ML19347F318
| ML19347F318 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/30/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19347F316 | List: |
| References | |
| 80NED275, NEDO-24268, NUDOCS 8105180362 | |
| Download: ML19347F318 (30) | |
Text
1 NEDO-24268 j
80NED275 Class I June 1980 PILGRIM NUCLEAR POWER STATION SINGLE-LOOP OPERATION I
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NUCLEAR POWER SYSTEMS DIVISION e GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORN!A 95125 GEN ER AL $ ELECTRIC
%Io 578'o361
NEDO-24268 DJ DISCLAIE R OF RESPONSIBILITI This document was prepared by or for the General Electric Company.
Neither the Gensval Electric Company nor any of the contributora to this doement:
A.
Makes any varranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information c>ntained in this document, or that the use of any infomatior, disclosed in this document may not infringe pri-vately owned righte; or B.
Assumes any responsibility for liability or damage of any kind which may result from the use of any infomation disclosed in this document.
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i NEDO-24268 CONTENTS Page 1.
INTRODUCTION AND
SUMMARY
l-1 2.
MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 2.1 Core Flow ?fncertainty 2-1 2.1.1 Core Flow Measurement During Single-Loep 2-1 Operation 2.1.2 Core Flow Uncertainty Analysis 2-2 2-4 2.2 TIP Reading Uncertainty 3.
MCPR OPERATING LIMIT 3-1 3.1 Core-Wi.-e Transients 3-1 3.2 Rod Wit.drawal Error 3-2 3.3 MCPR Operating Liinit 3-4 4.
STABILITY ANALYSIS 4-1 5.
ACCIDENT ANALYSES 5-1 5.1 Loss-of-Coolant Accident Analysis 5-1 5.1.1 Break Spectrum Analysis 5-1 5.1.2 Single-Loop MAPLHCR Determination 5-2 5.1.3 Small Break Peak Cladding Temperature 5-2 5.2 One-Pump Seizure Accident 5-3 6.
REFERENCES 6-1 APPENDIX A.
RESULTS OF THE SINGLE-LOOP ANALYSIS POR ECCS LIMITS WITH NO CREDIT FOR CORE SPRAY HEAT TRANSER A-1 f
l iii/iv
NED0-24268 TABLES Table Title Page 5-1 Limiting MAPLHGR Reduction Factors 5-5 ILLUSTRATIONS Figure Title Pjypi 2-1 Single Recirculation Loop Operation Flows 2-5 3-1 Main Turbine Trip with Bypass Manual Flow Control 3-5 4-1 Decay. Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 4-2 5-1 Pilgrim Nuclear Power Station Suction Break Spectrum Reflooding Times 5-6 5-2 Pilgrim Nuclear Power Station Suction Break Spectrum Uncovered Times 5-7 v/vi
INTRODUCTION AND
SUMMARY
I The current technical specifications for Pilgrim Nuclear Power Station do not allow plant operation beyond a relatively short period of time if an idle recirculation loop.cannot be returned to service.- Pilgrim Nuclear Power Sta-tion (Facility Operating License'3.E) shall not be operated for a period in excess 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
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 -
l loop inoperative. To ~ justify single-loop operation, the safety analyses docu-mented in the Final Safety Evaluation Reports and Reference 1 were reviewed i
.for one-pump operation.
Increased uncertainties in the core total flow and TIP readings'resulted in an 0.01 incremental increase in the MCPR fuel cladding
~
integrity safety limit during single-loop operation. This 0.01 increase is reflected in the MCPR operating limit. No oth'er increase in this limit is required as core-wide transients are bounded by the rated power / flow analyses j
performed for each cycle, and the recirculation flow-rate dependent rod block j
and scram setpoint equations given in the technical specifications are adjusted j
for one-pump operation. The least stable power / flow condition, achieved by tripping both recirculation pumps, is not affected by one-pump operation..
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During single-loop operation, the flow control should be in master nanual, since control oscillations might occur in the recirculation flew control system under automatic flow control conditions.
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The derived MAPLHGR reduction factor is 0.86 for the 8x8 and 8x8R fuel types.
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The analyses were performed assuming the equalizer valve was closed. The dis-charge valve in the idle recirculation loop is normally closed, but if its
)
closure is prevented, the suction valve in the loop should be closed to pre-(
vent the loss of Low Pressure Coolant Injection (LPCI) flow out of a postu-
!4' lated break in the idle suction line.
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1-1/1-2
NEDO-24268 2.
MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the statiscical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one of two recircula-tion pumps. Uncertainties used in the two-loop operation analyses are docu-mented in the FSAR for initial cores and in Table 5-1 of Reference 1 for reloads. A 6% core flow measurement uncertainty has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 2.
The random noise component of the TIP reading uncertainty was revised for singic recirculation loop operation to reflect the operating plant test results given in Subsection 2.2 below. This revision resulted in a single-loop operation process computer uncertainty of 9.1% for reload cores.
The comparable two-loop process computer uncertainty value is 8.7% for re. load cores. The net effect of these two revised uncertafuties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.
2.1 CORE FLOW UNCERTAINTY 2.1.1 Core Flow Measurement During Single-Loop Operation The jet pump core flow 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 sing 1c-loop operation, however, the inactive l
jet pumps will be backflowing.
Therefore, the measured flow in the back-i flowing jet pumps must be subtracted from the measured flow in the active loop.
In addition, 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 for this difference.
2-1
For single-loop operation, the total core flow is derived by the following formula:
(TotalCore Active Loop Inactive Loop
-C Flow Indicated Flo Indicated Flo 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 would have to be made.
Such calibration tests would involve calibrating core 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 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, except for some extensions. The core flow uncertainty analysis is described in Ref-erence 2.
The analysis of one-pump core flow uncertainty is summarized below.
For single-loop operation, the total core flow can be expressed as follows (Figure 2-1):
W C
A ~ I
- The expected value of the "C" coefficient is 40.88.
2-2 e_.
,~
~
NEDO-24268 where total core flow; W
=
C active loop flow; and W
=
A inactive loop (true) flow.
W
=
7 By applying the " propagation of errors" method to the above equation, the variance of the total flow uncertainty can be approximated by:
[#
I 2
2 2
1 2
a 2
2
+#
l
+
+
W W
l-W 1-a W
C C
sys Arand
(
rand
/
where uncertainty of total core flow; o
=
g uncertainty systematic to both loops; o
=
gsys random uncertainty of active loop only; o
=
Arand random uncertainty of inactive loop only; o
=
g Irand uncertainty of "C" coefficient; and
=
C i
ratio of inactive loop flow (W ) to active loop flow a =
7 (W }*
A Resulted from an uncertainty analysis, the conservative, bounding values of og
, og
, og and oC are 1.6%, 2.6%, 3.5% and 2.8%, respectively.
sys Arand Irand Based on above uncertainties and a bounding value of 0.36 for "a",
the variance of the total flow uncertainty is approximately:
2 2
(1-036 (2.6[+fy,0 (1.6) +
(3.5) + (2.8) o
=
W 36 C
(5.0%)2
=
2-3
NEDO-24268 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, the active coolant flow uncertainty is:
0 2 (5.0%)2 (5.0%)2,
ctive (4.1%)
=
=
coolant 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 way and its uncertainty has been conservatively evaluated.
2.2 TIP READING UNCERTAINTY To ascertain the TIP noise uncertainty for single recirculation loop operation, J
a test was performed at an operating BWR.
The test was performed at a power level 59.3% of rated with a single recirculation pump in operation (core flow
}
46.3% of rated). A rotationally symmetric control rod pattern existed prior to the test.
Five consecutive traverses were made with each of five TIP machines, giving a total of 25 tr.tverses. Analysis of their data resulted in a nodal TIP noise of 2.85%.
Use of this TIP noise value as a component of'the process computer total uncertainty results in a one-sigma process computer total uncertainty value for single-loop operation of 9.1% for reload cores.
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NEDO-24268 CORE i L
/
)
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L WC 3
WA We = 10TAL COHE FLOW WA = ACTIVE LOOP FLOW W3 = INACTIVE LOOP FLOW Figure 2-1.
Singic Recirculation Loop Operation Flows 2-5/2-6
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NED0-24268 i
3.
MCPR OPERATING LIMIT 3.1 CORE-WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which is 20 to 30% below that which is attainable for two-pump operation. Therefore, the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-loop opera-tional mode.
For pressurization, flow decrease and cold water increase tran-sients, previously transmitted Reload /FSAR results bound both the thermal and overpressure consequences of one-loop operation.
Figure 3-1 shows the consequences of a typical pressurization transient (tur-bine trip) as a f unction of power level. As can be seen, the consequences of one-loop operation are considerably less because of the associated reduction in operating power level.
The consequences from ficw decrease transients are also bounded by the full power analysis. A single pump trip from one-loop operation is less severe than a two-pump trip from full power because of the reduced initial power level.
Cold water increase transients can result from either recirculation pump speedup or restart, or introduction of colder water into the reactor vessel by eveats such as loss of feedwater heater.
The K factors are derived assuming f
that both recirculation loops increase speed to the maximum permitted by the M-G 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 M-G set will be less than that associated with both pumps increasing speed; therefore, the K factors derived with the two-pump assumption are cot.serva-g tive for single-loop operation.
Inadvertent restart of the idle recirculation pump would result in a neutron flux transient which would enceed the flow reference scram. The resulting scram is expected to be less severe than the 3-1
NEDO-24268 rated power / flow case documented in the FSAR. The latter event (loss of feedwater heating) is generally the most severe cold water increase event witn 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 trat.sient.
Since the initial power level during one-rump operation will be significantly lower, the one-pump cold 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 through 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 discrepancv 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.
3-2
NEDb-24268
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The two-pump rod' block equation is:
- m( 00)
=
100 The one-pump equation becomes:
RB mW +
- m(100)
- dW
=
100 where AW difference,-' determined by utility, between two-loop and single-
=
loop effective drive flow at the same core flow; RB power at rod block (%);
=
flow reference slope for the rod block monitor (RBM);
m =
i drive flow (% of rated); and W
=
l RB100 p eve r block at 100% flow.
=
If the rod block setpcint (RB100) is changed, the equation must be recalculated i
using the new value.
The APRM trip settings are flow biased in the same manner as the rod block j
monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the road block monitor trip set-ting discussed abcve.
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3-3
NEDO-24268 3.3 MCPR OPERATING LIMIT 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 cafety 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 K factor. This ensures that the 99.9% statistical limit require-g ment is always satisfied for any postulated abnormal operational transient.
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NEDO-24268 1160 3
1140 Y<w 5
=f 0
200 1120 x
5 d
e
{
1100 3
o E
iooy 3=
E 5
a a
E
!E E
d ioso a
5 z
W E< 1040 N
1o20 1000 -
980 RANGE OF EXPECTED 1
MAXIMUM ONE-LOOP POWFR OPER ATirN l
l sco 0
20 40 60 ao 100 120 i4o POWER LEVEL (% NUCLEAR BOILER RATED)
Figure 3-1.
Main Turbine Trip with Bypass Manual Flow Control 3-5/3-6
NEDO-24268 4.
STABILITY ANALYSIS The least stable power / flow condition attainable under normal conditions occurs 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 recirculation line with one pump running at minimum speed is more stable than operating with nat-ural circulation flow only, but is less stable than operating with both pumps operating at minimum speed..
During single-loop operation, the flow control should be in master manual, since control oscillations might occur in the recirculation ficw control system under automatic flow control conditions.
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4-1
NEDO-24268 1.2 ULTIMATE STABILITY LIMIT 1.0
- - SINGLE LOOP, PUMP MINIMUM SPEEr BOTH LOOPS, PUMPS MINIMUM SPEED 0.8
-?
/
n
.5 9
Q o.6 x
Uu NATURAL o
CIRCULATION RATED FLOW LINE CONTROL LINE
//
//
o.4 HIGHEST POWER
/
ATTAINABLE FOR SINGLE LOOP OPER ATION 0.2 I
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I o
20 40 so 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 events.- These events are the loss-of-coolant, recirculation pump seizure, control' rod drop, main steamline break, refueling,-and fuel assembly loading accidents. The analytical results for the loss-of-coolant and recirculation pump seizure accidents with one recirculation pump oper-ating-are given below. The results of the two-loop analysis for the.lastJfour
~
events are conser'vatively applicable for one-pump operation.
5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS A single-loop operation analysis utilizing the models and assumptions docu-mented in Reference 3 was performedLfor Pilgrim Nuclear Power Station.
Using this method, SAFE /REFLOOD computer code runs were made for a full spectrum of break sizes for both the suction and discharge side breaks.
Because the reflood minus uncovery time for the single-loop analysis is similar to the two-loop analysis, the Maximum Average Planar Linear Heat Generation Rate 4
(MAPLHGR) curves currently applied to each unit were modified by derived
-reduction factors for use during one recirculation pump operation.
I 5.1.1 Break Spectrum Analysis A break spectrum analysis was performed using the SAFE /REFLOOD computer codes and the assumptions given in Section II.A.7.2.2. of Reference 3.
Since non-LPCI model plants are limited by suction breaks, the analysis was l
performed for suction break spectrum only.
t
-The. suction break spectrum reflood times for one recirc.ulation loop opera-tion are compared to the standard previously performed two-loop operation in Figure 5-1.
The uncovered time (reflood time minus recovery time) for the
. suction break spectrum is compared in Figure 5-2.
For the standard two-loop analysis, the most limiting break was a 100% suction I~
DBA=with a total uncovered time of 154 sec and a boiling transition time of 5.84 sec.
5-1
NEDO-24268 For single-loop analysis, a bsiling transition of 0.1 see is conservatively assumed for all breaks larger than 1.0 f t2, and the reflooding times and
~
total uncovered times are similar'to the-two-loop analysis (Figures 5-1 and 5-2).
Therefore, the most limiting break for single-loop analysis is also the 100% suction DBA.
Comparison of the suction break spectrum reflood times between the single-and two-loop analysis shows that the reflood times are similar.
For the suc-tion break spectrum, the reflooding times for one-loop operation are within a few seconds of the two-loop operation reflooding times.
5.1.2 Single-Loop MAPLHGR Determination The small differences in uncovered time and reflood time for the limiting break size would result in a small increase in the calculated peak cladding tem-perature. 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~
MAPLHCR reduction factors for single-loop operation.
The most limiting reduction factors for each fuel type are shown in Table 5-1.
The MAPLHGR reduction factor with no credit for core spray hat transfer is given in Appendix A.
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One-loop operation MAPLHGR values are derived by multiplying the current two-loop operation MAPLHGR values by the reduction factor for that fuel type.
As discussed in Reference 3, single recirculation loop MAPLHGR values are con-servative when calculated in this manner.
5.1.3 Small Break Peak Cladding Temperature Section II.A.7.4.4.2 of Reference 3 discussed the small sensitivity of the calculated peak clad temperature (PCT) to the assumptions used in the one-
. pump operation analysis and the duration of nucleate boiling.
Since the slight l
increase (s50 F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent to 300 to 500 F in 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
NEDO-24268 5.2 ONE-PUMP SEIZURE ACCIDENT The one-pump seizure accident is a relatively mild event during two-recircu-lation-pump operation, as documented in References 1 and 2.
Similar analyses were performed to determine the impact this accident would have on one-recirc-ulation-pump operation. These analyses were performed with the models docu-mented in Reference 1 for a large core BWR/4 plant (Reference 4).
The analyses were initialized from steady-state operation at the following initial condi-tions, with the added condition of one inactive recirculation loop. Two sets of initial conditions were assumed:
(1)
Thermal Power = 75% and core flow = 58%
~(2)
Thermal Power = 82% and core flow = 56%
These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing MAPLHGR and MCPR limits at the same maximum pump speed. Pump seizure was simulated by setting the single operating pump speed to zero instantaneously.
The anc.icipated 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 flow in the loop in which the pump seizure occurs drops instantaneously to zero.
(2)
Core voids increase, which results in a negative reactivity insertion and a sharp decrease in neutron flux.
(3)
Heat flux drops more slowly because of the fuel time constant.
(4)
Neutron flux, heat 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.
5-3
NEDO-24268 It-is expected _that the transient will terminate at a condition of natural circulation and reactor operation will continue. There will also be a small
- decrease in system pressure.
The minimum CPR for the pump seizure. accident for the large core BWR/4 plant
~
was determined to be greater than the fuel cladding integrity safety limit; therefore, no fuel failures were postulated to occur as a result of this analyzed event.
These results are applicable to the Pilgrim Nuclear Power Station.
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5-4
J o
NEDO-24268 Table 5-1 LIMITING MAPLHGR REDUCTION FACTORS Fuel Type Reduction Factor 8x8 0.86 8 x 8R 0.86 5-5
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NEDO-24268 6.
REFERENCES 1.
. Generic Reload Fuel Application, General Electric Company, August 1979 (NEDE-24011-P-A-1).
2.
General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application, General Electric Company, January C 77 (NEDO-10958-A).
3.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 20CFR50 Appendix K Amendment No. 2 - One Recirculation
~ Loop Out of Service, General Electric Company, Revision 1, July 1918 (NED9-20566-2).
4.
Enclosure to Letter #TVA-BFNP-TS-117, O. E. Gray III to Harold R. Denton, September 15, 1978.
I 6-1/6-2
NEDO-24268 APPENDIX A This appendix provides the results of the single-loop analysis for the ECCS
~
limits with no credit for the core spray heat transfer system.
The CHASTE code was used to provide MAPLHGR values for single-loop analyses, where a boiling transition time of 0.1 see was conservatively assumed for all breaks larger than 1.0 ft, and the reflooding times and total une.'vered time were similar to the two-loop analyses.
In addition, no credit was taken for core spray heat transfer. However, credit was taken for fuel channel heat transfer after the bypass level became equal 'o the elevation of the highest relative-power fuel node to be consistent with the two-loop analysis with no core spray heat transfer credit.
The derived MAPLHCR reduction factor determined according to Reference 3 is 0.84 for 8x8 and 8x8R fuel types. This factor is to be applied in addition to the factors already derived for two-loop operation without credit for core spray heat transfer, but with the bypass region flooding credit.
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A-1/A-2
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NUCL'AR ENERGY DIVISIONS e GENERAL ELECTRIC CGMPANY SAN JOSE, CALIFORNIA 96126 G EN ER ALM E LECTRIC TECHNICAL INFORMATION EXCHANGE TITLE PAGE AUTHOR SUBJECT TIE NUM8ER uclear Science G. Zanardi DATE and Techt. ology June 1980 TITLE GE CLASS PILGRIM NUCLEAR POWER STATION I
SINGLE-LOOP OPERATION GOVERNMENT CLASS REPRODUCIBLE COPY FILED AT TECHNICAL NUM8ER OF PAGES SUPPORT SERVICES, R&UO, SAN JOSE, 26 CAllFORNIA 96125 (Mail Code 211)
SUMMARY
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 com-ponent renders one loop inoperative. To justify single-loop operation, the safety analyses documented in the Final Safety Evaluation Reports were reviewed for one-pump operation.
Increased uncertainties in the core total flow and TIP readings resulted in an 0.01 incremental safety limit during singic-loop operation.
This 0.01 increase is reflected in the MCPR operating limit. No other increase in this limit is required as core-wide transients are bounded by the rated power /
flow analyses performed for each cycle, and the recirculation-flow-rate-dependent rod block and scram setpoint equations given in the technical specifications are adjusted for one-pump operation.
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By aJtting out this rectangle and folding ir half, the above information can be fitted into a standard card file.
MCUMENT nut *BER NEDO-24268 FORMATION PREPARED FOR Nuclear Power Systems Division SECTION Safety and Licensing Operation SulLDING AND ROOM NUMBER K, Rm. 2602 M AIL CODE 682