ML20100M323
| ML20100M323 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/31/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20100M308 | List: |
| References | |
| FOIA-84-105 80NED279, NEDO-24272, NUDOCS 8412120275 | |
| Download: ML20100M323 (27) | |
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.NEDO-24272 0 80NED280 [C
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July 1980 '.
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DUANE ARNOLD ENERGY CENTER j 9
SINGLE-LOOT OPERATION .
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i NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORN1 A 95125 l
GENERAL $ ELECTRIC '
2 NEDO-24272 i
Ih?ORTANT NOTICE REGARDING -
CCNTENTS OF THIS REPORT PLEASE READ CAREFULLY nis report was prepared by General Electric solely for Iowa Electric Light and Power Company (IELP) for IELP's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending IELP's operating license of the Duane Arnold Energy Center Unit 1. He information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or pro-vided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract between Iowa Electric Light and Power Company and General Electric Company for nuclear. fuel and related services for the nuclear system for Duane Arnold Energy Center Unit 1, dated Februa'r y 8, 1968, and nothing contained in this document shall be construed as changing said contract. he use of this information except as defined by sai.d contract, or for any purpose other than that for which it is intended, is not authorized;
. and with respect to any such unauthorized use, neittier General Electric Company H
nor any of the contributors to this document makes any representation or warranty (express or implied) as to 'the completeness, accuracy or usefulness of the infor-
,, mation contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for lia-bility or damage of any kind which may result from such use of such information.
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, NEDO-24272 , ,
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g TABLE OF CONTENTS h!
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Page M lJ
- 1. INTRODUCTION AND
SUMMARY
1- 1 !]
- 2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2- 1 M 2.1 Core Flow Uncertainty 2- 1 !1 2.1.1 Core Flow Measurement During Single Loop Operation jj 2- 1 2.1.2 Core Flow Uncertainty Analysis 2- 2 ,e 2.2 -TIP Reading Uncertainty 2- 4 5
?
p
- 3. MCPR OPERATING LIMIT 3- 1 'ra 3.1 Core-Wide Transients 3- 1 9 3.2 Rod Withdrawal Error 3- 2 [!
3.3 MCPR Operating Limit 3 .4 [
- 4. STABILITY ANALYSIS 4- 1 j q
- 5. ACCIDENT ANALYSES 5- 1 J 5.1 Loss-of-Coolant Accident Analysis 5- 1 '1 5.1.1 Break Spectrum Analysis 5- 1 [j 5.1.2 Single-Loop MAPLHGR Determination 5- 2 ig 5.1.1 Small Break Peak Cladding Temperature 5- 3
~i4
~
5.2 One-Pump Seizure Accident 5- 3
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- 6. REFERENCES . 6- 1 j
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l_ ILLUSTRATIONS h.
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- .. t' Figure Title Page -
2-1 Illustration of . Single Recirculation Loop Operation Flows . 2-5 1 Main Turbine Trip With Bypass Manual Flow Control 3-5 h 4 Decay Ratio Versus Power Curve for Two-Loop and Single-j Loop Operation 4-2 5-1 Duane Arnold Suction Break Spectrum Reflood Times 5-6
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. 5-2 Duane Arnold Suction Break Spectrum Uncovered Times 5-7 1
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i NEDO-24272 -
k
. TABLES l] .
Tcble Title Page i l'
5-1 Limiting MAPLHCd Reduction Factors 5-5
.1
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,. NEDO-24272 j 1. INTRODUCIION AND
SUMMARY
I g The current technical specifications for Duane Arnold Energy Center do not j allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service. Duane Arnold Energy Center (Technical Specification 3.6.F) shall not be operated for a period in excess I of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop out of service.
t a
The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pump or other component renders one loop inoperative. To justify single-loop operation, the safety analyses docu-J mented in the Final Safety Evalua' tion Repor's and, Reference 1 were reviewed 2
~
for one-pump operation. Increased uncertainties in the core total flow and TIP readings resulted in an 0.01 incremental increase.'n i the MCPR fuel cladding 4f integrity safety limit during single-loop operation. This 0.01 increase is
- reflected in the MCPR operating limit. No other incr' ease in this limit is required as core-wide transients are bounded by the rated power / flow analyses
?
1 performed for each cycle, and the recirculation . flow-rate dependent rod block 1
and scram setpoint equations given in the technical specifications are adjusted .
l for one'-pump operation. The least stable power / flow condition, achieved by !
, tripping both recirculation pumps, is not affect,ed by one-pump operation.
I
- , During single-loop operation, the flow control should be in master manual, i
cince control oscillations might occur in tne recirculation flow control j system under automatic flow control conditions. I l !
t f Derived MAPLEGR reduction factors are 0.86, 0.87, 0.87 for the 7x7, 8x8
{
ind 8x8R fuel types, respectively.
l
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l The analyses were performed assuming the equalizer valve was closed. The dis- ,i
( charge valve in the idle recirculation loop is normally closed, but if its f- ' s, closure is prevented, the suetion valve in the loop should be closed to prevent 3
$ the loss of Low Pressure Coolant Injection (LPCI) flow back through the recirculation 4 f" e pump and into the downcomer region.
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. t i 1-1/1-2
- w
2 J- NEDO-24272 -
j '
1 . .
- 2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT -
i . A J 14 1 R
~l Except for core total flow and TIP reading, the uncertainties used in the E I 2 j statistical analysis to determine the MCPR fuel c1' adding integrity safety limit are not independent on whether coolant flow is provided by one or two
{ ,
9 recirculation pumps. Uncertainties used in the two-loop operation analysis are ;
f p
[ documented in the FSAR for initial cores and in Table 5-1 of Reference 1 for j f reloads. A 6% core flow measurement uncertainty has been established for I
I single-loop operation (compared to 2.5% for two-loop operation). As shown f d
below, this value conservatively reflects the one standard deviation (one l,i cigma) accuracy of the core flow measurement system documented in Reference 2. j t
j The random noise component of the TIP reading uncertainty was revised for j cingle recirculation loop operation to reflect the operating plant test results j given in Subsection 2.2 below. This revision resulted in a single-loop opera- j f tion process computer uncertainty of 9.1% for reload cores. The comparable f;.
a l two-loop process computer uncertainty value is 8.7% for reload cores. The net p cffect of these two revised uncertainties is a 0.01 incremental increase in [
o :
L the required MCPR fuel cladding integrity safety limit. ]
3 5
2.1 CORE FLOW UNCERTAINTY t .
2.1.1 Core Flow Measurement During Single Loop Operation The jet pu=p core flev measurement system is calibrated to measure core flow when both sets of jet pumps are in forward flow; total core flow is the sum of the indicated loop flows. For single-loop operation, however, the inactive h
l jet pumps will be backflowing. Therefore, the measured flow in the backflowing ;
jet pu=ps must be subtracted from the measured flow in the active loop. In cddition, the jet pump flow coefficient is different for reverse flow than for forward flow, and the measurement of reverse flow must be modified to account W for this difference.
li n
For single-loop operation, the total core flow is derived by the following i formula: h
- ?otalCore) , / Active Loop j In8ctive L ph h Flow / \ Indicated Flow / - C [ Indicated Flow /
\
2-1 4
%- --_____-______.__--_--__b
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. NEDO-24272 ,
I where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to
" Inactive Loop Indicated Flow," and " Loop Indicated Flow" is the flow indi-cated by the jet pump " single-tap" loop flow summers and indicators, which are set to indicate forward. flow correctly.
The 0.95 factor was the result of a conservative analysis to appropriately modify the single-tap flow coefficient for reverse flow.* If a more exact, '
less conservative core flow is required, special in-reactor calibration tests i would have to be made. Such calibration tests would involve calibrating core I support plate AP versus core flow during two-pump operation along the 100% flow control line, operating on one pump along the 100% flow control line, and cal-culating the correct value of C based on the core flow derived from the core f support plate AP and the loop flow indicator readings.
2.1.2 Core Flow Uncertainty Analysis '
The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump operation is essentially the same as for two-pump operation, h except for some extensions. .The core flow uncertainty analysis is described in Reference 2. The analysis of one-pump core flow uncertainty is su=marized
}
below. " '
For single-loop operation, the total core flow can be expressed as follows (Figure 2-1): .
i W " -
C A I
. i where W = t tal core flow; C
Wg = active loop flow; and W = inactive loop (true) flow.
7
, *The expected value of the "C" coefficient is 40.88. '
i 9
2-2
_ . . . _ _ U
m.
- ~
. i By applying the " propagation of errors" method to the above equation, the '
variance of the total flow uncertainty can be approximated by: ,
2 2 I 32 2
+
a [2 o +o C2\
og " og + ly,1 , og y_, .
g ' -
C sys ( A rand ( rand I where ,
i og = uncertainty of total core flow; op = uncertainty systematic to both loops; sys ,
op = random uncertainty of active loop only; A
rand i o, = random uncertainty of inactive Icop only; I
rand a = uncertainty of "C" Coefficient; and a = ratio of inactive loop flow (Wy) to active loop flow (WA*
I i
Resulted from an uncertainty analysis, the conservative, bounding values of and oC are 1.6%, 2.6%,'3.5% and 2.8%, respectively, ogsys, ogArand, owyrand Eased on above uncertainties and a bounding value of 0.36 for "a", the variance of the total flow uncertainty is approximately: !
i
,c =
(1.6)2 + 1-0.36 (2.6) + f Ib6 -
(3.5)2 + (2.8)2' C
(5.0%)2 O
e 2-3
l ;
.. i
. NEDO-24272 '
i ie When the effect of 4.1% core bypass flow split uncertainty at 12% (bounding
~
case) bypass flow fraction is added to the above total core flow uncertainty, [
I the active coolant flow uncertainty is:
F 1bf2 I (*
~
U setive
= (5.0%) + )
- coolant k }
l which is less than the 6% core flow uncertainty assumed in the statistical analysis.
In summary, core flow during one-pump operation is measured in a conservative l way and its uncertainty has been conservatively evaluated.
2.2 TIP READING UNCERTAINTY To ascertain the TIP noise uncertainty for single recirculation loop operation,
- a test was performed at an operating BWR. The test was performed at a power level 59.3% of rated with a single recirculatlion pump in operation (core flow ;
46.3% of rated). A rotations 11y symmetric control rod pattern existed prior - ,
to the test. 3 Five consecutive traverses were made with each of,five TIP machines, giving a total of 25 traverses. Analysis of their data resulted in a nodal TIP noise i of 2.85%. Use of this TIP noise value as a ec=ponent of the process computer total uncertainty results in a one-sigma process conputer total uncertainty value for single-loop operation of 9.1% for reload cores.
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, NEDO-24272 'J ,
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WC= TOTAL CORE FLOW '
WA= ACTIVE LOOP FLOW ['i!
We = INACTIVE LOOP FLOW j
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Figure 2-1. Illustration of Single Recirculation Loop Operation Flows i b.
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NEDO-24272 .f
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- 3. . MCPR OPERATING LIMIT l ..
1 3.1~ CORE-WIDE TRANSIENTS i,1 i
~0peration with one recirculation loop results in a maximum power output which j' is'20 to 30% below that which is attainable for two-pump operation. Therefore, f*
the consequences of~ abnormal operational transients from one-loop operation '.
will be considerably less severe than those analyzed from a two-loop opera-iS tional mode. For pressurization, flow decrease and cold water increase tran- a sients, previously transmitted Reload /FSAR results bound both the thermal and j overpressure consequences of one-loop operation. ,
Figure.3-1 shows the consequences of a typical pressurization transient (tur- '
bine trip) as a function of power level. As can be seen,.the. consequences of one-loop operation are considerably less because of the associated reduction -
i.
in operating power level.
F The consequences from flow decrease transients are also bounded by the full power analysis. A-single pump trip'from one-loop operation is less severe h than a two-pump trip from full pover because of the reduced initial power * ,1 1 ~1evel. ,)
i.
Cold water-increase transients can result from either recirculation pump I speedup or restart, or introduction of colder water into the reactor vessel by l events such as loss of feedwater heater. The Kg factors are derived assuming f that both recirculation loops increase speed to the maximum permitted by the H-C set scoop tube position. This condition produces the maximum possible
. power increase and, hence, maximum ACPR for transients initiated from less than rated power and flow. When operating with only one recirculation loop, the flow and power increase associated with the increased. speed on only one H-G set will be less than that associated with both pumps increasing speed; i therefore, the K factors f
derived with the two-pump assumption are conserva-tive for single-loop operation. Inadvertent restart of the idle recirculation k pump would result in a neutron flux transient which would exceed the flow reference scram. The resulting scram is expected to be less severe than the rated power / flow case documented in the FSAR. The latter event-(loss of 3-1 w n
NEDO-24272 feedwater heating) is generally the most severe cold water increase event with respect to increase in core power., This event is caused by positive reactivity insertion from core flow inlet subcooling; therefore, the event is primarily ,
dependent on the initial power level. The higher the initial power level, the ,
greater the CPR change during the transient. Since the initial power level _
during one-puep operation will be significantly lower, the one-pu=p cold I water increase case is conservatively bounded by the full power (two-pump) analysis.
From the above discussions, it can be concluded that the transient consequence from one-loop operation is bounded by previously submitted full power analysis.
3.2 ROD WITHDRAWAL ERROR The rod withdrawal error at rated power is given in the FSAR for the initial core and in cycle-dependent reload supplemental submittals. These analyses are performed to demonstrate'that, even if the operator ignores all instrument indications and the alarm which could occur during the course of the transient, the rod block system will stop rod withdrawal at a minimum critical power ratio (MCPR) which is higher than the fuel cladding integrity safety limit. Correc-tion of the rod block equation (below) and lower power assures that the MCPR safety limit is not violated.
One-pump operation results in backflow throu'gh 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps.
Because of the backflow through the inactive jet pumps, the present rod block equation was conservatively modified for use during one-pump operation because the direct active-loop flow measurement may not indicate actual flow above about 35% drive flow without correction.
A procedure has been established for correcting the rod block equation to account for the discrepancy between actual flow and indicated flow in the active loop. This preserves the original relationship between rod block and actual effective drive flow when operating with a single loop.
a 3-2
NEDO- 24272 The two-pump rod block equation is:
The one-pump equation becomes:
e where.
AW = dif ference, determined by utility, between two-loop and single-loop effective drive flow at the same core flow; RB = power at rod bibck in %; .
.. m = flow reference slope for the rod block monitor (RBM);
W = drive flow in % of rated; and RB " t p level rod block at 100% flow.
100 If the rod block setpoint (RB100) is changed, the equation must be recalculated using the new value.
The APRM trip settings are flow biased in the same manner as the rod block monitor t' rip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip set-ting discussed above.
3-3
NEDO-2427 3.3 OPERATING MCPR LIMIT t
For single-loop operation, the rated condition steady-state MCPR limit is increased by 0.01 to account for the increase in the fuel cladding integrity Psfety limit (Section 2). At lower flows, the steady-state MCPR operating .
limit is conservatively established by multiplying the rated flow steady-state limit by the Kg factor. This ensures that the 99.9% statistical limit require-ment is always satisfied for any postulated abnormal operational transient.
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3-4 A
- = , - x w _ :, , _ g . g .ye,y,
p m ,x m . m u m .. =. _ - . - .. -
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>>= -
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g 1060 Nw H
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E 1040 - '
h :
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6 1020 l
l 1000 -
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080 h -4 RANGE'OF EXPECTED MAXfMUM ONL LOOP POWER OPER ATf 0f 4 l
f I I I I 960 80 100 120 t40 20 40 60 O
POWER LEVEL (% NUCLEAR BOILER R ATED) l l
Figure 3-1. Main Turbine Trip with Bypass Manual Flow Control 3-5/3-6
_s NEDO-24272 -
. s - -
- 4. STABILITY ANALYSIS
, l The least stable power / flow condition attainable under normal conditions cccurs at natural circulation with the control rods set for rated power and flow. This condition may be reached following the trip of both recirculation
~
pumps. As shown in Figure 4-1, operation along the minimum forced recircula-
^
tion line with one pump running at minimum speed is more stable than operating with natural circulation flow only, but is less stable than operating with both pumps operating at minimum speed.
1
)
buring' single-loop operation, the flow control should be in master manual, i
cince ' control oscillations might occur in the recirculation flow control system under automatic flow control conditions.
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NEDO-24272 1.2 ULTIMATE STABILITY LIMIT 1.0 -- - -------
=== -===== SINGLE LOOP, POMP MINIMUM SPEED BOTH LOOPS, PUMPS MINIMUM SPEED (L8 -
"o
'9 0.6 - R ATED FLOW
$ CONTROL LINE n:
> NATURAL 8 CIRCULATION /
W LINE /
l if .
0.4 -
/
- HIGHEST POWER ATTAINABLE
- FOR SINGLE LOOP OPERATION 0.2 -
0 0 20 40 60 80 100 POWER (%)
Figure 4-1. Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation
- 4-2
- 5. ACCIDENT ANALYSES The broad spectrum of postulated accidents is covered by six categories of design basis eveats. These e. vents are the loss-of-coolant, recirculation pump .
ceizure, control rod drop, main steamline break, refueling, and fuel assembly '
loading accidents. The analytical results for the loss-of-coolant and recir-culation pump seizure accidents with one recirculation pump operating are i i given below. The results of the two-loop analysis for the last four events ;
cre conservatively applicable for one-pump operation.
5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS E J
s' n
A single-loop operation analysis utilizing the models and assumptions docu- .
-[
er.nted in Reference 3 was performed for Duane Arnold. Using this method, :
lt
{i SAFE /REFLOOD computer code runs were made for a full spectrum of break sizes hm for the suction side breaks. Because the reflood minus uncovery time for the I3 -
eingle-loop analysis is similar to the two-loop analysis, the Maximum Average ja Planar Linear Heat Generation Rate (MAPLHGR) curves currently applied to each
[i unit were modified by derived reduction factors for use during one recircula-i,e tion pump operation. -
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)
5.1.1 Break Spectrum Analysis l1 f
A break spectrum analysis was performed using the SAFE /REFLOOD computer codes
! cnd the assumptions given in Section II.A.7.2.2 of Reference 3. 2
. i h
Tho suction break spectrum reflood times for one recirculation loop operation ,je are compared to the sta2dard previously perfor.ned two-loop operation in (j Figure 5-1. The uncovered time (reflood time minus recovery time) for the rd 4
cuction break spectrum is compcred in Figure 5-2.
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5-1 l
-~ ._ -
, NEDO-24272 For Duane Arnold, the maximum uncovered time for the standard two-loop .
. analysis is 197.7 sec, with a boiling transition time less than 8 seconds, occurring at .100% of the DBA suction break, which is the most limiting -
' break .for the two-loop operation. For the single-loop analysis, the maximum !
uncovered time is 197.0 seconds at 100% DBA suction break. Consequently, for both the single- and two-loop ~ analysis, the limiting break is the 100% DBA
-suction break.
Comparison of the suction break spectrum reflood tLnes between the single- and two-loop analysis shows that the reflood times are similar.
For- the suction break spectrum, the reflooding times for one-loop operation are within 2 seconds of the two-loop operation reflooding times.-
5.1.2 Single-Loop MAPLHGR Determination The small differences in uncovered time and reflo'od time for the limiting _
break size would result in a small difference in the calculated peak cladding temperature. Therefore, as noted in Reference 3, the one- and two-loop SAFE /REFLOOD results can be considered similar and the generic alternative procedure described in Section II.A.7.4 of this reference was used to .
calculate the MAPLHGR reduction factors for single-loop operation. .)
', MAPLHGR reduction factors were determined for the 100% DBA suction break.
l The most limiting reduction factors for each fue] type are shown in i
Table 5-1.
One-loop operation MAPLHGR values are derived by multiplying the ' current-two-loop operation MAPLHGR values by the reduction factor for that fuel type, i
As discussed in Reference 3, single recirculation loop-MAPLHGR values are i
j' conservative when calculated in this manner.
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I i-5-2 p
r
, NEDO- 24272 5.1.3 Small Break Peak Cladding Temperature
- Section II. A.7.4.4.2 of Reference 3 discusses the small sensitivity of the
' calculated peak clad temperature (PCT) to the assumptions used in the one-pump operation analysis and the duration of nucicate boiling. Since the slight increase (450'F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent to 300* to 500*F 4 PCT) for one-pump operation, the calculated PCT values for small breaks will be well below the 2200*F 10CFR50.46 cladding temperature limit.
5.2 .ONE-PUhr SEIZURE ACCIDENT The one-pump seizure accident is a relatively mild event during two-recirculation-pump operation, as documented in References 1 and 2. Similar analyses were performed to determine the impact this accident would have on one-recirculation-pump operation. These analyses wer'e performed with the models documented in Reference 1 for a large co'e r BWR/4 plant (Reference 4).
The anal ses were initialized from steady-state operation at the following initial conditions, with the added condition of one inactive recirculation loop. Two sets of initial conditions were assumed:
5 (1) Thermal Power = 75% and core flow = 58% /
(2) Thermal Power = 82% and core flow = 56%
r I
These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing MAPLHG'R and MCPP limits at the same maximum pump speed. Pump seizure was simulated by setting the single operating pump speed to zero instantaneously.
L The anticipated sequence of events following a recirculation pump seizure which occurs during plant operation with the alternate recirculation loop out ;
of service is as follows:
i (1) The recirculation loop ficw in the loop in which the pump seizure occurs drops instantaneously to zero.
5-3 bL
~ NEDO-242h2 #
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"- (2)' Core. voids increase which results in a negative reactivity insertion-9= -and a sharp decrease in tieutron flux. _ ,
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(3) Heat flux drops more slowly because of the fuel time constant.
(4) Neutron flux, hea't flux, reactor water level, steam flow, and feed-water flow all exhibit transient = behaviors. However, it is not
-anticipated that the increase in water level will cause a turbine trip and result in scram.
,, It is expected that the t.ransient will terminate at a condition of natural-circulation.and_ reactor. operation will continue. There will also be a small
- decrease in system _ pressure. . ,
I
'The minimum CPR'for the pump seizure accident for the large core bht /4 plant ;
I was determined to be greater than the fuel cladding integrity safety limit'; j
~
therefcre, no fuel failures were. postulated to occur as a result of this
. analyzed event.. l
- .. l These results are applicable to the Duane Arnold Energy Center.
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.: HEDO-24272 * .
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Table'5-1 LIMITING MAPLHGR REDUCTION FACTORS Fuel Type Reduction Factors 7x7 0.86 8x8 0.87 8x8R 0.87 l.
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- 6. REFERENCES
- 1. " Generic Reload Fuel Application, General Electric Company", August 1979 (NEDE-240ll-P-A-1).
- 2. " General Electric BWR Thermal Analysis Basis (CETAB): Data, Correlation -
and Design Application", General Electric Cc=pany, January 1977 ,
(NEDO-10958-A).
I
- 3. " General Electric Company Analytical Model for Loss-of-Coolant Analysis '
in Accordance with 10CFR50 Appendix K Amendment No. 2 - One Recirculation [
isop Out-of-Service", General Electric Company, Revision 1, July 1978 r (NEDO-20566-2). >
I
- 4. Enclosure to Letter #TVA-BFNP-TS-ll7, O. E. Gray III to Harold R. Denton, j September 15, 1978. ;
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