ML20028F082

Single Loop Operation.

ML20028F082
Person / Time
Site:	FitzPatrick
Issue date:	08/31/1980
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20028F081	List: ML20028F080 ML20028F082
References
80NED285, NEDO-24281, NUDOCS 8301310130
Download: ML20028F082 (30)

Text

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! NEDO 24281

' 80N ED28's CLASS 4 I AUGUST 1980 I

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FITZPATRICK NUCLEAR POWER PLANT SINGLE-LOOP OPERATION l 1

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• i FITZPATRICK NUCLEAR POWER PLANT SINGLE-LOOP OPERATION l

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DISC: AIMER GF RESPONSIBILITY j Ynia doxer:ent was prepared by or for the General Electric Company.

Neither the General Electric Cor:pany nor any of the contributors to i this doc: cent:

l i A. Makaa any varranty or representation, express or implied, with recpect to the accuracy, completeness, or usefulness of the 1 infomation contained in this doc: cent, or that the use of any 3

infcmaticn dicolosed in this document may not infringe privately l

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i Ascumca any responsibility for liability or damage of any kind h uhich may result from the use of any infomation disclosed in

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NEDO-24281 TABLE OF CONTENTS Page l-1

1. INTRODUCTION AND

SUMMARY

2-1

2. MCPR FUEL CLADDING INTEGRITY SAFE"IY LIMIT 2-1 2.1 Core Flow Uncertainty 2.1.1 Core Flow Measurement During Single Loop 2-1 Operation 2-2 2.1.2 Core Flow Uncertainty Analysis 2-4 2.2 TIP Reading Uncertainty i 3-1 -

3. MCPR OPERATING LIMIT 3-1 3.1 Core-Wide Transients 3-2 3.2 Rod Withdrawal Error 3-4 3.3 MCPR Operating Limit 4-1 .

4. STABILITY ANALYSIS ,

5-1

5. ACCIDENT ANALYSES 5-1 5.1 Loss-of-Coolant Accident Analysis 5-1 ,

5.1.1 Break Spectrum Analysis 5-2 4 5.1.2 Single-Loop MAPLHGR Determination 1 Small Break Peak Cladding Temperature 5-2 5.1.3 5-2 5.2 One-Pump Seizure Accident 6-1 6.

REFERENCES h

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. NED0-24281 TABLES Table Title Page 5-1 MAPLilGR Multiplier Cases 5-5 5-2 Limiting MAPLHGR Reduction Factors 5-5 I

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'mi NEDO-24281 ILLUSTRATIONS Title Pg Figure Illustration of Single Recirculation Loop Operation Flows 2-5 2-1 Main Turbine Trip With Bypass Manual Flow Control 3-5 3-1 4-1 Decay Ratio Versus Power Curve for Two-Loop and Single- 4-2 Loop Operation FitzPatrick Discharge Break Spectrum Reflood Times 5-6 5-1 FitzPatrick Discharge Break Spectrum Uncovered Times 5-7 5-2 FitzPatrick Suction Break Spectrum Reflood Tines 5-8 5-3 FitzPatrick Suction Break Spectrum Unco *ered Times 5-9 5-4 l

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1. INTRODUCTION AND

SUMMARY

FitzPatrick Nuclear Power Plant The current technical specifications for the

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do not allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service. The FitzPatrick (Technical Specification 3.6.Gl shall not be operated Nuclear Power Plant 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.

reduced power with a single recirculation loop The capability of operating at in is highly desirable, from a plant availability / outage planning standpoint, d

the event maintenar.ce of a recirculation pump or other component ren ers one loop inoperative. To justify single-loop operation, the safety analyses docu- d mente'd in the Final Safety Evaluation Reports and Reference 1 were reviewe 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 This 0.01 increase is integrity safety limit during single-loop operation.

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 is not affected by one-pump operation. ,

tripping both recirculation pumps, L b

h the flow control should be in master manual, During single-loop operation, h since control oscillations might occur in tne recirculation flow control system under automat!c flow control conditions.

0.84, 0.35, and 0.84 for the 7x7, 8x8, Derived MAPLHGR reduction factors are '

and 8x8R r ue ! types, respectively.

The dis-The analyses were performed assuming the equalizer valve was closed.

if its charge valve in the idle recirculation loop is normally closed, but closure is prevented, the suction valve in the loop should be closed to prevent of a postulated the loss of Low Pressure Coolant Injection (LPCI) flow out break in the idle suction line, 1-1/1-2 I,

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NED0-24281

2. MCPk FUEL CLADDING INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the ctatistical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two racirculation pumps. Uncertainties used in the two-loop operation analysis are documented 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 As shown eingle-loop operation (compared to 2.5% for two-loop operation).

below, this value ccuservatively reflects the one standard deviation (ene sigma) accuracy of the core ilow 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-The comparable tion process computer uncertainty of 9.1% for reload cores.

The net two-loop process computer uncertainty value is 8.7% for reload cores.

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 Core Flow Measurement Durine Single Loop Operation Cl 2.1.1 l l

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-leop operation, however, the inactive 5 jet pumps will be backflcwing. Therefore, the measured ficw in the backflowing -

In jet pumps must be subtracted from the measured ficw in the active loop. i addition, the jet pump flow coefficient is different for reverse flow than for

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f orward flow, and the measurement of reverse flow must be modified to account  ;

for this difference. I f

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

fo rmula: l Inactive Loop }

Total Core} , Active Loop j - C IndicatedFlow}. '

Flow , Indicated Flow /

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NEDO-24281 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 measuremenc 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 i along the 100% flow control line, operating on one pump along the 100% flow I

control line, una calculating the correct value of C based on the core flow t derived f rom the core support plate AP and the loop flow indicator readings, f

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2.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used to establish the core flow uncertainty I

for one-pump operation is essentially the same as for two-pump operation, except for some extensions. The core flow uncertainty analysis is described j 17 Reference 2. The analysis of one-pump core flow uncertainty is summarized

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below.  !

l For single-loop operation, the total core ficw can be expressed as follows N (Figure 2-1): l

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C A ~ I i

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= total core flow;  ; y l q.

I W = active loop flow; and 3

W = inactive loop (true) flow, t 7 P>

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• The expected value of the "C" coefficient is NO.88.

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wm .h NEDO-24281 By applying the " propagation of errors" method to the above equation, the variance of the total flow uncertainty can be approximated by:

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Cy " C

+[Q1-a}l *W j2 + *C rand +[(1:a}k rand W 1 *W sys A

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where oy = uncertainty of total core flow; _;

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I oy = random uncertainty of inactive loop only; i I

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

P Resulting from an uncertainty analysis, the conservative, bounding vlaues of  ;

s and oc are 1.6%, 2.6%, 3.5% and 2.8%, respectively.

owsys, cuArand, cwyrand 6 Based on above uncertainties and a bounding value of 0.36 for "a", the variance 4 of the t tal flow uncertainty is approximately:

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4 NEDO-24281 k' hen 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 in:

2 = ( .0%)2 + j l 0 l' h l (4.1%)' = (5.0%)2 b active 1-0.12 coolant \ / j i fi t

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

f' analysis. q

! 4 In su:: nary, core flow during one-pump operation is measured in a conservative i way and its uncertainty has been conservatively evaluated. ,

2.2 TIP READING UNCERTAINTY

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To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating Bk'R. 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 syr: metric control rod pattern existed prior to the test.

Five consecutive traverses were made with each of five TIP machines, giving a i a

total of 25 traverses. Analysis df their data resulted in a nodal TIP noise h of 2.85%. Use of this TIP noise value as a component of the process computer, jf total uncertainty results in a one-sigma process computer total uncertainty .

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value for single-loop operation of 9.1% for relcad cores. >

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3. MCPR OPERATING LIMIT I

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3.1 CORE-k'IDE TRANSIENTS i hich Operation with one recirculation loop results in a maximum Therefore, power output w is 20 to 30% below that which is attainable for two-pump operation. l peration the consequences of abnormal operational transients from one- loop oop oopera-will be considerably less severe than those analyzed from a two-flow decrease and cold water increase tran- '

tional mode. For pressurization, hermal and sients, previously transmitted Reload /FSAR results bound both the t overpressere consequences of one-loop operation.

(tur-Figure 3-1 shows the consequencesAsofcan a typical pressurization transient be seen, the consequences of f bine trip) as a function of power level. d reduction one-loop operation are considerably less because of the associate in operating power level.

full The consequences from flow decrease transients are also bounded by the ,

A single pump trip from one-loop operation is less severe power analysis. d initial power than a two-pump trip from full power because of the reduce level.

Cold water increase transients can result from either recirculation pump l or introduction of colder water into the reactor vessel by speedup or restart, The K g factors are derived assuming events such a" loss of feeduater heater.

that both recirculation loops increase speed to the maximum permitted by the M-C set scoop tube position. This condition produce: the maximum possible ii d from less power increase and, hence, maximum ACPR for transients in t ate -

When operating with only one recirculation loop, than rated power and flow. d speed on only one the flew and power increase asscciated with the increase associated with both pumps increasing speed; M-C set will be less than that therefore, the K g factors 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 The resulting scram is expected to be less severe than the reference scrat. The latter event (loss of rated power / flow case documented in the FSAR.

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SEDO-24281 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 I

dependent on the initial power level. The higher the initial power level, the 1 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) i I

<,l analysis.

ll l Jd From the above discussions, it can be concluded that the transient consequence lE 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 l' are performed to demonstrate that, even if the operator ignores all instrument l 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 r.he rod block equation (see the following) and lower power assures that  ; .

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the MCPR safety limit is not violated.

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One-pump operation results in backflow through the inactive bank of jet pumps while the flow is being supplied into the lower plenum from the active bank of j l

jet piutp s . Because of the backflow through the inactive jet pumps, the present i

t rod block equation was conse natively modified for use during one-pump operation I

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because the direct active-loop flow measurement may not indicate actual flow l above about 35% drive flow without correction.

A precedure has been established for correcting the rod block equation to account for the discrepancy between actual flow and indicated flow in the active loep. This preserves the original relationship between rod block and actual effective drive flow when operating with a single loop.

3-2

NEDO-24281 The two-pump rod block equation is: i RB = mW + RB - m(100) 100 _.

The one-pump equation becomes:

= mW + R2 100 ~ 0 RB

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where aW = difference, determined by utility, between two-loop and single-loop effective drive flow at the same core flow; RB = power at rod blbck in %;

m = flow reference slope for the rod block monitor (RBM);

i W = drive flow in % of rated; and A j?

RB - t p level rod block at 100% flow.

100 .

i If the rod block setpoint (RB100) is changed, the equation must be recalculated

^l; using the new value. 3 i

The APRM trip settings are flew 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 set-ting discussed above.

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NEDO-24281 3.3 OPERATING MCPR LIMIT i 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 l l

safety limit bection 2) . At lower flows, the steady-state MCPR operating i limit is conservatively established by multiplying the rated flow steady-state i p 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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M AxiquM ONE LOOP POWER OPER ATION l I I I 1;ro 14o eso so ao too o 20 .o POWER LEVEL (% NUCLEAR SOILER RATEDI 1

Main Turbine Trip with Bypass Manual Flow Control Figure 3-1.

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NEDO-24281

4. STABILITY ANALYSIS The least stable power / flow condition attainable under for ratednormal conditions power and occurs at natural circulation with the control rods set j 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-ble than operating 2

tion line with one pump running at minimum speed is more sta is less stable than operating with both with natural circulation flow only, but

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pumps operating at minimum speed.

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1 During single-loop operation, the flow control should beflow in master manua ,

control system  :

since control oscillations might occur in the recirculation under automatic flow control conditions. 't

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NEDO-24281 1.2 ULTIMATE STASILITY LIMIT 1.0 ------ - = = = a-- - a-== - - ,- ]

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- SOTH LOOPS, PUMPS MINIMUM SPEED c.s -

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5. ACCIDENT ANALYSES l 6

The broad spectrum of postulated accidents is covered by six categories of d2aign basis events. These events are the loss-of-coolant, recirculation pump caizure, control rod drop, main steamline break, refueling, and fuel assembly l

lording accidents. The analytical results for the loss-of-coolant and recir-i culation pump seizure accidents with one recirculation pump operating are Fa 5

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

5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS I; A single-loop operation analysis utilizing the models and assumptions documented (f l

4I in Ref erence 3 was performed for the FitzPatrick Nuclear Power Plant. Using l ir

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this method, SAFE /REFLOOD computer code runs were made for a full spectrum of break sizes for the suction and discharge side breaks. The reflooding time for the single-loop analysis is similar to the two-loop analysis, and the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) curves currently applied were modified by derived reduction factors for use during one recirculation pump [i";

ope rat ion. V t.

5.1.1 Break Spectrum Analysis 'r

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A break spectrum analysis was performed using the SAFE /REFLOOD cceputer codes fi f,4 ar.d the assumptions giv1n in Section II.A.7.2.2. of Reference 3. .

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The discharge break spectrum and the suction break spectrum reflooding times for k one recirculation loop operation are compared to the standard previously per-t ormed two-loop operation in Figures 5-1 and 5-3. The uncovered tines (reflood time minus recovery time) for each break spectrum are compared in }

Figures 5-2 and 5-4  !

For the FitzPatrick Nuclear Power Plant, the reflooding time for the limiting break in standard two-loop operation is 247 seconds, occurring at 80% of the Desien Basis Accident (DBA) discharge break. The boiline transition time ranges to 10.i seconds for the three fuel types. For the sinele-loop analv+1s, B from 9 i 5-1 i

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. _ , . _ _ . _ - - m . . . . . i ,-:jf NEDO-24281 the most limiting break also occurs at the 80% DBA discharge break, and the reflooding time is 247 seconds. The uncovered time at the most limiting break is 213 seconds for the two-loop analysis and 214 seconds for the single-loop analysis.

5.1.2 Single-Loop MAPLHGR Determination The small difference in uncovered time for the limiting break size would result in a very small change in the calculated peak cladd ag temperature. Therefore, ,

as noted in Reference 3, the one- and two-loop SAFE /REFLOOD results can be con-sidered 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. y.

k MAPLHGR 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.

t One-loop operation MAPLHGR values arc 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 '

f conservative when esiculated in this manner. .

5.1.3 Small Break Peak Cladding Temperature f

Section II.A.7.4.f.2 of Reference 3 discusses the small sensitivity of the i r y

calculated peak clad temperature (PCT) to the assumptions used in the one-pump operation analysis and the duraticn of nucleate boiling. Since the slight I increase (%50*F) in PCT is overwhelmingly offset by the decreased MAPLHCR (equivalent to 300* to 500*F *. PCT) for one-pu=p operation, the calculated PCT ,

values for small breaks will be well below the 2200*F 10CFR50.46 cladding temperature limit.

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5.2 ONE-PUMP SEIZURE ACCIDENT j

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 '}

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0 one-recirculation-pump operation. These analyses were performed with the t{

models documented in Reference 1 for a large core BWR/4 plant (Reference 4).  !\i l to The analyses were initialized from steady-state operation at the following initial conditions, with the added condition of one inactive recirculttion loop. Two sets of initial conditions were assumed: g l\'

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,l (1) Thermal Power = 75% and core flow = 58% >

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'i I (2) Thermal Power = 82% and core flow = 56% ld 2 These conditions were chosen because they represent reasonable upper limits of lh 5

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q single-loop operation within existing MAPLHGR and MCPR limits at the same maximum pump speed. Pump seizure was simulated by setting the single operating 4 i pump speed to zero instantaneously. )

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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 dreps instantaneously to zero.

(2) Core volds increase which results in a negative reactivity inser- ,

tion and a sharp decrease in neutron flux. ,

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i' (3) Heat flux drops more slowly because of the fuel time constant. ]

.I (4) Neutron flux, heat flux, reactor water level, stean flow, and feed- 'j*i r

j water flow all exhibit transient behaviors. However, it is not 8 f

anticipated that the increase in water level will cause a turbine trip and result in scram.

1 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 I

decrease in system pressure.

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NEDO-24281 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.

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These results are applicable to the FitzPatrick Nuclear Power Plant. 'I,4 14 li 1,'

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Table 5-1 MAPLHGR MULTIPLIER CASES _

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Fuel Type Cases Calculated i 7x7 80% DBA Discharge Break *

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85% DBA Discharge Break p -

100% DBA Suction Break I' 8x8 80% DBA Discharge Break

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1 85% DBA Discharge Break I ,i 9 Is 100% DBA Suction Break j 8x8R 80% DBA Discharge Break * [,.

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85% DBA Discharge Break ( -

100% DBA Suction Break f;f 4

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• Most limiting break for MAPLHGR reduction factors. k

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l LIMITING MAPLHGR REDUCTION FACTORS p j k ,.,

l Fuel Type Reduction Factors f-7x7 0.84 l 8x8 0.85 8x8R 0.84 ,

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6. REFERENCES .

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

(NEDE-240ll-P-A).

Data, Correlation l.l

2. " General Electric BWR Thermal Analysis Basis (CETAB): .;

and Design Application", General Electric Company, January 1977 (NEDO-10958-A).

il

3. " General Electric Company Analytical Model for Loss-of-Coolant Analysis .

in Accordance with 10CFR50 Appendix K Amendment No. 2 - One Recirculation l Loop Out-of-Service", General Electric Company, Revision 1, July 1978 . ,

(NEDO-20566-2).

'f Enclosure to Letter #TVA-BFNP-TS-Il7, O. E. Gray III to Harold R. Denton, i .

September 15, 1978.

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