ML20031F366
| ML20031F366 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/30/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
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| References | |
| NUDOCS 8110190645 | |
| Download: ML20031F366 (22) | |
Text
.
Responses to Second Round Questions on the Statistical Combination of Uncertainties Program:
(CEN-124(B)-NP,Part 1, Part 3)
- Part 2 -
September 1981 COMBU3TI0ii E:1GIliEERIliG, If1C.
8110190645 811008
- PDR ADOCK 05000317
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LEGAL NOTICE l
THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED r
BY COMBUSTION ENGINEERING, INC. NEITHEli COMBUSTION ENGINEERING 1
NOR ANY PERSON AC11NG ON ITS DEHALF:
A.
MAKES ANY WARRANTY OR REPRESENTATION, EXPilESS OR IMPLIED INCLUDING TifE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH HESPECT TO.THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS OR D. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METilOD OR PROCESS DISCLOSED IN THIS ItEPORT.
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Abstract This report contains the answer to the last question on the Baltimore Statistical Combination of Uncertainties (Gas and Electtic(BGEE)Battelle Pacific fiorthwest Laboratories (BPfil).
SCU) program asked by The answer to this question provides the measured data used for the uncertainty evaluations of monitored parameters.
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e ii
Table of Contents,
Legal fiotice i
Abstract ii Table of Contents iii Title 1
Introduction 2
Question One 3
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I I
ontheSISonsetoSecondRoundQuestions R
tistical Combination of Uncertainties Program (CEN-124(Byf1P,Part 1, and Part 3) e W
- Part 2 -
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Introduction In a recent telecon between NRC and SGSE, the NRC reiterated its interest in the answer to the last outstanding question on the Statistical Combination of Uncertainties (SCU) program asked by its contracted technical reviewers Battelle Pacific Northwest Laboratories (BPNL).
e Ttis question had originally arisen daring a C-E, BPNL, NRC conference on the SCU reports at Richland, Washington.
The answer to the question supplies information about the measured
^
data used to justify the uncertainties for the monitored parameters included in the stochastic simulation analysis.
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puestion_i_
Provide measured data used for uncertainty evaluations on monitored parameters such as flow, tenperature, pressure and power. This is needed to assess the values given in Table 3-1 of Part 1 of the CEri 124(B)-P report.
]
Response
f The plant parameters directly monitored by the TM/LP trip system and the local power density trip system (Part 1 of Reference 1) and by the DilB LC0 and Linear Heat Rate LC0 monitoring systems (Part 3 of Reference 1) include:
Axial Shape Index 3-D Core Power Peak Hot and Cold Leg Primary Coolant Temperatures Pressurizer Pressure Core Thermal Power The evaluation of acceptable core performance, based'on these continually monitored paraneters, is predicated on periodic deteminations that other plant parameters are within specified bounds.
Some of these latter parameters are evaluated on c monthly basis, others less often.
Those monitored monthly include:
Planar Radial Peaking Factor 4
~
Integrated Radial Peaking Factor Azimuthal Tilt Those monitored less often include Core Flow Rate, which is evaluated E only at the beginning of each cycle.*
The uncertainties of these monitored and peridocially evaluated para.
meters were given in Tables 3-1 of Part I and 3-1 of Part 3.
As submitted,these tables include some errors. Tables 3-1A and 3-1B (attached) are the corrected versions of Tables 3-1 of Part I and 3-1 of Part 3 of BG1E's report.
The bases of the Axial Shape Index uncertainties were described in Appendix A of the report. However, the applicability of the reported shape annealing factor component of the shape index uncertainty to St. Lucie I has been questioned (2, 3, 4). This i.s discussed in a separate submi,tal.
Tle basis of the uncertainties in the power peaking factors Fx F4 is given in C-E's approved uncertainty topical (Reference 5)y, Fr.
- Changes in-core flow rate are monitored continuously via pressure drop measure-ments for the Low Flow Trip. This infomation is not incorporated into the Trip or LC0 systems iffected by the SCU methodology,
i The primary coolant temperature, the pressurizer pressure and the core l
power uncertainties are directly incorporated into the SCUmethodology via the s+ochastic simulation. They are also indirectly incorporated into the SCU methodology because they are components of the core fler uncertainty. Their involvement in the latter uncertainty will be discussed below.
Evaluation of the cold and hot leg primary coolant temperature neasurement uncertainties is based on analysis of the circuits of Figures 1-1 and 1-2, rerpectively. A breakdown of the components of these uncertainties is given in Tablel-l. The net uncertainty in temperature has a one standard deviation value of E 3and is assumedtobe[
- 3. The uncertainty in these temperatures due to the reactor protective system circuitry is
, N handled separately, in the electronic processing system uncertainty.
Analysis of the pressurizer pressure measurement is based on the circuit of Figure 1-3. A breakdown of its uncertainty components is given in Table 1-2. Because some of the components of this uncertainty -
are included in the electronic processing system uncertainty, this uncertainty has a one standard deviatj,on value of[
] psi and is assumed to be[
J.
The uncertainty in core power is derived from consioeration of the calnrimetric relationships for one coolant loop outlined in Figure; l-4A and 1-4B and Equation 1.
OLD + OCCll + L (Eq.1-1 )
Qcore " O + OBD ~ OFW - OPZR - O ~ OCH V
P where:
Q
= core thermal power to one steam generator, core Qy
= heat output due to steam flow ($y),
Q
= heat output due to blowdown (f1BD)'
BD Q
= heat input due to feedwater flow (k
),
g g
QPZR = heat input due to pressurizer heaters, Q
= heat input due to reactor coolant pumps,
p I
I Q
= heat input due to charging pumps (f1CH)'
Ctl Q
= heat input due to letdown flow di
)'
LD t.D Q
= heat output due to component cooling water, CCl!
L
= heat output due to reactor ' coolant system losses.
9,-
4 Because the charging flow rate and the letdown flow ra.tes f_
],
respectively, are neglibible compared to the full steam flow rate their heat flow can be eliminated from the heat balance equation.
Furthermore, since the calerinetric heat balance is determined for steady. state operation, it is reasonable to assume that there is no blowdown flow.
Because each of the remaining components of the heat balance equation are independent of each other, the calorimetric power uncertainty is given by Equation 1-2, where Psec is the secondary system pressure and TF is the temperature of the feedwater, 80 O
( core a g )2, f core A.TFW)2 + [ core a
sec)2 q
p FW 3
core 3 P
\\ EM T
FW pg sec
+fcore OPZR)
P} *(a
)
A PZR P
flE nQ CCW)2 core a g
o gW j
CC Table 1-3 lists the 2 o value of each uncertainty component for one coolant loop.
Since the core has two independent coolant loop _,
s the net calorimetric power uncertainty has a 2d value of
] power.
The reactor vessel coolant mass flow rate is determined from calorimetric data using the relationship.
O core WVESS =
(Eq. 1-3)
BULK BULK h
-h HOT COLD where:
W is the vessel mass' flow rate, lbm/hr, VESS Q
is the cora power determined from secondary side measurements, CORE primary side heat losses, and primary coolant pump input, BTU /hr, h
is the bulk enthalpy of the hot leg flow, BTU /lbm h
is.the bulk enthalpy of the cold leg flow,' BTU /lbm The bulk enthalpies are determined from the relationships, BULK _ f (THOT, PPRESS)
" HOT -
AVG BULK AVG hCOLD = f (TCOLD, PPRESS) 4 where:
T is the average hot leg temperature, F,
T is the average cold leg temperature, F,
D P
is the pressurizer pressure, psia.
PRESS The uncertainties associated with the parameters used in Equatinn 1-3 to detemine flow rate are given in Table 1-4.
The listed uncertainties include equipment tolerances, environmental effects, and instrumentation drift.
The aver Je cold'and hot leg temperatures are determined by taking the
. average of the measured RTD temperatures in the cuid and hot legs, respectively.
Since the ccid leg temperatures are measured just downstream of the primary coolant pumps, the average of the ;old leg RTD temperatures gives a good representation of the cold leg bulk coolant temperature.
Since there may be a difference between the hot leg average measured temperature and bulk temperature due to. temperature stratification effects in the upper plenum and hot leg regions, an additionzi uncertainty is. included on the hot leg temperature, as indicated.in Table
.1-4.
In this " application, a very conservative value of the nressurizer pressure uncertainty has been included for consistency with prior usage. The value includes the electronic processing. uncertainties incorporated into the stochastic simulation procedure separately.
Because of the small sensitivity of the flow uncertainty to pressurizer pressure uncertainty, this does not affect the total uncertainty significantly.
The uncertainties are combined using the[
,].
The distribution.of-the uncertainty for each variable in Table 1-4 is taken to be[
}. The resulting total flow measurement uncer-taint
}atthe2alevel.
This value supports the allowance of[ y is[]specified for the coolant flow measurement uncertainty.
\\
l References j
1.
CEN-124(B)P, " Statistical Combination of Uncertainties," Part 1 December 1979, Part 2. January 1980, Part 3, March 1980.
- 2.. G. M. Hesson (BPNL) to H. Balukjian (NP.C), Letter of April 9,1981,-
Question 9 3.
PNL Meeting held on April 14-15, 1981, Question 11 4.
Draft of SER on BG&E SCU Report.
5.
A. Jonsson, W. B. Terney, M. W. Crrop, " Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed In-Core Detector System," CENPD-153-P, Rev. 1-P-A, May 1980.
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Figure 1-1 COLD LEG TEMPERATURE POWER SUPPLY 50 51 l
+.01%
RTD TRANSMITTER
/
/
0-I RPS 2502 3\\
CIRCUITS
+. 01%
l I
i
/
l 50 D.
_.01%
+
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- 8'-
.. - ~..
Figure 1 -2
~
HOT LEG TEMPERATURE t
POWER SUPPLY 59 9 t
/
- ,01%
+
C y*g n
/
o-RTD TRANSMITTER 468.75 n _\\
t o-0,-
"I 1 01%
cI
/
A-l 7
50 R I
POWER
+. 01%
l.
SUPPLY 50.Q l
I
.01%
c
/
g RTD TRANSMITTER 468.75 R Oi D-
~!
i 4 01 %
C g_
/
TORPS
~
50 R CIRCUITRY e
- 4. 01 %
t
~9.
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Figure L-3 PRESSURIZER PRESSURE POWER SUPPLY INDICATOR
~-
2500
+. 01%
250%
J
.)
~
/;
I, _
TRIP
-i.017e 50 R BISTABLE
+. 01%
RPS TRANSMITTER CIRCUITRY
' 250 9 (
TRIP 101%\\
BISTA BLE 50R III 50n 101%
-+ 0,1%
CONVERTER
/
/
0 4
.9
Coolant Flow for One Loop of Nuclear Steam Supply System Component Cooling Water fu 9
L Coolant Pump Feedwater
'jt Blow-Core Generatv-
'down j
Flow Flo (v)
Let down Coolant Line Pump j
Component Cooling Water Figure 1-48, Heat Flow for One Loop of NSSS
~
^
OCCW QP QCH p.
~
v v
(
QFW Ocore (h Q reactor)
)
gBD i.
L L
OPzR
.v LD e
Hea't Balance Equation for 1 Loop of NSSS 4 OLD + l OV+O.BD + OCCW QP + QCil + OPZR Q
+ QFW
+
core
~OPZR +
+OCW Oy - Qpy-Q-QCil ' OBD
- OLD Q
=
p enre '
TABLE 3-~1A 1
UNCERTAINTIES A550 CIA?lD WITil THE LOCAL POWER DENSITY LSSS AND THE IM/LP LSSS 1
Uncertainty
+2
+2 Primary coolant mass flow (% measured flow)
NA Primary coolant pressure (psid)
NA Cure -coolant inlet temperature ( F) iia
^^
Power distribution (peaking factor, % power) 7-6 Axial Shape Index 1.
Separability (asiu)
See Table 1 of Appendir. Al 2.
Calibration (asiu) 3.
Shape Annealing (asiu)
Monitoring system processing (asiu)(20)
[
]
4.
5.
Monitoring system processing (psia)(2c)
Notes:
- For complete' description of'these uncertainties, see Appendix A.
- [
] values s
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t l
12
-~.
TAELE 3-1 B
\\
UtlCERTAINTIES ASSOCIATED WITH THE Dr1B AND LHR LCO'S Uncertainty
- LHR LCO Of!D LCO Core power (% of rated power)
+2
+2 Primary coolant mass flow (% measured flow)
NA
[
]
l
{
' ') **
Primary coolant pressure (psia)
!M Core coolant inlet temperature (*F)
NA
+2 Power distribution (peaking factor, % power) 7 6
Axial Shape Index (Excore Detector System) 1.
Separability (asiu)
See Table A-1 of Appendix-Al 2.
Calibration (asiu)
[
][
]
3.
Shape Annealing (asiu)
{
]
[
]
4.
Monitoring system processing (asiu)(2o)
[
]
[
]
Axial Shape Index ( Incore Detector System) (asiu)
[
]
Notes:
- For complete' description of these uncertainties, see Appendix A.
- [
] values O
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e COMP 0NENTS OF PRIMARY COOLANT SYSTU: TEM.iRATURE UNCERTAINTY REFERENCE ACCURACY 1.
RID (Sensor Error at 620 F) 2.
Transmitter I
(Transmitter Load Error)
(Po:ver Supply Variation)
Total a
ENVIRONMENTAL EFFECTS 1.
RTD (Frictional Heating)
(Stem Conduction) 2.
Transmitter (Ambient Temp. Change)
Total y
DRIFT IN RTD (Ma::imum Change in 1 Year)
- FIELD CALIBRATION
~
~
TOTAL PRIMARY COOLANT SYSTEM TEMPERATURE UNCERTAINTY
~
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4 i
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Table 1-2 COMPONENTS OF PRESSURIZER PRESSURE UNCERTAINTY Reference Accuracy e
l.
Sensor psi
~
2.
Current-To-Current Converter psi Total psi DRIFT psi CALIBP.ATION Digital Voltmeter Error
{
psi
~
SETPOINT SETTI!S Bistable Uncertainty
[
)si TOTAL PRESSURIZER PRESSURE UNCERTAINTY
[
.}si -
/
4
- These components are
-included in the electronic processing uncertainty..
See Appendix A of Reference 1.
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Table 1-:3 C!PONENTS OF I
CALORIMETRIC POWER UNCERTAINTY Magnitudes of Equivalent Power j.
Uncertainty Component Parametgyncertainty Uncertainty (% Core Power)
One Loop Feedwater Flow Rate Feedwater Temperature f
+ -
Secondary System Pressure I
s_ '
Pressurizer Heaters Reactor Coolant system Losses Coolant Pump Heat Component Cooling Water
~
}
Total TwoLoopTotalb S
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TABLE l-4 UNCERTAINTY COMPONENTS FOR VESSEL FLOW MEASUREMENT - SECONDARY SIDE CALORIMETRIC TECHN1QUE Parameter Magnitude of Parameter Fouivalent Uncertainty Uncertainty (2c)
% of Indicated Vessel Flow Rate
~
~
- l. Pressurizer Pressure
- 2. Temperature AVE A. T COLD AVE B. T HOT 3.
Core Power 12% of rated power
~
4.
Hot leg Temperature
[
]
Stratification Effect
- Listed uncertainty is for the average of eight RTD signals; the uncertainty on an individual RTD. signal is{
}.
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