Text

Revised Tech Specs 3/4.2 Re Power Distribution Limits.Util Justification Encl
	ML20003C952
Person / Time
Site:	Brunswick
Issue date:	03/13/1981
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20003C938	List: ML20003C937; ML20003C952;
References
	NUDOCS 8103180882
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O 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR 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 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 or 3.2.1-6.* l APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 or 3.2.1-6, initiate corrective action within 15 minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGR's shall be verified to be equal to or less than the appli-l cable limit detennined from Figure 3.2.1 -1, 3. 2.1 -2, 3.2.1-3, 3. 2.1-4, 3.2.1-5 or 3.2.1-6:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of l at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN ter APLHGR.

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In single reactor recirculation loop operation, the APLHGR limit shall be reduced to .65 of the values specified in the above tables. This portion of the Specification expires on March 15, 1981.

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l BRUNSWICK-UNIT 1 3/4 2-1 G10818 03%>

3/4.2.2 APRM SETPOINTS 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 relationship:

S 1 (0.66W + 54%) T S < (0.66W + 50.7%)T(Single Loop)*

S RB 1 (0.66W + 42%) T S RB 1 (0.66W + 38.M)T(Single Loop)*

where: S and S are in percent of RATED THERMAL POWER.

DB W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF -

obtained for any class of fuel in the core (T 5 1.0),and Design TPF for: 8 x 8 fuel = 2.45.

8 x 8R fuel = 2.48.

P8 x 8R fuel = 2.48.

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

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and SRB are within the required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS j 4.2.2 The MTPF for each class of fuel shall be determined, the value of T l calculated, and the flow biased APRM trip setpoint adjusted, as required:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 bnurs when the reactor is operating with a LIMITING CONTROL .50D PATTERN for MTPF.

This portion of the specification expires on March 15, 1981.

BRUNSWICK-UNIT 1 3/4 2-8 1

TABLE 3.3.4-2 c0NTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS ,

TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE N -

1. APRM (CSI-APRM-CH. A,B,C,D,E,F)

E a. Upscale (Flow Biased) -< (0.66W + 42%) T* **

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

7 MIPF MIPF E b. Inopera tive NA NA U c. Downscale > 3/125 of full scale > 3/125 of f 1 scale a d. Upscale (Fixed) i12%ofRATEDTHERMALPOWER i12%ofRATEDTilERMALPOWER

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

a. Upftcale 1(0.66W + 41%) _ T* *** 1 (0.66W + 41%) T* ***

MIPF MIPF

b. Inopera tive NA NA

c. Downscale > 3/125 of full scale > 3/125 of full scale

3. SOURCE RANGE MONITORS (C51-SRM-K600A,B,C,D) us a. Detector not full in NA 5

NA 5

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" b. Upscale < 1 x 10 cps < 1 x 10 cps

c. Inoperative RA RA

d. Downscale > 3 cps > 3 cps

4. INTERMEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,0,E,F,G,H)

a. Detector not full in NA NA

b. Upscale < 108/125 of full scale < 108/125 of full scale

c. Inoperative RA RA

d. Downscale > 3/125 of full scale

> 3/125 of full scale

T=2.43 for 8x8 fuel T=2.48 for 8x8R fuel T=2.48 for P8x8R fuel

- When in single loop, trip setpoint and allowable value shall be reduced to < (.66W + 38.7%) - T

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in single recirculation loop operation. This Specification expires oii March 15,1981. NTPT

. ***When in single loop, trip setpoint and allowable value shall be reduced to < (.66W + 35.7%) T

within 24 -

hours in single recirculation loop operation. MIPF This Specification expires on March 15, 1981.

3/4.4 REACTOR COOLANT SYSTEM 3/a.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant recirculation loops shall be in operation with the cross-tie valve closed, the pump discharge valves OPERABLE and the pump discharge bypass valves OPERABLE or closed.

APPLICABILITY: CONDITIONS 1* and 2*.

ACTION:

With one or both recirculation loops not in operation, operation may continue; restore both loops to operation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .** l SURVEILLANCE REQUIREMENTS 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if not perforned in the previous 31 days.

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

See Special Test Exception 3.10.4.

- Until March 15, 1981, with one recirculation pump not in operation, reduce l

power to less than 50% and reduce the setpoints as specified in Table 2.2.1-1, l Table 3.3.4-2, Section 3.2.1, and Section 3.2.2, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both l recirculation loops not in operation, operation may continue; restore at least one loop to operation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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BRUNSWICK-UNIT 1 3/4 4-1

JUSTIFICATION FOR THE OPERATION OF BRUNSWICK STEAM ELECTRIC PLANT WITH ONE RECIRCULATION LOOP OUT OF SERVICE 1.0 Introduction The technical Specifications for the Brunswick Steam Electric Plant (BSEP) (3.4.1.1) require that the plant be shutdown if an idle r :irculation loop cannot be returna.d to service within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At 1300 on 3/5/81, the Unit No. 1 "A" Loop Recirculation Pump Motor Generator Set field and relay alarm annunciated. An investigation of the Motor Generator Set found the field brushes arcing and a power reduction was commenced. ' At 1404 the "A" Loop Recirculation Loop Pump Motor Generator tripped on field under voltage. The cause for the cator generator failure was not known and a Technical Represenative was summoned to the site. It was expected that the Motor Generator Set could be repaired within approximately 10 days unless major repairs were required.

BSEP Unit Nos. I and 2 normally supply a large portion of the system load due to schedulee, ,otages of the fossil units. BSEP Unit No. 2 commenced an unscheduled outage'due to a snubber problem on 3/4/81 which will last approximately.2 - 3 weeks. This loss of capacity plus the loss of Unit No. I would have a severe impact on our customers.

For continuation of operation with one loop out of service, Carolina Power & Light Company contracted GE to provide an analysis as outlins!..

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 following conditions during normal operation.

1. The recirculation pump motor generator set field breaker will be pulled and placed under clearance to preclude operation of the pump or injection of a cold slug into the vessel. This will be done in order to run the motor generator set to allow " turning" the slip rings to repair damage.

2. Recirculation pump suction and discharge valves on th* idle loop will be left open to enable flow to prevent the loop-to-loop suction AT from exceeding the allowable value for an idle loop start. This can be done due. to the BSEP LPCI injection logic which automatically closes both discharge valves on a LPCI initiation signal thus assuring that the LPCI injection is directed into the vessel.

3.. The recirculation controls will be placed in the manual mode, thereby l'

eliminating the need for control system analyses.

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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.

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

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6. MAPLHGR will be limited to 0.65 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.

This new " rated flow" MCPR operating limit will then be increased by the appropriate Kg factor to determine the MCPR operating limit at reduced flow.

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3.0 MAPLHGR Adjustaent Factor i,

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

i Therefore, until the plant specific calculations can be verified (as required by 10CFR20), it is proposed that a multiplier of 0.65 be conservatively applied'for single loop operation.

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4.0 -Other Conside' rations for Sincle-Looo Ooeration Various conditions have been examined for the impact of single-loop ope ra tions . The following pages addrets 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. X f Curves C. Stability Analysis D. Thermal - Hydraulics I

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ONE T 'P SEIZURE *CCIDENT The pump seizure event is a very mild accident in relation to other accidents such as the LOCA. Tais has beer 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, tha recirculation driving loop flow is los: extremely rapidly; in the case of the seizure . stoppage of 'he pump occurs; for the LOCA, :ne severance cf the line nas a similar, but more rapid and severe influence.

.ollowing a pump seizure event, natural circulation flov linues, water level is maintained, the core remains submerged, and this pr.< ides a continuous core cooling mechanism. However, for the LOCA, complete flow stoppage occurs and the water level decreases due to loss-of-coolant resulting in uncovery of the reactor core and subsequent overheating of the fuel red cladding. In addition, for the pump seizure accident, reactor pressure does not decrease, wherear complete depressurization occurs for the LOCA. Clearly, the increased temp-erature of the cladding and reduced reactor pressure for the LOCA both combine to yield a much more severe stress and potential for cladding perferation for the LOCA than for the pump seizure. 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.

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ABNORM L OPERATIONAL TRANSIENTS TRANSIENTS AND CORE DYNAMICS Since ::eration wi:n one recirculation loco results in a maximum power output whicn is 20 to 30", below that which can be attained for : o-pump operation,

ne cm.equences of abnor .al operational transients from one-loop operation will be consideraciy less severe than those analy:ed from a two-loop operational mode.

For pressurization, flow cecrease, and cold water increase, transients previcusly transmitted for Reicad/FSAR results bound both the thermal and everpressure conse-quences'of ane-loco cperation. Figure 1 shows the consequences of a typical pressurizaticn transient (turbine trip) as a function of powcr level. As can be seen, the consequences of one-loop cperation are considerably less because of the associated reduction in operating power level.

The consecuences from ficw decrease transients are also bounded by the full power analysis. A single pump trip frca one-loop operation is obviously less severe than a two-pump trip frot .ull powerbecause of the reduced initial power level. .

Cold water increase transients can result from either recirculation pump speed-up or introducticn of colder water into the reactor vessel by events such as loss of feedvater heater. For the former, the X ffactors 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 maxi.mm possible power increase and hence maximum c.Mr0R for transients initiated frcm less than 'ratec power and ficw. When operating with only cne recirculation loop, the flow and power increase associated with the increased speed on only m .

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cne M-G Set will be less than that a:sociatec witn both ;umas increasing s eed, and therefore, the Xf factors derived with :ne two ;uro assumpti:n are c:n- .

servative for single-iccp operation. For the latter, the loss of feedwate,r heater event is generally the most severe c 10 water increase even with re-spe:: to increase in Ore power. Inis even is causec by positive reactivity insertion from core flow inle: subcocling; nerefore, ne even- is independent of two-pu=o or one-:uro c;eration. The severity of the event is crimarily cependen: On the initial power level. The higner the initial p wer level, the greater the CpR change during the transien:. Since the initial :cwer level durine one-pumo :peration will be significantly lcwer, the One-pump c 1d water increase ~ case is conservatively bounced by the full power (two-puto) analysis.

From :ne above discussions, it can be concluded Inat :ne transien: consequence from one-iccp cperation is bounded by previcusly submitted full power ar.alysis.

The maximum power level that can be attained en one-locp c:eration is eniy restricted by the MCPR and overp -:;ure limits established from a full pcwer analysis.

ROD WITHDPAWAL ERROR The rod withdrawal error at rated pcwer is given in reload licensing sub=ittals.

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

ne 1.07 safety limit. The .YCFR require:ent for ene-pu=p operatien will be equal to that for two-pump coeration because the nuclear characteristics are indepen-dent of whether the core flew is attained by one- or two-pu=p c:eration. The Only exceotions to this independence are possible ficw asymmetries wnich might result from ene-pump cperation. Flew asymmetries are shcwn to be of no concern by tests conducted at Quad Cities. Unde conditions of one-pump cperation and

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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 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 shown in the Technical Specification must be modified.

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

a. The two-pump rod block equation in the existing Technical Specification is of the form:

RB = (mW + K)% (1) v here RB = power at rod block in %

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

W = drive flow in % of rated K = power at rod block in % when W = 0.

For the case of top level rod block at 100% flow, denoted RB 100 I

.RB100 , m(100) + K or X = RB100 - m(100)

Substituting -for K in Equation 1, the two pump equation becomes:

RB = mW + (2)

[RB100 - m(100)]

b. Next, the core flow (F c) versus drive flow (W) curves are determined for the two-pump and one-pump cases. For the two-pump case the core flow and drive flow are derived 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'

je2 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 differential pressure measurement in the active recirculation loop. 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,4W = W) - W2 -

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The rod block equation corrected for one pump flow is:

RB = mW +

(RB100-m(100))-6RB where ARB = RB) - RB2 " " AW .

_ RB = mW + RB100 - m(100 + AW) (3)

d. For BSEP applicaticn, the constants from the Technical Specification are:

m = 0.66 RB100 = 108 From Figure 2:

AW = W) -W2 = 35 - 30 = 5 Evaluating in Equation 3, the one-pump rod block equation becomes:

RB = 0.66W + 10B - 0.66(100+5) = 0.66W + 38.7 (4)

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This line is depicted in Figure 2 as the future corrected rod block line for one-pump creration.

A?RM TRI? SETTING 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 _ciscussed above.

KfCURVE For single recirculation loop operation, the Kf curve contains sufficient conservatism to provide operational limits such that the fuel integrity safety limit is not violated for abnormal o, : rational events.

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s STABILITY ANALYSIS The least stable power / flow condition attain ble uncer normal conditions occurs at natural :irculation witn the control reds set for rated power and flow.

This conci:icn may be reached fo~ lowing the trip of both recirculation pumps.

Operation along the minimum forced rectrculation line with one : ump running a: minimum speed is more stable than operating with natural circulation fi:w only, but is less stable than operating with both pumos 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 maxim.um power for single pump operation and only manual flow control should be used. This is illustrated in Figure 3. .

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THERMAI.-HYDRAULICS Except for total core 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 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 uncertainty has been established for 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 Refe:rence 2. The random noise component of the TIP reading uncertainty was zevised for single recirculation

. . loop operation to reflect the operating ple.nt 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 l

increase in " rated flow" MCPR operating limit.

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I conservatively established by multiplying the Kg factor to the revised rated flow MCPR limit. This ensures that the 99.9% statistical limit requirement is i

always satisfied.

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REFERENCES

1. NEDO-20566-2, Revision 1, GE Analytical Model for LOCA Analysis in Accordance with 1C CFR 50 Appendix K Amendment No. 2 - One Recirculation Loop Out-Of-Service

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2.

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

3. General Electric BWR Thermal Analysis Basis (CETAB: Data, Correlation and Design Application, General Electric Company, January 1977 (NEDO-10958-A))

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A'DENDIX A UNCERTAINTIES IN TOTAL CORE FLOW FOR .

SINGLE LOOP OPERATION 1 CORE FLOW MEASUREMENT DURING SINGLE L;JP OPERATION The jet r ump 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 ir.dicated loop flows. However, for single loop operation, the inactive jet pumps will be backflowing, so the measured flow in the backflow-ing jet pumps must be subtracted from the measured flow in the active loop.

In addition, the jet pump flow coefricient is different in rever se flow than

, forward flow, and the measurement of reverse flow must be modified to account for th'is difference. -

Fo single loop operation,the total core flow should be measured by the follow-ing formula:

Total Core , Active Loop Inactive Loop Indicated Flow -C Indicated Flo 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 forward flow correctly.

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 csp versus cere flow during two pump operation along the 100% flow con-trol 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 co 6 support platai1P, along with the loop flow indicator readings.

Note: The expected value of the "C" coef ~icient is ~0.88.

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2. CORE FLOW UNCERTAINTY ANALYSIS The uncertainty analysis procedure used to establish the core flow encertainty for one-pump operation is essentially the same as for two pump operation, except for se=e extensions. The core flow uncertainty analysis is described in Refer-ence 3 The analysis of one pump core flow mcertainty is smrized below.

For single-loop operation, the total core flow can be expressed as follows (Refer to Figure A-1):

WC*WA~II where WC = total core flow; l WA = active Icop flow; and Ut = inactive loop (true) flow.

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

yC .ysys 7 1 12 2 *,,3 2 f2 21 l

1 h% Ard l-a k% Irand where

= uncertaioty of total core flow;

%c E,g = uncertainty systematic to both loops; og = randos uncertainty of active loop only; l

Og = random uncertainty of inactive loop only; Fe = uncertainty of "C" coefficient; and i

a =

ratio of inactive loop flow (WI ) to active loop flow (WA)-

A-2

Resulting f rom an uncertainty analyala, the conservative, bounding values of %, ,,, % ,% * ' * * * * * '

Irand roopectively.

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

2 2 f 1 1

/,1-0.36; 0.% \ g),$)2 + (2.8)2' (5.0%)2, "c = (l.612 g.2 (2.612 =

L1-0.367 - .

Wn the effect of 4.11 core bypass flow split uncertainty at 12I (bounding case) l bypcss flow fraction is added to the above total core flow uncertainty, the active coolanc flow uncertainty is:

2 g2 , . (5,gg)2 t1-0.12 0.M \ C4 lg)2 = (5.01)2 I

acttwe i j coolanc which is less than the 6% core flow ancertainty assumed in the statistiest analysis.

l l

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

l l

l I

A-3 l

l

i l e cons L 2 I

NC W1 WA VC -

Total Core Flow -

W -

Active Loop Flow A

Wy -

Inactive Loop Flow Figure A-1. Illustration of Sl*E e ltecirculation Loop Operation Flows A-4

_ , _ . . ~ . , . . - -. . . - n

APPENDIX B TIP READING UNCERTAIhTi FOR SINGLE LOOP OPERATION To ascertain the TIP noise uncertainty for single recirculation op operation, a test was performed at an operating 3W1. 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 machin.?s, giving a total of 25 traverses. Analysis of their data resulted in a nodal TIP noise of 2.851. Use of this T17 noise value as a compcoent of the process computer total uncertainty results in a one-signa process computer total oncertainty value for single-loop operation of 9.1% for reload cores.

B-1

.. , ..

ML20003C952

Text

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