ML20116M360

Proposed TS Table 3.2-1, DNB Parameters Reducing RCS Total Flow Rate to 375,000 Gpm

ML20116M360
Person / Time
Site:	Sequoyah
Issue date:	11/16/1992
From:	TENNESSEE VALLEY AUTHORITY
To:
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ML20116M358	List: ML20116M357 ML20116M360
References
NUDOCS 9211200210
Download: ML20116M360 (29)

Text

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i ENCLOSURE 1

• PROPOSLD TECHNICAL S?ECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328- ,

(IVA-SQN-TS-92-13 )

i 6

LIST OF AFFECTED PAGES 11 nit.1 l 3/4 2-16 IJniL_2 3/4 2-14 s

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DNB PARAMETERS I

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g, R142 vo ifAY 091990 SEQUOYAH - UNIT 1 3/4 2 ;

Amencment No, 41.133 s

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8 DNB PARAMETERS sx e

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it ENCLOSURE 2 )

PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE '

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SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 .

DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-92-13)

DESLRIPTION AND JUSTIFICATION FOR l

REACTOR C001 ANT SYSTEM MINIMUM FLOW RATE REQUIRFAENT REDUCTION 9

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, . Descriptir LaLChange TVA proposes to soodify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to reduce the required reactor coolunt system (RCS) total flow rate from greater than or equal to 378,400 gallons per minute (gpm) to greater than or equal to 375,000 gpm in TS 3.2.5. Tabic 3.?-1. The footnote associated with this flow rate requirement in TS Table 3.2-1 !s modified to reflect a 2.4 percent flow measurement uncertainty instead of the generic 3.5 percent utilized for initial licensing and incorporatss a typographical correction. The typographical error involves the "#" footnote for the Unit 2 TS Table 3.2-1 where the word " uncertainty" was misspelled. An additional clarification has been incorporated to provide consistency between the TSs for both units by using "Renetor Coolant System Total Flow Rate" as the title for this parameter in TS Table 3.2-1.

Reason _for_ Change The margin between the currently calculated RCS total flow rates and the present TS limiting value is less than one percent for both SQN Units 1 end 2. Changes in the methodt for calculating RCS flow rates in combination with the flow reductions associated with steam generator tube plugging and reactor coolant pump impeller wear have affected the calculated margin at SQN over time. With ;he potential for additional RCS flow reduction from future tube plugging, punp wear, fuel assembly design changes, or other factors, the less then one percent margin to the niinimwn TS-required RCS flow rate could result in conditions where the existing TS r(quirement could not be met. This would result in the shutdown of the unit or prevent unit start-up and require subsequent emergency TS changes or TS waivers of compliance to return to power operation. For these reasons, the reduction in the required RCS flow rate provided by the proposed request and justitied by the measurement uncertainty value wil) minimize the potential impact to power operation.

Justification _10t_ Change The justification for this change is based upon the Westinghouse Electric Corporation Safety Evaluation Cbecklist (SEch)92-288 and Letter TVA-91-349 provided in Enclosure 4 of this submittal. These evaluations determined the RCS flow measurement uncertainty based upon the use of RCS elbow tap dif ferential pressures normalized to baseline calorimetric flow rates obtained at the beginning of Cycle 1 for both SQN units. The result of these nyaluations is that a 2.4 percent measurement uncertainty is applicab'.e to the RCS flow rate instrumentation in contrast to the 3.5 percent value presently used in the SQN TSs. This 3.5 percent value was used for initial licensing of SQN based on generic uncertainty values for RCS flow measurement instrumsatation. The most recent evaluation of the measurement uncertainty for SQN is documented in Westinghouse Litter TVA-91-349 and coafirms the excessive conservatism in the 3.5 percent value. The reduction in the measurement uncertainty value is based solely on the Weatinghouse calculation that provides the basis for superseding the generic 3.5 percent value. No changes in plant equipment have been involved in this reduction.

,o I For this TS change, the proposed RCS flow rate is greater than or equal to 375,000 spm, which represents the RCS design flow rate of 365,600 gpm (91.400 gpm in each loop) plus 2.4 percent for measurement uncertainty and rouaded to the next highest thousand gpm. This rounding effort provides approximately 600 gpm additional margin in the proposed RCS total flow rate vaine. Therefore, this reduction in the flow measurement (meertainty and the related decrease in the RCS total flow rate minimum requirement do not alter an; of the accident analysis assumptions because only the excess margin for RCS flow instrumentation uncertainty is removed. The design flow rate assumed in the accident analysis remains unchanged by this proposal.

)

Additional discussions are included in SECL 92-288 for continued power operation with lower RCS flows at reduced reactor power. The proposed change does not incorporate those provisions at this timet however TVA plans to pursue this option along with upcoming TS changes for new flow mixing fuct assembly designs.

Enironmental_ImpacLEvaluation The proposed change request doec not involve an unreviewed environmental question because operation of SQN Unita 1 and 2 in accordance wi th this change would nott

1. Result in a significant increase in any adverse environmental impact previously ovaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety end Licensing Board, supplements to the FES, environmental impact appraisulo, or decisions of the Atomic Safety and Licenning Board.

2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact, a

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Enclosure 3 s.

PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLFAR PIANT UNITS 1 AND 2 L~ DOCKET NOS. 50-327 AND 50-328 (IVA-SQN-TS-92-13)

DETERMINATION OF Nu SIGNIFICANT !!AZARDS CONSIDERATION g

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Significant llazards Evaluation TVA has_ evaluated the proposed technical specification (TS) change and I has determintd that l' does not represent a significant hazards j consideration based on criteria established in 10 CFR 50.92(c).  ;

Ope rat ice.) of Sequoyah Nuclear Plant (SQN) in accordance with the proposed I

, amendent will nots ~

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1. Involve a significant increase in the proba*>ility or consequences of  ;

an accident previously evaluated. 1 The proposed changes do not alter any of the assumptions used in the accident analysis. The reduction in the minimum TS reactor coolant l system (RCS) flow rate only eliminates excess measurement uncertainty. Therefore, no change in any accident analysis assumptions or any plant configuration is involved in the proposed TS change. Base:1 on this, no increase in the probability or '

consequences of an accident can reruit from this TS change because RCS design flow rates remain unchanged, (nsuring no change in the plant response for normal or accident conditions.

2. Create the possibility of a new or different kind of accident from any previously analyzed. -

No plant design paramoters, equipment, or operating conditions are altered by the proposed TS change and therefore, no possibility of a new or different kind of accident la created. This elimination of the excess RCS flow measurement uncertainty margin will permit plant operac. ion below the existing RCS flow rate requirement, but not below the measured value that ensures operation at greater than or equal to the design flow asswoed in the accident analysis.

3. Involve a significant reduction in a margin of safoty.

The proposed TS change only affects the excess RCS flow measurement uncertainty above the required design flow. This reduction does not affect the " margin of safety" because the RCS design flow remains unchanged and thereby maintains the margin between design operating conditions and the RCS flow rate required to maintain the accident analysis assumptions. Therefore, the proposed. reduction in minimum 1108 flow te account for thu excess measurement uncertainty does not invoh e a reduction in a margin of safety, a

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3 Enclosure 4 I i

PA0 POSED TECHNICAL SPECIFICATION (TS) CllANCE ,

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SEQUOYAll NUCLEAR PIANT UNITS 1 AND 2 i

DOCKET NOS. 50-327 AND 50-328 l

('IVA-SQN-TS-92-13)

WESTINGl!OUSE El.ECTRIC CORPORATION EVALUATIONS l-FOR REDUCED REACTOR C001 ANT SYSTEM MINIMUM FLOW RATE >

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OCT 0 *92 15828 FROM t.lCEtGli0 TO TVA EEQUOYAH Pi4E.002/0G9' t

SECL 92-288 Customer Reference No(s).

N/A Westinghouse Paference No(s).

N/A WESTING 3OUSE NUCLEAR SAFETY SAFETY EVALUATION CIIECK LIST (SECL) 1.) NUCLEAR PLANT ($): Scouovah Units 1 & 2 _

2.) SUBJECT (TITLE): RCS Flow hitastemtDLilnsertainty Faction From 3.5 % to 2.4 %

3.) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59(b) has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.

Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.

CHECK LIST - PART A - 10CFR50.59(a)(1) 3.1) Yes __ No .X_ A change to the plant as described in the FSAR7 3.2) Yes _ No .X. A change to procedures as described in the FSAR?

3.3) Yes _ No 1 A test or experhneat not describe:I in the FSAR7 3.4) Yes .X. No _ A change to the plant technical specifications?

(See Note on Page 2.)

4.) CHECK LIST - PART B - 10CFR50.59(a)(2) (Justificatiou for Part B answers must be hicluded ou page 2.)

4.1) Yes __. No 1 Will the probability of an accidem previously evaluatal in the FSAR be increased?

4.2) Yes _ No .X_ Will the consequences of an accident previously evaluated in the FSAR be increased?

4.3) Yec __. No 1 hfay the possibility of an accident which is different than any already evaluated in the FSAR be created?

4.4) Yes _._ No X_ Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

4.5) Yes ._ No _X, Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

4.6) Yes _ No K. May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

4.7) Yes _ No _X_ Will the margin of safety as described in the bases to any technical specification be %Ked?

. 007 6 '93 15829 FRCr1 LICD61tG TO TWI SEQJOYfiH @ FEE.00$ 4 W

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SECL-93-288 NOTES:

If the answer to any of the above questions is unknown, indicate under 5.) REMARKS and explain below.

If the answer to any of the above questions in Part A (3.4) or Part D canuot in answered lu the negative, bued on written safety evaluation, the change review would require an applision for lkense arnendment as required by 10CTR50 59(c) and submitted to the NRC pursuant to 10CFR50.90.

5.) RCMARKS:

The answers given in Section 3, Part A, and Section 4, Part D, of the Safety Evaluation Checklist, are based on the attached Safety Evaluation. .

FOR FSAR UPDATE Section: _ Pages: _ Tables: Pigures: . .

Technical Specifice.lon Table 3.2-1, DNB Parameters (marked up Units 1 & 2 copies auached)

SAFETY EVALUATION APPROVAL LADDER:

Nuclear Safety Preparer:

M_. UdC 6 b L V. Tomasic 6o4 7 + 7 2.

Date: .,

Coordinated with Engirrers: CJL Tuley _ _ _

Date: _--

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Nuclear Safety Group Managec 'i. -'M.,Livine

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cCT 0 '93 13830 IF@FLTGRW Wr%-rvacate W M.vtw w SECl & 388 1.0 HACKGROUND Westinghouse was requested by TVA to modify the Sequoyah Unit 1 and 2 Technical Specification Table 3.21, DNB Parameters, to reflect a reduced RCS Flow measurement uncertainty from 3.5 %

to 2.4 %, as identified in previous documents addresslog Dueline Calorimetric Flow Uncertainty Combined With Elbow Tap Nonnalization U; certainty (Reference 1). His safety evaluation addresses the uncertainty reduction.

[lYA also requestcd that Westinghouse address a tradesif between RCS Flow and power to allow continued operation when less than the Technical Specification value would b:. indicated in the plant.

niis was revleted and determined to be feasible requiring additional Sequoyah plant specific effort involving tue evaluation of the DND parameters for the FSAR Chapter 15 Non LOCA Accidents.

The findings of this review are suinmarized for itJornation as follows:

Dased on earlier work performed for hicGuire ami V. C. Summer (which were subsequently ar> proved by the NRC) htinghouse has determined suitable modifications to allow continued opersion with flow reduced by up to 5 % from nermal Design Flow (as low as 356,000 gpm) at reduced power levels (a co'rrespotuling maximum limit on pcwer of 90 % RTP). He power to flow relationship used for this wotk is a 2.0 % power reduction for each 1.0 % RCS Flow is below Hermal Design Flow, both parameters in Integer units. His is a very conservative relationship since the recognized relarlopsbip is typically en ~0.5 to OA % RTP power reduction for each 1.0 % RCS Flow redaction to maintain a balance in DND space. h should be noted that Hennal Design Flow (TDF) is the system flow used as an initial condition in the various sdecy analyses (91,400 gpm /

loop x 4 loops - 365,600 gpm) without measurement uncertaloties and 375,000 gpm with measurement uncertainties. All values of flow quoted in this suminary are based on TDP - 375,000 gpta, I. e., with a measurement uncertainhy of 2.4 %.

De typleal relationship assur.2ed for RCS Flow to DNB 41.0/1.0. Dus a 1 % reduction in RC9 Flow would result in a loss of approxime.ely 1 % in DND margin. The typleal relationship assumed for Reactor Power to DND is 1.0/2.0, i.e., an increase in reactor power of 0.6 % would result in a loss of apprstimately I % in DND margin. Rese are typleal sensitivity values that must be verified on a plant specific basis end are a function of the DND x elation ut" red. As noted previously, similar changes to plant techn.lcal specificulons have been made on other plants. He licensing precedent set in these instances (McGuire and V. C. Summer) is the use of a Reactor Power to DND sensitivity of 2.0/1.0, i.e., a factor of four in the opposite direedon than the correlations demonstrate.

This is a very conservative assumption and results in a 2 % reduction in reactor power for each 1 %

reduction in RCS Flow from TDF.

With this conservative relationship established between RCS Flow and power, it should be apparent that the Core Limits (Safety Limhs in the Technical Specificatiors, Figure 2.1 1) should remain unchanged if the power level of the plant is interpreted somewhat differently with the reduced flow, i.e., Figure 2.1-1 is based on the Fraction of Rated nermal Power equal to 1.0 being 3411 hnVth.

When the power level is reduced due to low measuted RCS Flow, the Fraction of Rated Thermal Power shoald reflect the reduced power bebg considered 1.0, e.g., if RCS Flow is deterinined to be 98 % TDF (367,000 gpm), then reactor power should be reduced to 96 % of 3411 MWth, or 3275

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% $ECL-92 288 e MWth. Figme 2.1 1 would then be based on the new power level belag the equivalent of full power,  ;

i.e.,1.0 Fraction of Rated Thermal Power would be 3275 MWth, not 3411 MWth. With this rsvised i

interpretation of what power level corresponds to Rated Hermal Power, the Core Limits should remain unchanged.  !

l With the Core Limits unchanged, Overtemperature AT (OTDT) and Overpower AT (OPDT) reactor trips should be unaffected. This removes ag requirements to modify the constants K,, K, and K, for OTDT and K., K, and K, for OpDT. Ilowever, it is still necessary to scale OPDT to reflect the loop specific, todicated AT and T* values and to scale OTDT to reflect the loop specine, Indicated AT and T' values at the reduced power operating conditions.

Finally, the lower limit of 95 % TDP (356,000 gpm) was selected for several reasons.

1) The RCS Flow Low reactor trip se(point remains unchanged (90 % TDF) and it is believed r that further reduc 00ns la flow, with consideration of instmment uncertainties, could result in spurious actuation of this protection function.

2) The DND to RCS Flow and DNB to Reactor Power sensitivities are valid and linear for only relativeiy small changes in the parameters. Decreases in RCS Flow of greater than 5 %would result in changes to Tavg and AT which would fall outside the typical bounds of a full power sensitivity calculation. Thus new sensivity calculations would be requiredi-

3) A comparison of t'te loss le DNB margin due to the reduced RCS Flow vs the DNB margin available in the NIS Power Range reactor trip due to actuation prior to the Safety Analysis Limit ,

cancel at an approximate 5 % redue lon in RCS Flowi Further reductions la RCS Flow as a  !

steady state operating condition would require a corresponding rede: tion in the NIS Power Range .

Nominal Trip Setpoint to reflect the effect!"e change h allowed full power (Fraction of Rated Thermal Power equal to 1.0). ]

2.0 IJCENSING BASIS ,

ne werk performed is consistent with the requirements of 10CFR50.36 and information documented ,

in WCAP-11239, Revisions _1 - 6 for setpoint/ uncertainty calculations previously perforned by Westinghouse. As noted above, the uncertainty calculations are essentially the same as those v performed previously for Sequoyah Units 1 & 2. Dithrences frc~ previous calculations lie in the assumption of the normaliza !on of the Cold Leg Elbow Taps to a single, previously performed RCS Flow Calorimetric measurement (Cycle 1) which requires the inclusion of additional uncertainties in the determination of the indicated RCS Flow uncertainty.

3.0 EVALUATION ,

. In late 1991, et the request of TVA, We tinghouse performed calculations to determine the effect of a single normalization of the Cold leg Elbow Taps to a reference RCS Flow Calorunetric (Reference  ;

1). Based on the entrent plant configuration; RTD Bypass Elimination, Eagle-21 protection system 1 process racks, the performance of a single normalizarlou of the Cold Leg Elbow Taps to the RCS Flow Calorimetric performed BOL Cycle 1 and indication of RCS Flow via the plant process .

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,. SECL-92 288 cornputer, the calculated instament uncertainty is 2.4 % Flow. This is a redoction of - 1.1 % Flow from the current NRC mandated value of 3.5 % Flow. Westinghou.e he determmed that the reduced instrument uncertainty calculation is reasonable and bask <lly consistent with the Westinghouse approach approved by the NRC. 'Ibe only significant difference is the assumption of norma'imion to a single, previously performed RCS Flow calorimetric. Howner, this has been accounted for by 6e addition of instrument uncenainties usually considered to be zeroed out by the norma:laation performed each cycle. Based on continued Safety Analysis use of Thermal Design Flow (91,400 gym per loop) the minimum RCS Flow that must be measured in the plant changes from 378,400 gpm to 375,000 gpm (rounded to the nearest thousand gpm).

4.0 DETERMINATION OF UNREVIEWED S & *ETY QUESTION While modifications to the plant technical specifications h:ve been determined to be necessary, cc unreviewed safety questions have becn Identified. The seven questions typleally answered for a 10CFR50.59 evaluation are noted as follows.

4.1 Will the probability of an accident previously evaluated in the SAR be lucreased?

With the reduced uncertainty, no increase in the probability of an accident has been octed. No changes to reactor trip setpoints are required, no changes to control system setpoiots or gains are required, no changes to installed equipment or hardware in the plant are required, thus the probability of an accident occurring remains unchanged.

4.2 Will the consequences of an accident previously evaluated in the SAR % lacreased?

With the reduced uncertainty, the initial conditions for all accident scenarios modeled rsmala unchanged. Therefore, the consequences of an accident will be the same as those previously ana!yzed.

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4.3 May the possibility of an accident which is different than any already evaluated in the SAR be created?

With the reduced uncertainty, no new accident scenarios have been identified. Operation of the plant will be consistent with that previously modeled, i.e., reactor trip setpoints and control function setpoints are the same, thus plant

• ponse will be the same and will not lutroduce any diffe.ent accident scenarios that have not been evaluated.

l 4.4 Will the probability of a malfunction of equipment important to safety previously waluat d in the SAR be lucreased?

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No changes to equipment installed in the plant are required. Protection function trip setpoints and control function setpointa remain unchanged and do not introduce additional constraints on equipment important to safety, thus there is no increase in the probability of a malfbaction of this; equipment.

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4.5 Wul the consequences of a malfunction of equipa. cat important to safety previously evaluated !n the SAR be increued?

With the reduced uncertainty the Ldtlal conditions present at the irdtlation of an acc! dent wul wul tot change. Therefore it is expected that the consequences of a malfunction of equipment important to asfety will not change.

4.6 May the possibility of a malfunctbu of equipment important to sMety different than any already

. evaluated in the SAR be created?

No changes to equipment installed in the plant are required. Protectiot, function trip setpoints and control function setpaints remain unchanged and do not intstdue:c additional constraints on equiprnent Itaportant to safety, thus no fa'h:re mode not previously evaluated is inaoduced.

4.7 Wul the margin of safety as defined in the BASES to any technical specifications be reduced?

With the changes to the technical specifications required as noted, the margin of safety as defined in the BASES will remain tht same.

5.0 CONCLUSION

S Based on the above it has been determined that the changes noted on the attached for specification Table 3.21, DNB Parameters, are acceptable for use at Sequoyah Units 1 and 2.

6.0 REFERENCES

1. TVA 91349 (L'I' NSL-OPL 1-91-628), dated November 6,1991, entided Baseline Calorimetric Flow Uncertainty Combined With / Elbow Tap Nonnalization Uncertainty.

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Reactor Coolant $ystem lv Total Flow i fM845 gpa# . '

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1. N8561 Mr. P. G. Trudel (B25 410f07 o16) TVA 91-349 Project Engineer ET-NSL-0PL-1-91-628 Tennessee Valley Authority November, 6,.1991 P. O. Box 2000 Ref: N-021 Soddy Daisy, TN 37379 Tennessee Valley Authority Sequoyah Nuclear Plants Units 1 and 2 Baseline Calorimetric Flow Uncertainty Combined with Elbow Tan Normalization Uncertainty

Dear Mr. Trudel:

This is in response to the request of Reference 1. .

Attachment A provides the elbow tap measurement repeatability justification 'for

- RCS flow verification.

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Attachment B provides the results of the uncertainty calculations for the RCS flow measurement employing elbow tap delta - pressure measurements normalized-to the baseline calorimetric flow performed at 100% power at the beginning of Cycle 1. The results of the instrument uncertainty for the loss of Flow Reactor trip and revisions to the Setpoint Study (WCAP ll239) are also included (as page mark ups). WCAP-ll239 will be revised and forwarded under separate cover.

If you havo any J0C DUFT cDM ~M o Nm please do not hesitate to contact us.

g "[ . *, my .co0Y ccPY DATE Very truly yours, lre gwuam 1 I 3nw ,j & Y

• f EmYN [ 1 ,-] O LB, J. Garry, Manager

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Attachment To TVA-91-349 l Attachment A '

i Page I cf 3 ELB0W TAP FLOW MEASUREMENT REPEATABILITY The elbow tap dp measurements on the RCS pump suction piping are being i used with increasing frequency, to determine if, and by how much RCS flows might have changed from one fuel cycle to the next'. Elbow taps are not considered to be accurate enough to define absolute flows, but the dp ,

measurements have been found to be repeatable and to provide accurate flow change indications. ,

Elbow flow meters (Ref 1) are a form of centrifugal meter, using the momentum forces developed by the change in flow direction. The principal narameters that determine the dp for a specified flow are the radius of-curvature of the elbow and the diameter of the flow channel through the elbow. Experiments on elbow meters have determined that the flow measurements are not affected by differences in surface roughness and have a high degree of repeatability.

Specific phenomena that have affected other types of flow meters, ur that might affect the elbow meters in the RCS application have been evaluated to determine if any of these phenomena would affect elbow meter repeatability. In addition, data from an cperating plant equipped with a -

highly accurate flow meter has been compared with the elbow meter .

measurements at that' plant to demonstrate elbow meter repeatability. The E"

~

results of these evaluations and comparisons are provided in the following sections.

1. Venturi Fouling Venturi meters are affected by crud deposits, called fouling, that affect surface roughness and throat area. The fouling is.apparently caused by an electrochemical ionization plating of copper and magnetite particles in the feedwater, a process associated with the large velocity increase as the flow approaches the venturi throat.

This condition is not present in an RCS elbow; there is no large 4 change in cross section to produce a velocity increase and ionization, and changes in surface roughness do not affect the elbow flow measurement.

2. Meter Dimensional Changes The elbow meter is primarily part of the RCS pressure boundary, so there would be only minimal dimensional changes associated with pipe stresses, and pressure and temperature would-be.the same (full power conditions) whenever measurements are made. Erosion of the elbow surface is unlikely since stainless steel is used, and the velocities are not large-(42 fps) relative to erosion. The effects.of any dimensional change or of erosion could only affect flow by changing elbow radius or pipe diameter, both large relative to any possible dimensional change. Therefore, the elbow meter is considered to be a highly stable flow measurement element.

t

. l e

s Page 2 Of 3 1

3. Upstream Velocity Distribution Effects I The velocity distribution entering  :

e elbow meter will be skewed by the upstream 40' elbow on the steam generator outlet nozzle, and the velocity distribution entering the steam generator outlet nozzle may be skewed due to its off-center location relative to the tube sheet.

These velocity distributions, including the distribution in the elbow meter, will remain constant through a fuel cycle, so the elbow meter dp would not change.

Steam generator tube plugging is usually randomly distributed across the tube sheet, so the velocity distribution approaching the outlet nozzle would not change. The velocity distribution could change if extensive tube plugging occurred in one locatir,n on the tube sheet, but the change would not be transmitted through the outlet nozzle to the elbow meter. The velocity approacM v) the outlet nozzle is small (6 fps) compared to the pipe velocity, 30 tr.is largt change in flow area wouid significantly decrease or flatten any upstream velocity gradient. Therefore, any tube plugging, even if asymmetrically distributed, would not affect the elbow flow measurement repeatability. .

4. Flow Measurement Comparisons k The Leading Edge Flow Meters (LEFM) installed in both reactor coolant loops at Prairie Island Unit 2 provide a mt.ans to confirm repeatability of the elbow flow meters. The comparisons covered 11 years of operation, during which a significant change in system hydraulics was made. Or,e of the reactor coolant pumps was replaced, and the replacement pump produced additional flow. The LEFM measurements after pump replacement were in agreement with the predicted change, and the elbow flow meters indicated similar changes, but slightly lower flows than measured by the LEFM, The comparisons over 11 years show that the average difference between elbow meter flows and LEFM flows was less than 0.3% flow; the largest single difference was 0.5% flow, with the elbow meter indicating a lower flow than the LEFM. Another com'parison, performed before and after the pump replacement, showed that the two measurements agreed to within an average of 0.2t. on the ratio of flows when one and two pumps were l operating, thus further confirming the relative flow measurements from i elbow taps. These flow comparians are listed in the following tables.

, . ~ , . - - _ ._.__m,_ .-- . . _ . _ _ _ _ -- -

Page 3 of 3 LOOP / METER: A/LEFM A/ELB0W B/LEFM . B/ ELBOW DATE FULL POWER RCS FLOW MEASUREMENT COMPARISONS Feb 1980 97519 gpm (same) gpm 97950 gpm (same) gpm Jul 1981 98673 98309 9 763 97267 Aug 1991 98724 98557 97543 97607 1 PUMP /2 PUMP FLOW RATIOS Dec 1974 1.0819 1.0777 1.0852 ' 0875 Jul 1981 1.0794 1.0816 1.0820 1.0820 i

~~

Ref. 1: "Y1uid Meters, Their Theory and Application", 6th Edition, -g M Howard S. Bean, ASME, New York, 1971.

4 I

., .} i

e Attac!tment To TVA-91-3 49 i

Attachant B j Page 1 of 8

~

BASELINE CALORIMETRIC FLOW UNCERTAINTY COMBINED WITH ELBOW TAP NORMALIZATION UNCERTAINTY CALCULATIONS (EVALUATION OF THE EFFECTS OF [LQI PERFORMING A PRECISION RCS FLOW CALORIMETRIC FOR NOT;MALIZATION OF COLD LEO ELBOW TAPS EACH CYCLE) 1 The effects of not performing a normalization of the cold leg elbow taps exh cycle were evaluated.

I) The evaluation is baseri on tbe premise that a normnization of the Cold Leg l Elbow Taps is made to a previously performed Precision RCS Flow  !

Calorimetric (or group of ctJorimetrics). This effectively establishes the flow coefficient for the elbow. For subsequent cycles, it is assurt.ed tha* the transmitter is calibrated, either on a bench or in place, such that it responds within the calibratiot tolerunce tor a given ap input. , -.

~

II) For each cycle after the normalization, it is assumed that adequate RCS flow -

M is confirmed by verification that the indicated Ap of the Cold Leg Elbow Taps is as expected for that cycle and set of measurmant conditions, Le., if no systematic changes have been made to the pt ide, then indicated op should agree with previous cycle (s) indicated ., for that loop and transmir*er (within indication tolerances). If modifications have been made, the inoicated op should be within indication tolerances of the post mod' cation predicted value for the measurement conditions.

The evaluation was performed based on the following assumptions:

1) '.nstallation of the orginal, Westingbouse supplied Foxboro E13DH transmitters for measurement of cold leg elbow tap ap.

2) Meuurement and Test Equipment accuracy is per TVA specifications for other protectic functions. i.e., SMTE = SCA. In this instance SCA = 0.5 % ap span, thus SMTE = 0.5 % ao span.

3) The cold leg elbow tap op transmitter span is 110 % TDF.

4) The cold leg elbow tap Ap transmitters are calibrated in place once per 18 months.

Typically, this would occur during a reload shutdown.

J 5) The nominal arabient temperature at calibration is at least 70 "F and peak ambient i s temperature during normal operation is le , than 120 *F. .

it

1

u l

Page 2 of 8 Based on the above, calculations were performed for the Loss of Flow reactor trip and the RCS Flow indication error with use of the plant process computer. The calculations resuk . in the following conclusions:

A) Westinghouse has evaluated the RCS Flow measurement uncertainty noted in Sequoyah Technical Spe:i6 cation d 1.5, Table 3.2-1, and has found that the value remains unchanged, i.e., a cha .ge to  : Technical Specifications for this specific value is not required. The ve:- . :t is 3.5 % Flow. This is conservative in that the s '* ' adicated valt supported by Westinghouse was 2.4 % Flow (calculated value was 2.38 % Flow). The revised value is 2.4 % Flow (calculated value is 2,44

% Flow). The calculation is based on indication via the plant process computer. Any indication process in front of the process computer using a reasonably accurate DVM would result in a smaller indication error B) Changes are required to WCAP-11239 Rev. 5 and to the plant Technica! ,

Specifications as noted by the attached marked-up pages. Even though the changes are minor (only a slight increase in the Allowable Value for item 12, Loss of Flow,

_ Table 2.2-r, page 2-5 of the Sequoyah Unit I version), it is recommended that either y the Technical Specifications be modified, or the NRC advised and administrative control be put in place, prior to implementation.

j

4 WESTINGHOUSE PROPRIETARY CLASS 2 .3

's A r'fEdbZK .h I TABLE 3 9 LOSS OF FLOW Parameter Allowance' Process Measurement Accuracy ~

. . .e Idensity effects on Ap cell . ! 0.3 percent flowl'"

20.3 (precision f%w calorimetnc 12.3 percent flow

+ 0.05% flow bias) *"  ? 2.1

+ C:0/ 6 . C '/

Pnmary Element Accuracy (Elbow tap repeatability. t 0.5 percent Ap] " 10.3 c.c Sensor Calibration l' : o.T cr ee d v

• W1 -

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- b!!miec ;' ,- rc:-- E22!!:r a cA naw,d "' rn ( r,- ( e.r p.-e o v I , ge- 3 e, y Sensor Pressure Effects ..

e!!-

:ts N m,rm symatic- te- c:'ar,. t :ci " E eo.r fenaM 'e ' F1 .c ,0 e. o. 3 Sensor Temperature Effects 3p-22tg by 4erm slinti" to cafer4 met-u;?*18 L

• o i pr*T V Y" ',4 u M i C' 3 Sensor Drift [ t 1.0 percent Ap span} *"  : 0.6 Environmental Allowance 0.0 E M #

Rack Calibration Rack Accuracy [: 0.4 percent do span] *"  ! 0.2 M&TE [: 0.4 porcent ao span]" t0.2 Rack Temperature Effects (t 0.3 oercent op span]*"  : 0.1 Rack Onft 9 0.3 percent Ap span

10. '

In percent flow span (110 percent Thermal Design Flow) percent Ap span converted to flow span via Equation 3-26.1, with Fm.; = 110% and Fs = 100%

Channel StatJstical Allowart ' =

c. 3 + 0.3 c.3 0,3 , - a .c

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. + (0.2 + 0.2 + 0.1)2 + (0.1)'}"' + 0.0 + 0.0/ =,2:5% span ,

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!i TABLE 2 2-1(continued) 9 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Iotal Sep<or' functional Unit Allowance (TA) 1 Dritt (5) Trip 5etpoint Allowable Value .

12. Loss of Flow . 2.8 . JAnf p6' 2 90% of loop design .

M.6 2.fg- . 2)94% of loop design

., J, f flow

• flow" --

13. Steam Generator Water

~ Level-Low Low (Modified -

Barton Transmitters) '

a . Vessel AT Equivalent Io Power - 6.0 I' 1.74 1.6 Vessel AT variableinput

's 50% RTP Vessel AT variableinput "

s 50% MTP s trip setpoint +'2.5%

RTP Coincident with .

Steam Generator Water 14.8 13.34 ,2.0

- Level-Low-Low (Adverse) 214.8% of Narrow Range 214.2% of Narrow Range and Instrument span Instrument 3 pan Containment Pressure-EAM - 4.4 ' 2.94 1,5 s 0.5 psig 4

50 6 psig -

or- ,4 Steam Generator Water . 10.7 9.24- 2.0 i 210.7% of NarrowRange & 10.t% of Narrow Range ~

.' Level Low-Low (EAM) instrument span instrument span With a time delap (T s) .

sT,(Note 5) - J 5(1.01) T,(Note 5) .

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