ML20010C958

Proposed Tech Specs 3/4.2,3/4.4 Changes Re Single Loop Operation.Justification for Operation W/One Recirculation Loop Out of Svc Encl

ML20010C958
Person / Time
Site:	Brunswick
Issue date:	08/12/1981
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20010C956	List: ML20010C955 ML20010C958
References
NUDOCS 8108210271
Download: ML20010C958 (32)

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8108210271 810812

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• PDR ADOCK 05000324 P

PDR 3RIT55*." ICE - ECT 2 2-4 j

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- 3/4.2 POWER DISTRIBUTION LIMITS l

3/4.2.1 AVERAGE PLdNAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR's) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not

,i exceed the limits shown in Figures 3.2.1 -1, 3.2.1 -2, 3.2.1-3, 3.2.1-4, 3.2.1 -5, 3.2.1 -6, 3. 2.1-7 or 3. 2.1-8:.

• i APPLICABILITY:

CONDITION 1, when THERMAL POWER > '25% o.f RATED THERMAL POWER.

i ACTION:

i With an APLHGR exceeding the limits of Figures 3.2.1 -1, 3. 2.1 -2, 3. 2.1 -3, 3.2.1-4, 3.2.1-5, 3.2.1-6, 3.2.1-7 or 3.2.1-8, initiate corrective action within 15 minutes and continue corrective action so that APLHGR is i

within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

I 1

SURVEILLANCE REOUIREMENTS 4.2.1 All APLHGR's shall be verified to be equal to or less than the applicable limit detere.ir.ed from Figures 3. 2.1 -1, 3. 2.1 -2, 3. 2.1 -3, 3. 2.1 -4,

3.2.1-5, 3.2.1-6, 3.2.1-7 or 3.2.1-8:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMA' POWER increase of at least 15% of RATED THERMAL POWER, and j

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

i l

• In single reactor recirculation loop operation, the APLHGR limit shall be reduced to.65 of the values specified in the above tables.

I l

BRUNSWICK - UNIT 2 3/4 2-1

i I

POWER DISTRIBUTION LIMRTS

'l 3/4.2.2 APRM SETPOINTS j

J l

LIMITING CONDITION FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (S) and rod block trip set-point (SRB) shall be established according to the following relationships:

5 1 (0.66W + 54%) T,

5 < (0.66W + 50.7%)T(Single Loop) l S

1 (0.66W + 42%) T,

SRB 1 (0.66W + 38.7%)T(Single Loop)

RD where:

S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design 7?F divided by the MTPF 1

0), and j

obtained for any class of fuel in the core (T 1

Design TPF for: P8 x BR fuel = 2.48

~

8 x 89 fuel = 2.48 7x7 fuel = 2.60 8x8 fuel = 2.45 i

APPLICABILITY:

CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

I/[

ACTION:

I With 5 or S exceeding the allowable value, initiate corrective action within 15 mS$utes and continue corrective action so that S and are B

within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POL to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i 1

i SURVEILLANCE REOUIREMENTS l-4.2.2 The MTPF for each class of fuel shall be determined, the value i

of T calculated, and the flow biased APRM trip setpoint adjusted, as required:

l a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at leest 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MTPF.

BRUNSWICK - UNIT 2

,3/4 2-10 e

TADLE 3.3.4-2 h

CONTROL R00 WITil0RAHAL BLOCK INSTRUMENTATION SETPOINTS

.55

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TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE 1.

APRH(C51-APRM-CH.A,B,C,0,E,F)

N a.

Upscale (Flow Biased)

< (0.66 W + 42%).

T* **

< (0.66 W + 42%) T* **

l b.

Inoperative fiA HIPF RA HIPF c.

Downscale

>- 3/125 of full scale

> 3/125 of full scale d.

Upscale (Fixed)

T 12% of RATED TilERMAL POWER 712% of RATED TilERMAL POWER 2.

R00 BLOCK MONITOR (C51-RBM-CH.A,8)

T* ***

< (0.66 W + 39%)

T* ***

l a.

Upscale

<(0.66W+39%)RTPT b.

Inoperative WA RA HIPF c.

Downscale

> 3/125 of full scale

> 3/125 of full scale s"

3.

SOURCE RANGE MONITORS (C51-SRM-K600A,B,C,0) w b

a.

Detector not full in NA NA 5

< 1 x 10 cps

< 1 x 10 cps b.

Upscale c.

Inoperative RA NA

~

d.

Downscale

> 3 cps

> 3 cps 4.

INTERMEDIATERANGEMONITORS_(C51-IRH-K601A,B.C,0,E,F,G,il) a.

Detector not full in NA NA b.

Upscale

< 108/125 of full scale

< 100/125 of full scale c.

Inoperative NA NA d.

Downscale

> 3/125 of full scale

> 3/125 of full scale

• T=2.60 for 7.x 7 fuel.

T=2.45 for 8 x 8 fuel.

T=2.48 for 8 x 8R fuel.

T=2.48 for P8 x BR fuel.

• • When in single loop, trip setpoint and allowable value shall be reduced to < (.66W + 38.7%)M hours in single recirculation loop operation.

T

• within 24

• • • When in single loop, trip setpoint and allowable value shall be reduced to < (.66W + 33.7%) T

• within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in single recirculation loop operation.

MIPF

O l

REACTOR COOLANT SYSTEM 3/4.4 l

3/4.4.1 RECIRCULATION SYSTEM i

l RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 With two reactor coolant recirculation loops in operation, the cross-tie valve shall be closed, the pump discharge valves shall be OPERABLE and the pump discharge bypass valves shall be OPERABLE or closed.

With only one reactor coolant recirculation loop in operation, the cross-tie valve shall be closed and the pump discharge valves on the operating loop shall be OPERABLE and the operating loop pump discharge bypass valves shall be OPER/.0LE or closed. For the inoperable loop, the valves shall be either OPERABLE or closed and de-energized. Thermal power shall be limited to no more than 50% within 15 minutes and the setpoints reduced as per Tables 2.2.1-1, 3.3.4-2 and Sections 3.2.1 and 3.2.2.

APPLICABILITY:

CONDITIONS 1* and 22 ACTION:

With both recirculation loops not in operation, operation may continue; restore at least one loop to operation within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l l

l SURVEILLANCE REOUIREMENTS l

l 4.4.1.1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each COLD SHUTDOWN which exceeds 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if not per-formed in the previous 31 days.

4.4.1.2 Each pump discharge bypass valve, if not OPEPAGLE, shall be verified to be closed at least once per 31 days.

• See Special Test Exception 3.10.4.

i BRUNSWICK - UNIT 2 3/4 4-1 l

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JUSTIFICATION FOR THE OPERATION OF A GENERAL ELECTRIC JET PUMP BOILING WATER REACTOR WITH ONE RECIRCULATION LOOP OUT OF SERVICE 1

1 i

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BRUNSWICK STEAM ELECTRIC PLANT 4

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JUSTIFICATION FOR THE OPERATION OF BRUNSWICK STEAM ELECTRIC PLANT UNIT NO. 2 WITH ONE RECIRCULATION LOOP OUT OF SERVICE 1.0 Introduction The recirculation pump on Brunswick Unit 2 Loop "A" experienced a 1

step increase in shaft vibration level early on August 11, 1981.

l No cause was apparent. Attempts were made (August 11) to contact the pump manufacturer to discuss the vibration levels and make decisions regarding continued service of this pump.

Since we could not reach the manufacturer expeditiously, we determined that shutting down 1

the pump would be prudent; however we have kept it in service because 4

Brunswick Unit 1 is out of service due to turbine bearing repairs and lube oil flushing until the end of September, and H. B. Robinson Unit 2 is also out of service with steam generator tube problems until early September.

4 4

If the Brunswick Unit 2 Loop "A" recirculation pump were shut down, Technical Specifications (3.4.1.1) require that the unit be shutdown if the idle recirculation loop is not returned to service within 9

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4 CP&L's nuclear units normally supply a large portion of the system i

load, especially during peak periods such as late summer.

Since Brunswick Unit 1 and Robinson Unit 2 are out of service now due to.,

l I

forced outages, CP&L is having to replace this capacity by resorting f

to much more expensive generation and to purchasing power and capacity I

from neighboring utilities to meet demand.

If Brunswick Unit 2 were also tcken out of service, the impact on our customers would be severe.

For these reasons, CP&L is continuing to operate the Brunswick Unit 2 Loop "A" recirculation pump in order to keep the unit on-line. However, we are requesting emergency relief from Technical Specifications to allow single loop operation so that the unit can be kept in service at power levels up to 50%.

General Electric Company has provided analyses to support this request 1

as outlined herein.

l l

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2.0 Special Operating Conditions for Single Loop Operation In order to ensure operation of this derated condition is in accordance with the assumptions utilized by GE, Carolina Power & Light Co. commits to the I

following conditions during normal operation.

I 1.

The recirculation pump motor generator set field breaker on the inoperable loop will be pulled' and placed under clearance to preclude operation of the pump or injection of a cold slug into the vessel.

This l

will be done unless an attempt to re-start the pump is underway.

l l

2.

R circulation pump suction and discharge valves on the idle loop may be

.ef t open to enable flow to prevent the loop-to-loop suction &T from exceeding the allowable value for an idle loop start.

This can be done j

due to the BSEP LPCI injection logic which automatically closes both l

discharge valves on a LPCI initiation signal thus assuring that the LPCI injection is directed into the vessel.

If the pump conditions which required the shutdown of one loop are such that there are no good prospects of repair in a short period of time while the unit remains on-line, then the recirculation pump suction ar.d discharge valves on the idle loop will be shut and the valve motor breakers opened and placed under clearance.

3.

The r: circulation controls will be placed in the manual mode, thereby eliminating the need for control system analyses.

4.

The settings for the rod block monitor, APRM rod block trip, and flow bias scram will be modified as necessary to provide for single loop operation.

t

~3-

5.

Administrative Controls in addition to technical specifications restricting pump startup will prevent startup of the pump in the idle loop.

6.

MAPLHGR will be limited to 0.6$ of the rated flow (two loop) limit.

7.

The limitation on power level as described in FSAR Section 14.3.6.2 is 65 percent.

Carolina Power & Light will further limit the power level to 50%.

8.

The safety limit MCPR must be increased by a value of 0.01, hence the rated flow (two loop) MCPR operating limit would be increased by 0.01.

1 This new " rated flow" MCPR operating limit will' then he increased by the i

appropriate Kg factor to determine the MCPR operatink limit at reduced l

I flow.

I I

l I

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I l.

3.0 MAPLHGR Adjustment Factor GE has performed a large number of single loop analyses for similar plants; in no case has a multiplier of less than 0.70 been required.

Therefore until the plant specific calculations can be verified (as required by 10.CFR.20), it is proposed that a multiplier of 0.65 be conservatively applied for single loop operation.

.l 1

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4.0 Other Considerations for Single-Loop Operation Various conditions have been examined for the impact of single-loop l

operations.

The following pages address several issues including:

A.

One pump seizure accident B.

Abnormal Operational. Transients 1.

Transients and Core Dynamics 2.

Rod Withdrawal Error 3.

APRM Trip Setting 4.

K Curves f

C.

Stability Analysis D.

Thermal - Hydraulics 1

1 i

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=**

ONE-PUMP SEIZURE ACCIDENT The pump seizure event is a very mild accident in relation to other accidents such as the LOCA.

This has been demonstrated by analyses in Reference 2 for the case of two pump operation, and that it is also true for the case of one pump operation is easily verified by consideration of the two events.

In both accidents, the recirculation driving loop flow is lost extremely rapidly; in the case of the seizure, stoppage of the pump occurs; for the LOCA, the severance of the line has a similar, j

but more rapid and severe influence.

Following a pump seizure event, natural circulation flow continues, water level is maintain 3d, the core remains submerged, and this provides a continuous core cooling mechanism.

However, for the LOCA, complete flow stoppage occurs and the water level decreases due to loss-of-coolant result'ig in uncovery of the reactor core and subsequent overheating of the fuel rod cladding.

In addition, for the pump seizure accident, reactor pressure does not decrease, whereas complete depressurization occurs for the LOCA.

Clearly, the increased temperature of the cladding and reduced reactor pressure for the LOCA both combine t yield a much more severe stress and potential for cladding perforation for the LOCA than for the pump seizurc.

Therefore, it can be concluded that the potential effects of the hypothetical pump seizure accident are very conservatively bounded by the effects of a LOCA and specific analyses of the pump seizure are not required.

ABNORMAL OPERATIONAL TRANSIENTS TRAN0IENTS AND CORE DYNAMICS I

Since operation with one recirculation loop results in a maximum power output which is 20 to 31% below that wnich can be attained for two-pump operation, the consequences of abnonnal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-loop operational mode.

For pressurization, flow decrease, and cold water increase, transients previously transmitted for Reload /FSAR results bound both the thermal and overpressure consequences of one-loop operation.

Figure 1 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level.

As can be seen, the consequences of one-loop operat:on are considerably less because of the associated reduction in operating power level.

t The consequences from ficw decrease transients are also bounded by the full power anal Jis.

A single pump trip from one-loop operatica is obviously 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 speed-up or introduction of colder water into the reactor vessel by.

events such as loss of feedwater heater.

For the former, the K factors f

are derived assuming that both recirculation loops increase speed to the maximum permitted by the M-G Set scoop tube position set screws.

This condition produces the maximum possible power increase and hence maximum AMCFR for transients initiated from isss than rated power and flow.

When operating with only one rec;rculation loop, the flow and power increase associated with the increased speed on only on M-G Set will be less than that associated with both pumps increasing speed, and therefore, I

the K factors derived with the two pump assumption are conservative for f

single-loop operation.

For the latter, the loss of feedwater heater event is generally the most severe cold water increase event with respect to increase in core power.

This event 's caused Dy positive reactivity insertion from core flow inlet subcooling; therefore, the event is independent of two pump or one pump operation.

The severity of 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 signifi-cantly 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 i

full power aaalysis.

The maximum power level that can oe attained on one-loop operation is only restricted by the MCPR and overpressure limits established from a full power analysis.

R0D WITHDRAWAL' ERROR The rod withdrawal error at rated power is given in reload licensing submittals.

These analyses demonstrate that even if the operator ignores all indications and alarm which could occur during the course of the transient, the rod block system will stop rod withdrawal at a critical power ratio which is higher than the 1.07 safety simit.

The MCPR requirement for one pump operation will be equal to that for two pump operation because the nuclear characteristics are independent of whether the core flow is attained by one-or two-pump operation.

The only exceptions to J

this independence are possible flow assymmetries which might resuit from one pump operation.

Flow assymmetries are shown to be of no concern by tests conducted at Quad Cities.

Under conditions of one pump operation and equalizer valve closed, flow was found to be uniform in each bundle.

One pump operation results in backflow through 10 of the 20 jet pumps 4

while the flow is being supplied into the lower plenum from the 10 active jet pumps.

Because of the backflow through the inactive jet J

pumps, the present rod block equation shown in the Technical Specifi-cation must be modified.

The procedure for modifying the rod block equation for one pump operation is given in the following subsections.

The two pump rod block equation in the existing Techniccl specifi-a.

c.ation is of the form:

RB = (mW + K)%

(1) where RB = power at rod block in %

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

W = drive flow in % of rate' K = power at rod block in % when W = 0 For the case nf top level rod block at 100% flow, denoted RB100; RB100 = m(100) + K or l

K = RB

- m(100) 100 Substituting for K in Equation 1, the two pump equation becomes:

RB = mW + [RB

- m(100)]

(2) 100 b.

Next, the core flow (F ) versus drive flow (W) curves are determined e

for the two pump and one pump cases.

For the f.wo pump case the core flow and drive flow are aerived by measuring the differential pressures in the jet pumps and recirculation loop, respectively.

Core flow for one pump operation must be corrected for the backflow through the inactive jet pumps thus:

Actual core flow (one pump) = Active jet pump flow - inactive jet pump flow. -

Both the active and inactive flows are derived from the jet pump differential pressures.

The drive flow is derived from the differ-

)

ential pressure measurement in the active recirculation loop.

1 These two curves are plotted from a BWR data in Figure 2.

The maximum difference between the one pump and two pump core flow is determined graphically.

This occurs at about 35% drive flow which is denoted W.

c.

Next., a horizontal line is drawn from the 35% drive flow point on the one pump curve to the two pump curve and the corresponding flow, W, is determined.

Thus, AW = W -W' 2

y 2

The rod block equation corrected for one pump flow is:

RB = mW + [RB m(100)] - ARB 100 where ARB = RB - RB y

2*

RB = mW + RB

- m(100 + AW)

(3) 100 d.

For BSEP application, the constants from the Technical Specification are:

m = 0.66 RB100 = 108 From Figure 2:

AW = W -W2 = 35 - 30 = 5 y

i

Evaluating in Equation 3, the one pump rod block equation becomes:

RB = 0.66W + 108 - 0.66(100 + 5) = 0.66W + 38.7 (4)

't This line is depicted in Figure 2 as the future corrected rod block line for one pump operation.

i APRM TRIP SETTING i

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

i

I t

i K CURVE f

1 For single recirculation loop operation, the K curve contains sufficient f

4 conservatism to provide operational limits such that the fuel integrity safety limit is not violated for abnormal operational events.

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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.

Operation along the minimum forced recirculation i

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

The core stability along the forced circulation, rated rod pattern line for single loop operation is the same as that for both loops operable except that rated power is not attainable.

Hence, the core is limited to maximum power for single pump operation and only manual flow control should be used.

This is i

illustrated in Figure 3.

1 i

j

,. I

THERMAL-HYDRAULICS Except for total core flow and TIP reading, the uncertainties ussd in the statistical analysis to determine the MCPT, fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two recirculation pumps.

Uncertainties used in the two-loop operation analysis are documented in Table 5-1 of Reference 1 for reloads.

A 6% core flow measurement ur.tertainty has been established fc" single-loop operation (compared to 2.5% for two-loop operation).

As shown in Appendix A, 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 Appendix B.

This revision resulted in a single-loop operation process computer uncertainty of 9.1% for reload cores.

A comparable two-loop process computer uncertainty value is 8.7% for reload cores.

The net effect of the revised core flow and TIP uncertainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit, and therefore a similar increase in the " rated flow" MCPR operating limit.

The steady-state operating MCPR with single-loop operation will be conservatively established by multiplying the K factor to the revised f

rated flow MCPR limit.

This ensures that the 99.9% statistical limit requirement is always satisfied.

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REFERENCES t

1.

NED0-20566-2, Revision 1, GE Analytical Model for LOCA Analysis in Accordance with 10CFR50 Appendix K, Amendment No. 2 - One Recircula-tion Loop Out of-Service j

2.

Generic Reload Fuel Application, General Electric Company, August 1979 (NEDE-24011-P-A-1) 3.

General Electric BWR Thermal Analysis Batis (GETAB):

Data, Correia-tion, and Design Application, General Electric Company, January 1977 (NEDO-10958-A) 1 f

I 4

APPENDIX A UNCERTAINTIES IN TOTAL CORE FLOW FOR i

SINGLE LOOP OPERATION 1.

CORE FLOW MEASUREMENT DU. RING 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 and the total core flow is the sum of the indicated loop flows.

However, for single loop operation, the inactive jet pumps will be backflowing, so 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 in reversa flow than forward flos, and the measurement of reverse flow must be modified to account for this dit arence.

]

For single loop operation the total core flow should be measured by the following formula:

(TotalCore Active Loop Inactive Loop

-C Flow Indicated Flo Indicated Flow 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 up to indicate for.vard flow correctly.

A-1

The 0.95 factor is 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 could be made.

Such calibration tests may involve calibrating core support plate AP versus core flow during two p ap operation along the 100% flow control line, then 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, along with the loop flow indicator readings.

2.

CORE FLOW UNCERTAINTY ANALYSIS i

i 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 uncer-i tainty analysis is described in Reference 3.

The analysis of one pump core flow uncertainty is summarized below.

t For single-loop operation, the total core flow can be expressed as l

follows (refer to Figure A-1):

4 WC* A I

l

• The expected value of the "C" coefficient is s0.88.

i 1

A-2

I' where WC = total core flow; WA = active loop flow; and WI = inactive loop (true) flow.

By applying the " propagation of errors" method to the above equation, 4

the variance of the total flow uncertainty can be approximated by:

i 1a 1a

+

C sys A

y rand rand f

where i

i

{

"WC uncertainty of total core flow;

=

i "W

uncertainty systematic to both loops;

=

sys i

W

=

A random uncertainty of active loop only;

}

rand UWy random uncertainty of inactive loop only;

=

i rand C

=

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

=

y (W )'

A i

j Resulting from an uncertainty analysis, the conservative, bounding values of W W

W and "C are 1.6%, 2.6%, 3.5%, and 2.8%,

sys' Ag' I g respectively.

i i

A-3

4 Based on the above uncertainties and a bounding value of 0.36 for "a",

the variance of the total flow uncertainty is approximately:

= (1.6)2 (2.6)2 1b6 (3.5)2 + (2.8)

= (5.0%)2

+

o 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:

\\

(5.0%)2 y!gf2 (4. E ) = (5.0%)2

+

etive 12 coolant which is less than the G% core flow uncertainty assumed in the statistical analysis.

In summary, core flow during one pump operation is determined in a conservative way, and its uncertainty has been conservatively evaluated.

1

?

i A-4

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I CORE j

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i 4

4 l

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i "C

W, f

"A I

OTAL CORE FLN W

=

C W,

ACTIVE LOOP FLOW I

W, INACTIVE LOOP FLOW l

i i

Figure A -1.

Illustration of Single Recirculation Loop Operation Flows l

I A-5 l

e a

APPENDIX B TIP READING UNCERTAINTY FDR SINGLE LOOP OPERATION 1

5 To ascertain the TIP nnise uncertainty for single recirculation loop operation, a test was parformed at an operating BWR.

The test was performed at a poaer level 59.3% of rated with a single recirculation pump in operation (core flow 46.3% of rated).

A rotationally symmatric control rod pattern existed prior to the test.

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

Analysis of these 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.

(

l

'l 9

1 t

B-1

....