ML112230966

Application to Amend 59 to App a Tech Specs for License DPR 49.Amend Covers Reactor Startup & Operation Longer than 24-h W/One Recirculation Loop Out of Svc.Class III License Change Fee Encl

ML112230966
Person / Time
Site:	Duane Arnold
Issue date:	05/02/1980
From:	Root L IES Utilities, (Formerly Iowa Electric Light & Power Co)
To:
Shared Package
ML112230967	List: ML112230966 ML19309G791
References
LDR-80-127, NUDOCS 8005070573
Download: ML112230966 (30)

Text

REGULATORY -INSMATION DISTRIBUTION SYSTL(RIDS)

ACCESSION NBR:8005070573 DOC,DATE: 80/05/02 NOTARIZED: YES DOCKET #

FACIL:50-331 Duane Arnold Energy Center, Iowa Electric Light & Pow 05000331 AUTHNAME AUTHOR AFFILIATION ROOTL,D. Iowe Electric Light & Power Co.

RECIP,NAME RECIPIENT AFFILIATION

SUBJECT:

Application to amend Tech Specs App a to License DPR*49.

Amend is for reactor startup & operation greater than 24-h w/one recirculation loop out of svcClass III license change fee encl DISTRIBUTION CODE: AQO1S COPIES RECEIVED:LTR, /ENCL . SIZE: it.

TITLE: General Distribution for after Issuance of Operating Lic NOTES:"1. & ... .. . . . .. . . . . .. . . .. . . . . . . . .

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL ACTION: 05 BC 0AS *3 .7 7 INTERNAL: 1 1 02 NRC PDR 1 1~

12 I&E 2 2 15 CORE PERF BR 1 1 17 ENGR BR 1 18 REAC SFTY BR .1 1 19 PLANT.SYS 8R 1 1 20 EEB 1 1 21 EFLT TRT SYS 1 EPB-DOR 1' 1 OELD 1 0 STS GROUP LEADR 1 1 EXTERNAL: 03 LPDR 1 1 04 NSIC 1 1 23 ACRS 16 16 AY 9 1980 TOTAL NUMBER OF COPIES REQUIRED: LTTR 38 ENCL 37

9 8 005070t5 7 Iowa Electric Light and Power Company May 2, 1980 LDR-80-127 LARRY D. ROOT ASSISTANT VICE PRESIDENT NUCLEAR GENERATION Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

Transmitted herewith in accordance with the requirements of 10 CFR 50.59 and 50.90 is an application for amendment to Appendix A (Technical Specifications) to operating license DPR-49 for the Duane Arnold Energy Center (DAEC). The pro posed amendment is for reactor startup and operation for greater than 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. Justification for this proposed change is also enclosed.

This application has been-reviewed by the DAEC Operations Committee and the DAEC Safety Committee.

It has been determined that this is a Category III amendment and a pay ment of $4000.00 is herewith enclosed.

Three signed and 37 additional copies of this application are transmitted herewith. This application consistin6g of the foregoing letter and enclosures hereto, is true and accurate to the best of my knowledge and belief.

IOWA ELECTRIC LIGHT AND POWER COMPANY Larry D. ot Assistant Vice President LDR/RFS/mz Nuclear Generation Attachment cc: R. Salmon Subscribed To And Sworn To Before Me On D. Arnold This P~wee&day of 19 L. Liu S. Tuthill D. Mineck J. Van Sickle T. Kevern (NRC)

File: A-117 Notary Public In and For The State of Iowa Poo'I

~1/

GeneralI Ofic

• PO. RoX 351

• Cedar Rapids, Iowa 52406

• 319/398-4411

8oeo0 Proposed Change RTS 119 to The Duane Arnold Energy Center Technical Specifications The holders of License DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by deleting current pages and replacing them with the attached new proposed pages. A list of the affected pages is attached with the proposed new pages.

List of Affected Pages 1.1-2 1.1-3 3.2-16 3.2-17 3.6-7

0 DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 16.C Power Transient Where: S = Setting in percent of rated power (1,593 MWt)

To ensure that the Safety Limits established.in Speci W = Recirculation loop flow fication 1.1.A and 1.1.B are in percent of rated flow.

not exceeded, each required Rated recirculation loop scram shall be initiated by flow is that recirculation its primary source signal. loop flow which corresponds A Safety Limit shall be to 49x10 8 lb/hr core flow.

assumed to be exceeded when scram is accomplished by a means other than the Phidry Source Signal.

D. With irradiated fuel in the reactor vessel, the water level shall not be less than 12 in.

above the top of the normal For a MFLPD greater than FRP, the active fuel zone. Top of the APRM scram setpoint shall be:

active fuel zone is defined to be 344.5 inches above S < (0.66W + 54) -'~D for two vessel zero (See Bases 3.2) recirculation loop operation and S < (0.66W + 50.7) FRP for one MFLPD recirculation loop operation.

NOTE: These settings assume operation within the basic thermal design criteria.

These criteria are LHGR . 18.5 KW/ft (7x7 array) or 13.4 KW/ft (8x8array) and MCPR > values as indicated in Table 3.12-2 times Kf, where Kf is de fined by Figure 3.12-1. Therefore, at full power, operation is not allowed with MFLPD greater than unity even if the scram setting is reduced. If it is determined that either of these design criteria is

-being violated during operation, action mut be taken immediately to return to

/operation within these criteria.

L2. APRM High Fl ux Scram When in the REFUEL or STARTUP and

.HOT STANDBY MODE. The APRM scram shall be set at less than or equal to 15 RercentQf rated-power.

Amendment No. 59 1.1-2

DAEC-l SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

3. APRM Rod Block When in Run Mode.

For operation with MFLPD less than or equal to FRP the APRM Control Rod Block setpoint shall be as shown on Fig. 2.1-1 and shall be:

S-. (0,66W + 42)]

The definitions used above for the APRM s'cram trip apply.

For a MFLPD greater than FRP, the APRM Control Rod Block set point shall be:

(0,66W + 42) FRP for anFL PD two recirculation loop operation, and S < (0.66W + 38.7) FRP for MFLPD one recirculation loop operation.

4, IRM - the IRM scram shall be set at less than or equal to 120/125 of full scale, B. Scram and Iso > 513.5 inches lation on reac above vessel tor low water zero (+12" on level level instru ments)

0. Scram - turbine < 10 percent stop valve valve closure closure D, Turbine control valve fast closure shall occur within 30 milliseconds of the start of turbine control valve fast closure, Amendment No, 59 1,1-3

TABLE 3.2-C Minimum No.

of Operable Instrument Number of Channels Per Instrument Channels Trip System Instrument Trip Level Setting Provided by Design Action 2 APRM Upscale (Flow Biased) for 2 recirc loop operation c(O.66W + 42) FRP C (2) 6 Inst. Channels (1)

MFLPD for 1 recirc loop operation

<(0.66W + 38.7) FRP (2)

MFLPD 2 APRM Upscale (Not in Run Mode) <12 indicated on scale 6 Inst. Channels (1) 2 APRM Downscale >5 indicated on scale 6 Inst. Channels (1) 1 (7) Rod Block Monitor for 2 recirc loop operation (Flow Biased) <(0.66W + 39) FRP 2 Inst. Channels (1)

MFLPD for 1 recirc loop operation

<(0.66W + 35.7) FRP (2)

MFLPD 1 (7) Rod Block Monitor >5 indicated on scale 2 Inst. Channels (1)

Downscale 2 IRM Downscale (3) >5/125 full scale 6 Inst. Channels (1) 2 IRM Detector not in (8) 6 Inst. Channels (1)

Startup" Position 2 IRM Upscale <108/125 6 Inst. Channels (1) 2 (5) SRM Detector not in (4) 4 Inst. Channels (1)

Startup Position 2 (5) (6) SRM Upscale <105 counts/sec. 4 Inst. Channels (1)

DAEC-l NOTES FOR TABLE 3,2-C

1. For the startup and rundpositions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.

The- SRM and IRM blocks need not be operable in "Run" mode, and the APRM [except for APRM Upscale (Not in Run Mode)] and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven .days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped, If the. first column cannot be met for both trip systems, the systems shall be tripped.

2. W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (1593 MWt). A ratio of FRP/MFLPD (1.0 is permitted at reduced power, 3, IRM downscale is bypassed when it is on its lowest range,

4. This function is bypassed when the count rate is >100 cps.

3.2-17

LIMITING CONDITIONS PG OOPERATION SURVEILLANCE REdCOREMENTS LIMITING CONDITIONS F~OPERAT(ON SURVEILLANCE RE*REMENTS

b. The indicated value a@f core flow rate varies fromrthe value derived from loop flow measurements.by uaore than 10%.

c. The diffuser to lower plenum differential pressure read ing on an individual jet pump varies from the mean of all jet pump differeatial pressures by more than 10%.

2. Whenever there is recircu lation flow with the reactor in the Startup or Rum -mode, and one recirculation- pump is operating, the diffuser to lower plenum differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pres sures in that loop by more than 10%.

F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch

1. When both recirculation 1. Recirculation pump' speeds pumps are in steady state shall be checked and :Uogged operation, the speed of at least once per day.

the faster pump may not exceed 122% of the speed of the slower pump when core power is 80% or more of rated power or 135% of the speed of the slower pump when core power is below 80% of rated power.

2. If specification 3.6.F.1 cannot

-be.met, one-recirculation-pump shall be tripped. The reactor may.Jbe startedad oneratetcwith one recirculation loop out of ser viceorovidedthat:

a. A MAPLHGR multiplier of I 0.65 is applied.

b. The power level is limited to a maximum of 50% of licensed power.

3.6-7

JUSTIFICATION FOR THE OPERATION OF DUANE ARNOLD ENERGY CENTER WITH ONE RECIRCULATION LOOP OUT OF SERVICE 1.0 Introduction The Technical Specifications for the Duane Arnold Energy Center (DAEC) (3.6.F.2) require that the plant be shutdown if an idle recirculation loop cannot be returned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At approximately 5:00 AM, April 30, 1980, the "A"Loop Recirculation Pump Motor Generator (MG) Set tripped. Subsequent investigations indicate that a short to ground in the motor end of the MG Set exists requiring rewind of the motor prior to returning the "A" Loop to service. Present estimates indicate that a minimum of approximately three weeks are necessary to affect repairs.

The DAEC was returned to service after the spring refueling outage on April 18, 1980 which started February 9, 1980. The DAEC carries approximately 70% of Iowa Electric and partners load at this time of year. At 50% power capability, the DAEC would carry about 35%

of the load. The effect of the DAEC not being operational is a severe impact to the rate payers.

In order to resume operation with one loop out of service, Iowa Electric has contracted GE to provide an analysis as outlined.

-'1.-

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, Iowa Electric commits to the following conditions during normal operation.

1. Recirculation loop A recirculation pump is electrically disarmed and the motor is inoperable precluding operation of the pump or injection of a cold slug into the vessel.

2. The recirculation controls will be placed in the manual mode, thereby eliminating the need for control system analyses.

3. The settings-fTor the rod block monitor, APRM rod block trip, and flow bias scram will be modified as necessary to provide for single loop operation. Technical Specifications are enclosed hereto.

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

5. MAPLHGR restrictions, Figure 3.12-2 through 7,will result in a 35 percent reduction for all fuel.

6. The limitation on power level as described in FSAR Section 14.5.6.2 is 55 percent. Iowa Electric will further limit the power level to 50%.

3.0 MAPLHGR Adjustment Factor for DAEC General Electric is performing analyses for single loop operation of DAEC. Preliminary evaluation of these calculations performed according to the procedure outlined in NEDO-20566-2 indicates that a multiplier of 0.86 should be applied to the MAPLHGR limits for single loop operation of the DAEC, Further, GE has performddea large number of single loop analyses for similar plants; in no case has a multiplier of less than 0.70 been required. Because DAEC does not have LPCI modification and because the limiting break is a suction line break, the single loop MAPLHGR multiplier is expected to be significantly better than for most other BWR's. Until the DAEC 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 at the DAEC.

4.0 Other Considerations for Single-Loop Operation Various conditions have been examined for the impact of single-loop 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. Kf Curves C. Stability Analysis D. Thermal -- Hydraulics 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, but hore rapid and severe influence, Following a pump seizure event, natural circulation flow continues, water level is maintained, 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 resulting 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 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 perforation 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, ABNORMAL OPERATIONAL TRANSIENTS TRANSIENTS AND.CORE DYNAMICS Since operation with one recirculation loop results in a maximum power output which is 20 to 30%'below that from which can be attained for two-pump operation, the consequences of abnormal operational transients from one-loop operation will be considerably less severe than thospanalyzed from a two-loop operational mode.

For pressurization, flow decrease, and cold water increase, transients previously transmitted Ifor Reload/FSAR results bound both the thermal and 6verpfessure -conse quences 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 operation are considerably less because of the associated reduction in operating power level.

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

Cold water increase transients can result from either recircultation pump speed up or introduction of colder water into the reactor vessel by events such as loss of feedwater heater. For the former, the Kf factors 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 fMCPR 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, and therefore, theKf factors derived with the two pump assumption are con servative for single-loop operation. For the latter, the Toss of feedwater heater event is generally the most severe cold water increase event with re spect to increase in core power. This event is caused by 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 significantly lower, the one-pump cold water increase case is conservatively bounded by the full power (two-pump) analysis.

From the above discussions, it can be concluded that the transient consequence from one-loop operation is bounded by previously submitted full power analysis.

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

ROD WITHDRAWAL ERROR The rod withdrawalP 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 limit. The MCPR requirement for one-pump operation will be equal to that for two-pump operation because the nuclear characteristics are indepen dent of whether the core flow is attained by one- or two-pump operation. The only exceptions to this independence are possible flow asymmetries which might result from one-pump operation. Flow asymmetries are shown to be of no concern by tests conducted at Quad Cities. Uhdet'ddnditions of one-pump operation and equalizer valve closed, flow was found to be uniform in each bundle. The DAEC does not have an equalizer line.

One-pump operationrresults in backflow through 8 of the 16 jet pumps while the flow is being supplied into the lower plenum from the 8 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 modifying 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) where 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 10' RB100 m(100) + K or K = RB10 0 - m(100)

Substituting for K in Equation 1, the two pump equation becomes; RB = mW + [RB 100 - m(100)] (2)

b. Next, the core flow (Fc) 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 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 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.

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, [2, is determined. Thus,AW 1 - W2*

The rod block equation corrected for one pump flow is:

RB = mW + LRB 10 0 - m(100) - ARB where ARB = RB - RB mAW RB = mW + RB100 - m(100 +AW) (3)

d. For DAEC application, the constants from the Technical Specification are:

m = 0.66 RB10 0 = 108 From Figure 2:

W=W 1 - 2 35 -30= 5 Evaluating in Equation 3, the one-pump rod block equation becomes:

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

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

APRM TRIP 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 discussed above.

Kf CURVE 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 operational events.

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 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 illustrated in Figure 3.

THERMAL-HYDRAULICS 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. The only exception is the core total flow. The standard deviation for this quantity from Table 4-2 is 2.5%. For single-loop operation this value may increase to about 6% of rated core flow.* The 3.5% increase in core total, flow uncertainty corresponds to an increase in the safety limit of about 0.004%which can be neglected.

The steady-state operating MCPR with single-loop operationewill be conser vatdively established by multiplying the Kf factor to the rated flow MCPR limit. This ensures that the 99.9% statistical limit requirement is always satisfied.

• See Appendix A for justification, SHORT TERM LESSONS LEARNED Iowa Electricowas requested by the NRC to evaluate single-loop operationF in light of the Short-Term Lessons Learned from TMI. Iowa Electric and General Electric have condicted a review of the Sh6rt-Term Lessons Learned and have determined that there is no unique consideration which would affect any actions taken in accordance with the Short-Term Lessons Learned.

Iowa Electric will conduct training with all Control Room operators prior to those personnel operating the plant in single loop. This training will consist of review of appropriate Technical Specifications, core limits, precautions, surveillance tests, and bases for single loop operation. As the plant is normally operated in local manual recirculation control at less than 50% power the operations are not unique and specific training is not required.

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APPENDIX A UNCERTAINTIES IN TOTAL CORE FLOW FOR SINGLE LOOP OPERATION

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 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 backflow ing jet pumps must be subtracted from the measured flow in the active loop.

In addition, the jet pump flow coefficient is different in reverse flow than forward flow, and the measurement of reverse flow must be modified to account for this fact.

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

Total Core Active Loop' - C Inactive Loop Flow Indicated Flow) (Indicated Flow where C = 0.95 and "Loop Indicated Flow" means the flow indicated by the jet pump "single-tap" loop flow summers and indicators, which are set up to indi cate 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 AP versus core 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 core support plate AP, along with the loop flow indicator readings.

• Note: The expected value of the "C" coefficient is-0.89.

I.

2. CORE FLOW UNCERTAINTY ANALYSIS The uncertainty analysis procedure used to establish the core flow uncertainty for one pump operation is basically the same as for two pump operation, except for some extensions. The core flow uncertainty analysis is described in Refer ences 1 and 2. The analysis of one pump core flow uncertainty can be summarized as follows:

a. During one pump operation, the core flow is measured by the following formula:

Total Inactive Loop Core = Active Loop C Flow (Indicated Flow, (Indicated Flow) 10 P ST K / 20 S

=K K s K forward reverse forward i=l i=11 where APST is the "single tap" differential pressure for jet pump "i".

The constant "C" is required to modify the inactive loop flow indication since the jet pump diffuser flow coefficient is different for reverse flow compared with thefforward flow coefficient used for the core flow instru mentation calibration.

b. The core flow uncertainty analysis must now account for the uncertainty in "C". The value of "C" has been determined analytically, using a conserva tive bounding analysis: therefore, the core flow input to the process com puter during one pump operation has a conservative bias, since "C" was analyzed in a conservative manner. However, the following uncertainty analysis is based on the uncertainty in the true (or nominal) value of "C;" not the uncertainty in the conservative value of "C" used in the reactor flow-measurement.

A-2

0s "C" can be defined as:

where:

Kforward =

The forward flow loss coefficient resulting from in-reactor calibration tests assumed for the analytical derivation of "C."

2 Note: K (flow) ( aP)

Kreverse = The loss coefficient calculated for reverse flow.

Combining the uncertainties in Kforward and Kreverselit can be shown that d 1 2

+ 2 c FdKforward K ]

Bounding values are < 2.5% and 6.4%, thus:

dK dK forward reverse dc = 3.4%

c. Now the effect of this reverse flow coefficient uncertainty must be related to total core flow uncertainty. Assuming that 33%* of the flow in the active (forward flowing) jet pumps backflows throughtthe inactive pumps, it can be shown that:

dW 2 2 / 0.33 2 2 C5

= 6A += 1-0.33 ) C

• Nbte: This value can vary from about 20% to 30%, depending on plant type and operating conditions. 33% is a conservative bounding value.

A-3

J . -

0 where:

- the uncertainty in the total core flow.

WT WA = the uncertainty in the active loop flow.

The flow uncertainty in the active jet pumps during single loop operation (dw )

is presently analyzed to be <3.5%. To produce a conservative, bounding A analysis, 6W = 4.0% was used in this calculation. Then, A

2 2 2 = (4.34%)2

-= (4.0%) +{0.33 } (3/4%)2

.WT When the effect of 4.1% core bypass flow uncertainty at 12% (bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncertainty is:

2 2

= (4.34%)2 + 0.12 = (4.38%)2 active (4.1%)2 1-0.12) cool ant This verifies the assumption of core flow uncertainties of 6%. Actually, the core flow accuracy is expected to be much better, as shown above.

In summary, core flow during one pump operation is measured in a conservative way, its uncertainty has been conservatively evaluated, and thernet effect on MCPR is insignificant.

REFERENCES - Appendix A

1. Letter to Walter R. Butler (AEC/NRC).

Subject:

Response to the Third Set of AEC Questions on the General Electric Licensing Topical Reports NEDO-10958 and NEDE-10958, "General Electric BWR Thermal Analysis (GETAB):

Data, Correlation and Design Application," July 11, 1974.

2. J.F. Carew, "Process Computer Performance Evaluation Accuracy," June 1974 (NEDO-20340).

A-4

REFERENCES:

1. NEJO-20566-2, Revision 1, GE Analytical Model for LOCA Analysis in Accordance with 10 CFR 50 Appendix K Amendment No. 2 - One Recirculation Loop Out-of-Service

2. "GE/BWR Generic Reload Licensing Applications for 8x8 Fuel," Rev. 1, Supplement 4, (NEDO-20360)