ML19331C220

Single-Loop Operation.

ML19331C220
Person / Time
Site:	Cooper
Issue date:	05/31/1980
From:	Zanardi G GENERAL ELECTRIC CO.
To:
Shared Package
ML19331C218	List: ML19331C217 ML19331C219 ML19331C220
References
8ONED267, NEDO-24258, TAC-42418, NUDOCS 8008140380
Download: ML19331C220 (31)

Text

NEDO-24258 80NED267 g Class 1 May 1980 COOPER NUCLEAR STATION SINGLE-LOOP OPERATION l

NUCLEAR POWER SYSTEMS OtvlSION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNi A 95125

?iEDO-24258

- m D.r.eCLA.rLer.o C.r" .c..r.e. rC.n..e..r.p.ur 2. y This doc-ren: uaa -repared e b.u cr .*'cr the General Electric Ccqany.

Neither the General Electric Ccepany nce anL of the ccn:ributcra to thia dorrent:

A.  !.'akca any carranty or representaticn, caprese cr implied, with respec: to the accuracy, ccqle ences, cr usefulness of the infcmation acntained in thic dce:r:cnt, cr that the use of any infomaticn discicaed in chia doewent may not infringe pri-xcaly ~.med righ a; or

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c. nas:c ce any respona:ctis.t.u s

~,c" usabu. . .

. .;:u or danaae c.r. any cna.

uhich may reault frcn the use of any infomaticn disclosed in

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NEDO-24258 TABLE OF CONTENTS Page,

1. INTRODUCTION AND

SUMMARY

1-1

2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 2.1 Core Flow Uncertainty 2-1 2.1.1 Core Flow Measurement During Single Loop Operation 2-1 2.1.2 Core Flow Uncertainty Analysis 2-2

, 2.2 Tip Reading Uncertainty 2-4

3. MCPR OPERATING LIMIT 3-1 3.1 Core-Wide Transients 3-1 3.2 Rod Withdrawal Error 3-2 3.3 Operating MCPR Limit 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 liAPLHCR Determination 5-2 5.1.3 Small Break Peak Cladding Temperature 5-3 5.2 One-Pump Seizure Accident 5-3

6. REFERENCES 6-1 1

I j

i iii/iv

. . - ._. _ - .. e _ _.___ . ._ - -,-- . . -

i I NEDO-24258

,l ILLUSTRATIONS Figure- Title Page

2-1 Illustration of Single Recirculation Loop Operation Flows 2-5 3-1 Main Turbine Trip With Bypass Manual Flev Control 3-5 4-1 Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 4-2 5-1 Cooper Reflooding Time vs. Break Area 5-7 4

5-2 Cooper'Reflooding Time vs. Break Area 5-8 5-3 Cooper Total Uncovered Time vs. Break Area 5-9 5-4 Cooper Total Uncovered Time vs. Break Area 5-10 I

i 1

1 l

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k e

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4 v/vi g c---- -- -- --6 - - -we m- e +-.-; - -t---y w,m--

• 8% -

t NEDO-24258 TABLES Table Title Page 5-1 MAPLHCR Multiplier Case 5-5 5-2 Limiting MAPLHCR Reduction Factors 5-6 vii/viii

NEDO-24258

1. INTRODUCTION AND SU!DfARY The current technical specifications for Cooper Nuclear Station do not allow plant operation beyond a relatively short period of time if an idle recircula-tion loop cannot be returned to service. Cooper Nuclear Station (Technical Specification 3.6.F.3) 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.

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-mented in the Final Safety Evaluation Reports and Reference 1 were reviewed 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 du,ing single-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. The least stable power / flow condition, achieved by tripping both recirculation pumps, is not af fected by one-pump operation.

Under single-loop operation, the flow control should be in master manual, since control oscillations may occur in the recirculation flow control system under these condtions.

Derived MAPLHGR reduction factors are 0.84, 0.86, 0.77 for the 7x7, 8x8 and 8x8R fuel types, respectively.

The analyses were performed assuming the equilizer 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 prevent the loss of Low Pressure Coolant Injection (LPCI) flow out of a postulated break in the idle suction line.

1-1/1-2

mfd 0-24358

2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the statistical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two recircula-tion pumps. Uncertainties used in the two-loop operation analysis 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 single recirculation loop operation to reflect the operating plant test results given in Subsection 2.2 below. This revision resulted in a single-loop opera-tion process computer uncertainty of 9.1% for reload cores. Comparable two-loop process computer uncertainty value is 8.7% for reload cores. The net effect of these two revised uncertainties 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 single-loop operation, however, the inactive jet pumps will be backflowing. Therefore, the measured flow in the backflowing 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.

For single-loop operation, the total core flow is derived by the following formula:

Active Loop Inactive Loop

-C (TotalCore Flow Indicated Flow Indicated Flow 2-1

NED0-2425k 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 indicated 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 measurement is required, special in-reactor calibration tests would have to be made. Such calibration tests would involve calibracing core support plate AP versus core flow during two-pump operation along the 100% flow control line, operati 3 on one pump along the 100% flow control line, and calculating 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 = W -W 7 where Mg = total core Gom Wg = active loop flow; and W = inactive loop (true) flow.

7

• The expected value of the "C" coefficient is %0.88.

2-2

I!EDO-24258 By applying the " propagation of errors" method to the above equation, the variance of the total flow uncertainty can be approximated by:

2 2 2 "

2

+ 1 2

+ a [2

1. +#

2

• tfC W l-a W l-a W

sys A rand kIrand C) where cp = uncertainty of total core flow; i C cy = uncertainty systematic to both loops; sys og =

random uncertainty of active loop enly; A

rand og =

random uncertainty of inactive loop only; a

C

= uncertainty of "C" coefficient; and a = ratio of inactive loop flow (W 7) to active loop ficw (W ).

Resulted from an uncertainty analysis, the conservative, bounding values of og ,o g ,o g ande are 1.6%, 2.6%, 3.5% and 2.8%, respectively.

C sys A I rand rand Based on tbove uncertainties and a bounding value of 0.36 for "a", the variance of the total flow uncertainty is approximately:

2

• 0.36 U[4 C (1.6)2+('1 l-0.36 ( .6)~ + (1-0.36 ) -(3.5)2 + (2.8)2"

=

(5.0%)~

2-3

l l

NED0-24258 When the effect of 4.1% core bypass flow split uncertainty at 12% (bounding case) bypass flow fraction is added to the abcve total core flow uncertainty, the active coolant flow uncertainty is:

c 2 = (5.0%)2 , 0.12 g 2 (4.1%)2 =

(5.0%)2

. coolant which is less than the 6% core flow uncertainty assumed in the statistical analysis.

r 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 fer single recirculation loop operation, a test was performed at an operating BUR. 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 truverses. Analysis of their data resulted in a nodal TIP noise of 2.83%. Use of this TIP noise value as a component of the process computer total uncertainty results fn a one-signa process computer total uncertainty value for single-loop operation of 9.1% for reload cores.

2-4

NEDO-24258 CORE J L 4k

] l J [

WC w, WA WC

• TOTAL CORE FLOW WA = ACTIVE LOOP FLOW We = INACTIVE LOOP FLOW Figure 2-1. Illustration of Single Recirculation Loop Operation Flows 2-5/2-6

NEDO-24258

3. MCPR OPERATING LIMIT 3.1 CORE-WIDE TRANSIENTS i

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 o'rerpressure consequences of one-loop operation.

j 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.

l The consequences from flow decrease transients are also bounded by the full power analysis. A single punp trip from one-loop operation is less severe than a two-pump trip from full power because of the reduced initial power level.

i Cold water increase transients can result from either recirculation pump speedup or restart, or introduction of colder water into the reactor vessel by events such as loss of feedwater heater. The Kgfactors are derived assuming l 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 l M-G set will be less than that associated with both pumps increasing speed; therefore, the K factors g derived with the two-pump assumption are conserva-tive for single-loop operation. Inadvertent restart of the idle recirculation 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 feed-water heating) is generally the most severe cold water increase event with 3-1 4

NED0-24258 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-pump 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 UITHDRAWAL 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. Corree-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 flaw 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.

3-2

. . . . = - _- _ _ = _ - .

NEDO-26258 The two-pump rod block equation is:

RB = mW + RB 100 - m( 00)

The one-pump equation becomes:

I _ _

RB = mW + RB 100 - m(1 0) - maw 4

l where aW =

difference, determined by utility, between two-loop and single-

loop effective drive flow at the same core flow; RB =

pot'er at rod block in %;

1 I

m = flow reference slope for the rod block monitor (RBM), and W =

drive flow in % of rated.

i RB100 " t p level rod block at 100% flow.

If the rod block setpoint (RB100) is changed, the equation must be recalcu-laced using the new value.

l The APRM trip settings are flow biased in the same manner as the rod block i

monitor trip setting. Therefore, the APRM rod block and scram trip setting ~

are subject to the same procedural changes as the rod block nonitor trip set-ting discussed above.

3-3

_ __. -_ __ ~ - _- - - - . . . _ . _ _ - - . - . . -- -

NEDO-24258 3.3 OPERATING MCPR 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 safety limit (Section 2). At lower flows, the steady-state operating MCPR limit is conservatively established by multiplying therated flow steady-state limit by the Kg factor. This ensures that the 99.97. statistical limit require-ment is always satisfied for any postulated abnormal operational occurrence.

3-4

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RANGE OF EXPECTED & 1 MAXIMUM ONE-LOOP POWER OPERATION geo m I I I I O 20 40 60 80 100 120 140 POWER LEVEL M NUCLEAR BOILER RATED)

Figure 3-1. Main Turbine Trip with Bypass Manual Flow Centrol 3-5/3-6

NEDO-24258-

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 natural circulation flow only, but is less stable than operating with both pumps oper-ating at minimum speed. Under single-loop operation, the flow control should be in master manual, since control oscillations may occur in the recirculation flow control system under these conditions.

1 6

4-1

1 NED0-24258 1.2 ULTIMATE STABILITY LIMlf 1.0 -- ------- --- - -

== = === === SINGLE LOOP. PUMP MINIMUM SPEED

- BOTH LOOPS, PUMPS MINIMUM SPEED O. 8 -

^o A

9 Q o.6 -

=

NATURAL O CIRCULATION / RATED FLOW CONTROL LINE LINE

/

/ Y o.4 -

HIGHEST POWER A TTAIN AS LE

! FOR SINGLE LOOP OPER ATION o.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

NEDO-24258

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 operating are given below. The results of the two-loop analysis for the last four events are conservatively applicable for one-pump operation.

5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS A single-loop operatien analysis utilizing the models and assumptions documen-ted in Reference 3 was performed for Cooper Nuclear 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 (MAPLHGR) curves cur-rently applied to Cooper were modified by derived reduction factors for use during one recirculation pump operation, s

5.1.1 Break Spectrum Analysis l

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.

The suction and discharge break' spectrum reflood times for one recirculation loop operation are compared to the standard previously performed two-loop oper-ation in Figures 5-1 and 5-2. The uncovered time (reflood tiue minus recovery time) for the suction and discharge break spectrum is compared in Figures 5-3 and 5-4.

For Cooper Nuclear Station, the maximum uncovered time for the standard two-loop analysis is shown in Figure 4, occurring at 80% of the DBA discharge break, 5-1

WEDO-24258 l

i which is the most limiting break for the two-loop operation. For the single-loop analysis, the maximum uncovered time occurring at 77.5% DBA discharge is withir 4 seconds of the uncovered time at the 80% DBA discharge break. Hence, the larger break (80% DBA) will be more limiting because of the earlier uncovery and corresponding higher decay heat during the uncovered period. Consequently, 4 for both the single- and two-loop analysis, the limiting break is the 80% DBA discharge break.

Comparison of the suction and discharge break spectrum reflooding times between the single- and two-loop analysis shows that the reflooding 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. In the dis-charge break spectrum, the single-loop reflood times are approx 1 ately equal to or less char. the two-loop reflood times for breaks equal to or greater than 80% DBA.

[

5.1.2 Single-Loop MAPLHGR Determination The small differences in uncovered time and reflood tiee 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 /

i

REFLOOD results can be considered similar and the generic alternative procedure described in Section II.A.7.4. ?f this reference was used to calculate the MAPLHGR reduction factors for single-loop operation.

i l MAPLHCR reduction factors were determined for the cases given in Table 5-1.

The most limiting reduction factors for each fuel type is shown in Table 5-2.

Slightly longer calculated boiling transition time for the 8x8R fuel required use of curve 3 on Figure II. A.7.4-1 of Reference 3, rather than curve 2 used for the other fuel types. 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 conservative when calculated in this manner.

5-2

i fiED0-24358 7

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 nucleate bbiling. Since the slight increase (N50*F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent to 300' to 500*F % PCT) for one pump cperation, the calculated PCT values for small breaks will be well below the 2200'F iUCFR50.46 cladding temperature limit.

i 5.2 ONE-PUMP SEIZURE ACCIDENT

The one-pump seizure accident is a relatively mild event during two recircula-tion pump operation as documented in References 1 and 2. Gimilar analyses

were performed to determine the impact this accident would have on one recir-I culation' 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 flov = 58%

l (2) Thermal Power = 82* and core flow = 56%

These conditions were chosen because they represent reasonable upper limits of

'l 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 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:

(1) The recirculation loop flow in the loop in which the pump seizure occurs drops instantaneously to zero.

5-3

NEDo-24258 1

(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.

It is expected thet the transient will terminate at a condition of natural circulation and reactor.operat'on i will continue. There will also be a small decrease in system pressure.

E t

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 '.imit; therefore, no fuel failures were postulated to occur as a result of this

. analyzed event.

l T

5-4 -

NEDO-34258 Table 5-1  ;

MAPLHGR MULTIPLIER CASES Fuel Type Cases Calculated 7x7 100% DBA Discharge Break 80% DBA Discharge Break

• 100% DBA Suction Break 8x8 100% DBA Discharge Break 80% DBA Discharge Break

• 100% DBA Suction Break 8x8R 100% DBA Discharge Break 80% DBA Discharge Break

• 100% T:BA Suction Break

• !iost limiting break for MAPLHGR reduction factors.

i t

i a

i

5-5

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NED0-24258 Tabic 5-2 LIMITING MAPLEGR REDUCTION FACTORS Fuel Type Reduction Factors 7x7 0.84 8x8 0.86 8x8R 0.77 5-6

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NEDO-24258

6. REFERENCES

1. " Generic Reload Fuel Application, General Electric Company", August 1979 (NEDE-240ll-P-A-1).

2. "Genaral Electric BtTR Thermal Analysis Basis (CETAB): Data, Correlation, and Design Application", General Electric Company, January 1977-(NED0-10958-A).

3. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K Amendment No. 2 - One Recirculation Loop Out-of-Service", General Electric Company, Revision 1. July 1978 '

(NEDO-20566-2).

4. Enclosure to Letter #TVA-BFNP-TS-il7, O. E. Gray III to Harold R. Denton, September 15, 1978.

N*

t 4

t 6-1/6-2

i NUCLEAA ENERGY DIVISIONS ,e CENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 96125 GENER AL h ELECTRIC TECHNICAL INFORMATION EXCHANGE TITLE PAGE AUTHOR SUBJECT TIE NUMBER G. ZM!ARDI and Technology oATE May 1980 TITLE GE CLASS I

COOPER NUCLEAR STATION SINGLE LOOP OPERATION GOVERNMENT CLASS NUMBER OF PAGES REPRODUCIBLE COPY FILED AT TECHNICAL SUPPORT SERVICES. R&UO. SAN JOSE. 31 CALIFORNIA 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-l ponent renders one loop inoperative. To justify single-l 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 incre-mental safety limit during single-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 specifica-tions are adjusted for one-pump operation.

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DOCUMENT NUM8ER INFORM ADON PREP ARE D FOR Nuclear Power Svstems Division SECTION Safety and Licensing Operation K Rs. 2602 M All CODE 632 BUILDING AND ROOM NUMBER