ML20053E394

Single-Loop Operation.

ML20053E394
Person / Time
Site:	Brunswick
Issue date:	09/30/1981
From:	GENERAL ELECTRIC CO.
To:
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ML20053E363	List: ML20053E362 ML20053E375 ML20053E394
References
81NED277, NEDO-24344, NUDOCS 8206070868
Download: ML20053E394 (32)

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BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 SINGLE-LOOP OPERATION

J.a u g. GENER AL h ELECTRIC

NEDO-24344 81NED277 Class I September 1981 l

l BRUNSWICK STEAM ELECTRIC PLANT i

UNITS 1 AND 2 SINGLE-LOOP OPERATION l

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NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORN! A 95125 GEN ER AL h ELECTRIC L----__ -- - - _ - - - - _ _ - - _ - - - _ - - _ - -

NEDD-24344 DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the General Electnc Company nor any of the contnbutors to this document:

A Makes any warranty or representation. express or empiled, with respect to the accutacy. completeness. Or usefulness of the information contained in this docu-ment. or that the use of any information disclosed in this document may not untnnge privately owned nghts; or B Assumes any responsibtInty for Isability or damage of any kind which may result from the use of any information disclosed in thls document.

- ~ ._.

NEDO-24344 CONTENTS Page A. INTRODUCTION AND

SUMMARY

l-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 Analysis 5-1 5.1.1 Break-Spectrum Analysis 5-1 5.1.2 Single-Loop Maximum Average Planar Linear Heat Generation Rate 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 iii/iv

NEDO-24344 TABLES Table Title Page 5-1 MAPLHGR Multiplier Cases 5-4 5-2 Limiting MAPLHGR Reduction Factors 5-5 l

l

! ILLUSTRATIONS Figure Title Pjtj gi 2-1 Illustration of 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 Brunswick 1 Suction Break Spectrum Reflood Times 5-6 5-2 Brunswick 1 Discharge Break Spectrum Reflood Times 5-7 5-3 Brunswick 2 Suction Break Spectrum Reflood Times 5-8 5-4 Brunswick 2 Discharge Break Spectrum Reflood Times 5-9 5-5 Brunswick 1 Dischargu Break Spectrum Uncovered Times 5-10 5-6 Brunswick 1 Suction Break ~ Spectrum Uncovered Times 5-11 5-7 Brunswick 2 Discharge Break Spectrum Uncovered Times 5-12 5-8 brunswick 2 Suction Break Spectrum Uncovered Times 5-13 v/vi

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1. INTRODUCTION AND

SUMMARY

The current technical specifications for the Brunswick Steam Electric Plant, Units 1 and 2, do not allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service. The Brunswick 1 and 2 nuclear power plant (Technical Specification 3.4.1.1) ahall 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 documented in the Final Safety Analysis Reports and Reference 1 were reviewed for one-pump operation. Increased uncertainties in the total core flow and Traversing In-Core Probe (TIP) readings resulted in a 0.01 incremental increase in the Minimum Critical Power Ratio (MCPR) fuel cladding integrity safety limit during single-loop operation. This 0.01 increase is also added to 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 set-point equations given in the technical specifications are adjusted for one-pump operation. The least stable power / flow condition, achieved by trip-ping both recirculation pumps, is not affected by one-pump operation. Under single-loop operation, the flow control must be in master manual, since con-trol oscillations may occur in the recirculation flow control system under these conditions. Derived MAPLHGR reduction factors for single recirct lation pump operation are given in Table 5-2.

The analyses were performed as:aming the two recirculation manifolds are isolated from one another by closure of appropriate valves in the cross-tie (equalizer) line between the loops. The discharge 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 partial loss of Low Pressure Coolant Injection (LPCI) through the recirculation pump into the downcomer degrading the intended LPCI performance.

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2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Most of the uncertainties used in the statistical analysis presented in Table 4-2 of Reference 2 are independent of whether flow is provided by two loop or single loop except the core total flow and the TIP reading uncertain-ties. The one standard deviation (1-sigma) of the core total flow for single loop operation may increase to about 6% of the rated core flow from 2.5% as that for the two-loop operation. The process computer effective TlP reading uncertainty, which considers the single-loop operation TIP data noise measured lI at Browns Ferry 1, is increased to 6.8% from 6.3% for the initial cycle and to j 9.1% from 8.7% for the reload cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in the required General Electric l Thermal Analysis Basis (CETAB) safety limit MCPR.

The steady-state operating MCPR limit with single-loop operation is conserv-atively established by the multiplication of the existing K g factor and the rated flow steady-state operating limit (which should be increased by 0.01 to reflect the safety limit MCPR change). This ensures that the 99.9% statiutical limit requirement is always satisfied.

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 loop jet pumps will be backflowing. Therefore, the measured flow in the back-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.

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

~ ~

" Total Core " Active Loop Ir. active Loop -

-C

. Flow _ . Indicated Flow. . Indicated Flow.

2-1

NEDo-24344 where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to "Inac-tive 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 rppropriately modify the single-tap flow coefficient for reverse flow.* If u more exact, less conservative core flow measurement is required, special in-reactor calibration tests would have to be made. Such calibration tests would involve calibrating ,

I core support plate AP versus core flow during two pump operation along the g 100% flow control line, operating 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 Refer-ence 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 (refer to Figure 2-1):

WC"WA-WI where WC = total core flow; WA = ac tive loop flow; and WI = inactive loop (true) flow.

By applying the " propagation of errors" method to the above equation, the vari-ance of the total flow uncertainty can be approximated by:

• The expected value of the "C" coe f ficient is JO.88.

2-2

NEDD-24344 2 2 1 2 2 a 2

+ +

f 2 2 WC = og,y, l-a UW%g UW l-a Irand where cWC = uncertainty of total core flow; O = uncertainty systematic to both loops; g,

Og = random uncertainty of active loop only; l$ Og = random uncertainty of inactive loop only;

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F 1

1 OC = uncertainty of "C" coefficient; and a = ratio of inactive loop flow (W I ) to active loop flow (W A)*

Resulting f rom an uncertainty analysis, the conservative, bounding values ,

of a ,o g ,o g , and o are 1.6%, 2.6%, 3.5% and 2.8%,

C sys A '

rand rand respectively.

Based on the above uncertainties and a bounding value of 0.36 for "a," the variance of the total flow uncertainty is approximately:

2 0.36 2C = (1*6)2 1

(2.6)2 -(3.5)2 + (2.8)2' = (5.0%)2 l-0.36)2 OW 1-0.36 - -

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:

12 ojetive = (5.0%)2 } ,2 (4.1%)2 = (5.0%)2 coolant which is less than the 6% core flow uncertainty assumed in the statistical analysis.

/

2-3

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n summary, core flow during one-pump operation is determined in a conserva-tive way, and its uncertainty has been conservatively evaluated.

2.2 TIP READING UNCERTAINTY To ascertain the TIP noise uncertainty for single recirculation loop opera-tion, 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.

'l Five consecutive traverses were made with each of five TlP machines, giving a total of 25 traverses. 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.

2-4

f NED0-24344 CORE i

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t L i i "C W, A

WC = TOTAL CORE FLOW Wg = ACTIVE LOOP FLOW W, = INACTIVE LOOP FLOW Figure 2-1. Illustration of Single Recirculation Loop Operation Flows 2-5/2-6

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) 3. MCPR OPERATING LIMIT

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I 3.1 CORE-WIDE TRANSIENTS I

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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. There-fore, the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-l loop operational mode. For pressurization, flow decrease, and cold water increase transients, previously transmitted Reload / Final Safety Analysis '

Report (FSAR) results bound both the thermal and overpressure consequences of one-loop operation.

f Figure 3-1 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be seen, the consequences of the transient during one-loop operation are considerably less because of the associated reduction in operating power level.

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 than a two-pump trip from full power because of the reduced initial power I level.

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 K factors g are derived assum-ing 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; there-fore, the Kg factors derived with the two-pump assumption are conservative for single-loop operation. Inadvertent startup of aa idle recirculation pump is j not the limiting reactivity insertion transient. In addition, the restart of an idle pump would actually result in a neutron flux transient which would exceed the flow reference scram. The resulting transient with scram is 3-1 e _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ - _ _ _ _ _ _ _ _ _

HEDO-24344 l

l expected to be less severe than the worst-case cold-water transient from rated power / flow.

The latter event (loss of 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 increased subcooling of core inlet flow; 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) analy s is .

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 alarms which could occur during the course of the transient, the rod block system will stop rod withdrawal at a minimum critical power ratio which is higher than the fuel _ladding integrity safety limit.

During single-loop operation, correction of the flow-biased Rod Block Monitor (RBM) equation (below) and the lower reactor power obtainable assures that the MCPR safety limit would not be violated during the postulated RWE.

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.

3-2

1 I

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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 ef fective i

drive flow when operating with a single loop.

The two pump rod block equation is:

RB = mW + (RB100 - m(100))

f

\

I

The one pump equation becomes

} RB = mW + (RB100 - m(100)) - mow where LW = d i f f e rence, determined by utility, between two-loop and single-loop effective drive flow when the active loop indicated flow is the same; RB = power at rod block in %;

i m= fl ow re f erence slope for the RBM; W = drive flow in % of rated; and P. B100 = top level rod block at 100% flow.

If the rod block set point (RB100) is changed, the equation must be recal-culated using the new value.

The APRM trip settings are flow-biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip setting

\

discussed above.

I I

3-3

NEDO-24344 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 the rated flow steady-state limit by the Kg factor. This ensures that the 99.9% statistical limit requirement is always satisfied for any postulated abnormal operational occurrence.

3-4

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3-5/3-6 o

NEDO-24344 l

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4. STABILITY ANALYSIS I 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 recircula-tion line with one pump running at minimum speed is more stable than operation with natural circulation flow only, but is less stable than operation with i both pumps operating at minimum speed. Under single-loop operation, the flow control should be in master manual, since control oscillations may occur in ,

y the recirculation flow control system under these conditions.

i 1

4-1

NEDD-24344 12 ULTIMATE STABILITY LIMIT

- - - - - - " " " " " " " " ~ * " " " " "

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- - BOTH LOOPS, PUMPS MINIMUM SPE ED 08 -

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Figure 4-1. Decay Ratio versus Power Curve for Two-Loop and Single-Loop Operation 4-2

i NEDO-24344 I , .

5. ACCLDENT ANALYSES i

e brond' spectrum of postulated cccidentr.is covered by six categories of

(! ?

design 'sasis- events. These events are the loss-of-coolant, recirculation pump seizure',. control rod drop, main steam line break, refueling, and fuel assembly loading accidents. The analytical results for loss-of-coolant and recircula-tion pump seizure. accidents with one recirculation pump operating are givea below. The results of the two-loop analysis for the last four events are I conservatively applicabic for one-pump operation.

5.1 LOSS-OF-COOLANT ANALYSIS 5.1.1 Break-Spectrum Analysis A break spectrum analysis for each unit was performed using the model and assumptions given in Section II.A.7.3.2 of Reference 3. The suction and dis-charge break spectrum reflood times for one recirculation loop operation are compared to the standard previously performed two-loop operation in Figures 5-1 and 5-2, respectively, for Unit 1. Suction and discharge break spectrum reflood time comparisons for Unit 2 are shown in Figures 5-3 and 5-4. The uncovered time (reflood time minus recovery time) for the Unit 1 discharge and suction break spectrum and the Unit 2 discharge and suction break spectrum is compared in Figures 5-5, 5-6, 5-7, and 5-8, respectively.

For the Unit 1 standard two-loop analysis, the most limiting break was an 80%

discharge Design Basis Accident (DBA) with a total uncovered time shown in Figure 5-5 and boiling transition times ranging from 9.7 to 10.3 seconds for the two fuel types.

For the Unit 1 single-loop analysis, a boiling transition time of e 'acond is conservatively assumed for all breaks larger than 1.0 ft and the reflooding times and total uncovered times are similar to the two-loop analysis. The most limiting break for single-loop analysis is the 86% discharge DBA which has a total uncovered time of 200.6 seconds. The single-loop reflooding time

) is 233.1 as opposed to a two-loop limiting reflooding time of 235 seconds.

t 5-1 u___ _

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NEDO-24344 I

For the Unit 2 standard two-loop break-spectrum analysis, the most limiting break in also an 80% discharge DBA with a total uncovered time shown in Figure 5-7 and a boiling transition time ranging from 9.7 to 10.3 seconds for the three fuel types.

For the Unit 2 single-loop analysis, a boiling transition time of 0.1 second is conservatively assumed for all breaks larger than 1.0 ft and the reflooding times and total uncovered times are similar to the two-loc, analysis. The most limiting break for single-loop analysis is also the 80% discharge DBA which has a total uncovered time of 215.3 seconds. The single-loop reflooding time is 248.6 as opposed to a two-loop reflooding time of 248.3 seconds.

5.1.2 Single-Loop Maximum Average Planar Linear Heat Generation Rate Determination Since the reflooding time for the limiting break in single-loop operation for both Units 1 and 2 is similar to the reflooding time for two-loop operation, the procedure described in Section 11.A.7.4 of Reference 3 is conservatively applicable.

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) reduction factors were determined for the cases given in Table 5-1. The most limiting reduction factors for each fuel type are shown for both units in Table 5-2. One-loop operation MAPLHGR values are derived by uultiplying the current two-loop operation MAPLHCR values by the reduction factor for that fuel type. As dis-cussed in Reference 3, single recirculation loop MAPLHGR values are conserva-tive when calculated in this manner.

The analyses were performed assuming the two recirculation manifolds are isolated from one another by closure of appropriate valves in the cross-tie (equalizer) line between the loops. The discharga valve in the idle recircula-tion loop is normally closed, but if its closure is prevented, the suction valve in the loop should be closed to prevent the loss of LPCI flow out of a postulated break in the idle suction line.

5-2

NEDO-24344 f

( 5.1.3 Small Break Peak Cladding Temperature Section II.A.7.4.4.2 of Reference 3 discusses the small sensitivity of the calcu-lated Peak Clad Temperature (PCT) to the assumptions used in the one-pump opera-tion analysis and the duration of nucleate boiling. As this slight increase (50'F) in PCT is overwhelmingly of f set by the decreased MAPLHGR (equivalent to 300* to 500*F PCT) for one-pump operation, the calculated PCT values for small breaks will be significantly below the 2200*F cladding temperature limit speci-fled in 10CFR50.46.

[

5.2 ONE-PUMP 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 per-formed to determine the impact this accident would have on one-recirculation-pump operation. These analyses were performed with the models documented in Refer-ence 1 for a large core Bh'R/4 plant (Reference 4). The analyses were initialized f rom steady-state operation at the following initial conditions, with the added condition of one inactive recirculation loop:

thermal power = 75% and core flow = 58%, and 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 anticipated sequence of events following a recirculation pump seizure which occurs during plant operation with the alternate recirculation loop out of ser-vice is as follows:

3

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

5-3 L_ __ -_ -_ _

NEDO-24344

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 feedwater flow all exhibit transient behaviors. However, it is not anticipated that the increase in water level will cause a turbine trip and result in a scram.

It is expected that the transient will terminate at a condition of natural cir-culation and reactor operation vill continue. There will also be a small decrease in system pressure.

The minimum CPR f or the pump seizure accident for the large core BWR/4 plant was determined to be greater than the fuel cladding integrity safety limit; there fore ,

no fuel f ailures were postulated to occur as a result of this analyzed event.

These results are applicable to Brunswick Steam Electric Plant, Units 1 and 2.

Table 5-1 MAPLHCR MULTIPLIER CASES Unit Fuel Type Cases Calculated 1 8x8, 8x8R, 100% DBA Suction Break and P8x8R 100% DBA Discharge Break 80% DBA Discharge Break

• 2 7x7, 8x8, 100% DBA Suction Break 8x8R, and P87.8R 80% DBA Discharge Break *

• Most limiting break.

5-4

NEDo-24344 Table 5-2 LIMITING MAPLHGR REDUCTION FACTORS Unit Fuel Type Reduction Factor 1 8x8 0.85 8x8R and P8x8R 0.85 2 7x7 0.84 8x8 0.85 8x8R and P8x8R 0.84 f

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6. REFERENCES

1. General Reload Fuel Application, General Electric Company, August 1979

( NE DE- 2401 1 -P- A- 1 ) .

2. General Electric BWR Thermal Analysis Basis (CETAB): Data, Correlation, and Design Application, General Electric Company, January 1977 (NEDO-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 No. TVA-BFNP-TS-Il7, 0.E. Gray, III, to Harold R. Denton, September 15, 1978.

I

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l 6-1/6-2

GENER AL h ELECTRIC

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