.e 1e.
%. /
e O
em d
F -
i e
A a
/.
o
- 24-%.74 QC
<l e
Evaluate all the remaining coefficients:
N
.f_.
=
661 x [- O.0492) f bTp 2.
St. 2./.
- 27 8 8. Ibs/gr, _.g
=
15770I 4 3 88 lbs/r. -
s
=
es M.
r p
IN
~
K (around 596F and,910 osia)
S 0.835 Bru/lb o
=
592F to 600F 3"S (around 596F and 910 psia)
-0.112 Beu/lb - F
=
3P 890 to 930 psia 3
IN
^
K (around 460F and 960 psia) 1.126 stu/lb - F F-
=
453 to 467 F s
C 3H K
(around 460F and 960 psia)
F
= - 0.00042 Beu/lb - psi 900 to 1020 psia 2H (around 605.5F and 2160 psia) 1.47 acu/lb - F H
=
ST 604.5 to 606.5F H
3H (around 605.5 F and 2160 psia) = ' b.0056 stu/lb - psi H
2P 2140 to 2180 psia g
ANC (around 558.8F and 2220 psia) 1.260 stu/lb - F
=
ST 557.8 to 559.8 F C
3H
= - 0.0020 stu/lb - psi K
(around 558.8F and 2220 psia)
C 2200 to 2T40 psia Substituting each of the calculated coefficients into the equation for d WRC we obtain:
p.
'. ^ *.... '..
W $LfS NU dW (12.85)2(-2780)2(dT)2+(12.85)2(4388)2(daP)2 RC p
(91460)2 (0.835)2 (dT )2 + (91460)2 (-0.112)2 (dPs)2 e
3
+ (91460)2 (1.125)2 (dT )2 + (91460)2(.00042)2 (dP )2 p
p
+ (1.170X10 )2 (3,47)2(dT )2 (3,37gx3g )2(_g,gg33)2(de )2 6
6 g
g
+ (1.170.X10 )2(1.26)2(dT )
+(1.170x10 ) (-0.0020)2(dP )
6 C
C
+ 7.624 X 109 (dT )2 h
p dWRC =
3.179 X 109 (daP)2 + 2.953 X 1012 (dT )2 + 2.173 X 1012 (dT )
g C
+ 1.952 X 1010 (dT )2 + 0.764 X 1010 (dT )2 + 0. 429 X 108 (dP )2 p
3 g
+ 0.055 X 108 (dP )
+ 1.049 X 108 (dP )2 +.148 X 104 (dP ) '
C 3
F hap)2 = ( 12.) =1h (d T )2 = (0.79)2 = 0.62 g
s (d T ) = (0.79)2 = 0.62 C
[
(d T )2 = (6.8)2', 4g, p
(d T )2 = (4.2)2 = 18.
3 (d P )
- (I9)
" 30I
~
H (d P )
= (19
= 361.
C (d P )2 = (23)
= 529 g
hPg)2 = (15 )2 = 22 b78X10b+
h l
for (daP) dWRC =
1.834 X 1012 +
go,(d T )H-1.347 X 1012 +
for (d T )
C 0.898 X 1012 +
for (d T )
~
p 0.138 X 10_2 +
. for (d T )
1 3
1.549 X 1010,,
for (d P )
g 9
1.986 X 10
+
for(d P )
C 0.555 X 10Il +
for (d P )
3
[
forCd P )
F
- 24
3...
e X-1.L( W e.
6 d h*g b
474 0
x 10 lbs/ hour
=
or for either loop.
Since the flowrates for the two RC loops were calculated from measurements taken with two completely different sets of sensors, the total RC flowrate percentage error from the heat balance measure =ents is:
1 0.707 (2.95) {
As shown on p.4, the RC flowrate string error for the Gentile flowrates is 1.046%.
After the RC flowmeter 4 P transmitters are calibrated against the results obtained from the heat balance determination, then the total fractional error in each loop flowrate will be:
. v.
=
(.0295)
+ (.010k6)2
=
.000980 or. =
Tha total RC flowrate is simply the sum, of both loop flowrates,but the extremes of crroneous signals will not occur simultaneously; there{ ore, we can say: that the parcentage error in the total RC flowrate from the Gentile flowrate measurements, ca calibrated by the heat balance results,. is (
3.1%)
Q,
,F
(
Ik 1 %
= f% ' a w,)$ 3 % A w [
w L
A,
2% -
t Pw.
3%
lhL i
- - d% '
(Aw,f+(dog sa,kI Aw, i de.
'h.
m s
1 1
(us%Q + (.r.6ed gg,,
- 0) w, u g, A
- 2.. %,i w
ev kf A A
^
L:
k%
+.,. =
l i
2.
a
_ fW
_'1(2.ss't.ir ae G.sf y,
m;,,
z g
1 I
=
TOLEDO EDISON COMPANY LETIER DATED May 26, 1978; Sarici No. 436 32-9:.74 oc c
SfATEMENT ADDRESSING FOREIGN MATERIAL AND DEPOSITS IN SECONDARY SYSTEM The OTSG Feedviater Chemistry control is designed to minimize the ingress of contaminants to the units.
These contaminants include both suspended and dissolved solids.
Chemistry control utilizes the all-volatile chemicals ' ammonia and hydrazine which will not deposit or form insoluble solids which could deposit on critical surfaces. ' The use of these chemicals is designed
(
to minimize the corrosion of feedwater train materials and thus the input of corrosion product oxides into the steam generators.
Water purity is further maintained through the use of full flow condensate polishers (powdered resin) which, in addition to re-moving disso1ved solids, also provide., excellent filters.
These filters a a located such that they process all the water coming 1
from the condenser.
As a result, damage to orifices in the feed-water train is hignly improbable due to the aforementioned feed-water chemistry and puaity control.
Shot blasting for the feedwater piping was done in the sh2p before installation l
of the piping.
Thus7 no shot blasting was done with the feedwater flow elements 1
installed.
l 1
a e
m. - 7[
k