ML20043H228

Informs of Issue Recently Identified During Startup of Facility from Cycle 4 Refueling Outage & How Issue Addressed to Support Continued Escalation to 100% Power,Per 900613 & 14 Telcons

ML20043H228
Person / Time
Site:	Sequoyah
Issue date:	06/18/1990
From:	Medford M TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9006220267
Download: ML20043H228 (16)

Text

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TENNESSEE' VALLEY-AUTHORITY V

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JUN 181990

'U.S. Nuclear-Regulatory Commission

' ATTN: Document Control Desk.

Washington, D.C.

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Centlemen:

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.In the Matter of.

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' Docket No. 50-327 Tennessee Valley Authority

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i SEQUOYAH NUCl. EAR PLANT (SQN) - UNIT 1 CALORIMETIC The' purpose.of this letter is to !nform NRC'of an issue =recently identified eduring startup of=SQN. Unit 1 from the Cycle 4 refueling outage and how it-was

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addressed to support' continued escalatlon.to 100-percent power.

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information has.been previously discussed with NRC'in continuing communication

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( 1, with the onsite senior resident inspector and in telephone-conference calls held between TVA-and NRC staff on June 13 and 14, 1990.

During the performance of the startup secondary and primary? calorimetric an unexplained

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increase in core delta T was. discovered.,This increased and anomalous delta T 1

measurement resulted in a reactor coolant system (RCS) flowrate calculation less?than the required technica11 specification (TS) value.

Preliminary assessment of: implemented changes to plant equipment, test data (recent and historical) -and core parameters provided high confidence that the RCS flowrate had not actua11y' degraded. However,-power escalation was. temporarily

suspended pending confirmation of the condition and cause. Status of this issue was communicated to the senior resident inspector, and_ ongoing communication continued throughout the issue investigation and resolution.

process. To ensure.that:no safety concerns existed during resolution-of the

. issue, TVA requested Westinghouse' Electric Corporation to evaluate the worst case ~ scenario of an. actual reduction in RCS flowrate. The resultant

-justification for continuedLoperation verified acceptability of operation at 100 percent power, and a copy:is enclosed for reference.

~As'a result.of numerous changes, which had been implemented to both primary and secondary. equipment, an in-depth investigation was initiated to determine

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the'cause of the-anomalous indications. Af ter extensive compilation and evaluation of data. TVA' concluded that RCS flowrate was in fact greater than-

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the TS :value' and that the earlier. calculated low value had-resulted from y

errors in indicated RCS hot leg temperature (Tn). The following summarizes' i

the. basis for this determination.

RCS flowrate, Macs, is normally calculated (inferred) from the following

. equation:

-(Macadh) prim.ry = (Mrwdh).. cone ry

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U.S. Nuclear Regulatory Commission The primary enthalpy change is derived from measured core delta T, i.e., Tn minus cold leg temperature. A secondary plant calorimetric is performed to 3

establish the right side of the equation, RCS hot and cold leg temperatures are measured, and the Macs is then calculated.

As previously mentioned, a variety of plant changes had been implemented ~

duringLthe Cycle 4 refueling outage, a number of which were considered to have potential'for impacting the. primary or secondary data.. Feedwater venturi's were cleaned, tested, and new calibration curves were provided; Eagle 21 protection sets were installed; RCS narrow range resistance temperature detector. (RTD). bypass manifold elimination was implemented, replacing the previous manifold with thermowells and fast acti:tg RTDs; Vantage 5H fuel was loaded to enhance fuel economy and reliability. Other key modifications such as upper head injection removal and boron injection tank deactivation were not considered to affect calorimetric data.

f The following actions were taken to verify secondary plant data validity.

A precision feedwater calorimetric was performed and verified with results utilizing condensate flows; calculated secondary plant power was verified consistent with both electrical output and turbine impulse pressure; the

- feedwater venturi calibration was checked and transmitter output verified to be consistent with raw differential pressure data; power output and data was reviewed against previous cycle data for consistency.

'In eval ~uation of primary side data, Tn RTD leads were lifted upstream of Eagle 21 processing to verify consistent input and output; power distribution was reviewed against previous cores and incore thermocouple (TC) maps; RCS elbow tap pressure-drop data was reviewed against previous data; core exit TCs were compared to Tn; and RCS parameters-were compared to design data and operating data for plants

of similar configuration.

Completion of these reviews confirmed the validity of secondary plant data and that errors in core RCS temperature measurement were not being introduced by

' Eagle _21 impicmentation. The review also confirmed that RCS flowrate had not

~ changed since initial startup as indicated by consistent elbow tap pressure

drop data.- 'The review did determine that RCS delta T had increased from previous operation values and Tn had increased from expected values as compared to core exit thermocouple data without apparent cause. Review of the previous equation shows that this indicated temperature increase thereby results in a corresponding lower calculated RCS flowrate; in fact, a significant reduction in calculated flowrate for a small increase in indicated y

delta T.- Close review of.this situation by Westinghouse's thermal hydraulics specialist confirmed TVA's' previous data evaluation results and concluded that Tn was'incicating erroneously high.because of changes in hot leg flow streaming resulting from indicated chanbes in core exit temperature distribution. A similar condition had been previously observed at several other' sites, although to a lesser extent. While still under evaluation by Westinghouse, it is co'nsidered to have resulted from depression of the radial power distribution at the peripht:y of the core. This type of profile causes colder water streaming along the bottom of the hot leg pipe.

Some of this colder flow is not included in the average temperature measured by the RTDs,

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.. U.S. Nuclear Regulatory Commission resulting;in the erroneously high Tn indication.

(Tu utilizes RTDs in three=thermowells-scoops located at 0 degrees, 120 degrees, and 240 degrees from the-top center of the loop piping.) Review of the SQN Unit i radial power profile substantiates the potential for this phenomenon.

A variety of options are under consideration by Westinghouse to address this situation for' affected plants.

In the interim, the impact of higher indicated Tn on protection and control tunctions was evaluated and determined not to represent a safety issu9 or to adversely' affect SQN analyses.

Discussions.were held with the NRC staff on June 13 and 14, 1990, to provide the staff vith information concerning this issue and respond to any questions that the staff may have had with regard to the present status or TVA's plans for increasing power on SQN Unit 1 to 100 percent. During the discussions, TVA provided a detailed description of the issue, investigation efforts, and resolution status. Also included in these discussions was a description of the-effect of operating with an increased indicated Tn on the reactor protection set setpoints and control circuits that use RCS average temperature (Tavg).and delta T as inputs.

It was concluded that all effects were in the conservative direction, and no safety concerns would be introduced by high indication.-

At the conclusion of the June 14, 1990, telephone call, NRC indicated agreement with TVA's approach in resolving this issue. The staff found that

. power escalation to 100 percent using the secondary side calorimetric program pcrformed by the plant process computer is acceptable. The staff did,

'however,. express continued interest in TVA's long-term resolution.

It is recognized that uncorrected, the higher Tu will result in adverse operational.ef fects,- e.g., reduced margin between 100 percent power and runback / trip setpoints and depressed actual Tavg and steam pressure.

Accordingly a number.of options are being evaluated for both short and long-term resolution.

In the short term TVA rescaled indicated delta T to slightly above 100 percent power when actual power is verifled by the secondary side calorimetric, to be at 100 percent. A process has been

. implemented to monitor delta T fer further changes so that appropriate scaling-changes.are implemented as the streaming phenomenon is expected to dampen over core burnup. Rescaling to slightly above 100 percent provides acceptable margin to runback / trip setpoints while additionally providing margin for potential decreases in delta T prior to rescaling. Long-term actions being pursued include development of scaling correction factors based on core exit thermocouples, reprogramming of Tavg control systems, and possible testing to validate and-further define the observed streaming phenomenon.

Actions are

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being coordinated with Westinghouse and will be carefully evaluated for full assessment on safe plant operation.

Preliminary results of ongoing eval'uations are expected to be available by mid July. Condition Adverse to Quality Report SQP 900286 documents this issue and will be used to track short and long term corrective actions. TVA will continue to keep the senior NRC resident inspector briefed on both related changes in plant status and long term issue resolution developments.

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U.S. Nuclear Regulatory Commission gl(lhl 18 j@[)Q In summary. TVA has determined both by calculation of RCS flowrate utilizing RCS elbow taps and by comprehensive review of data that the RCS TS flowrate has been satisfied. TVA has further determined that RCS Tn is indicating higher than actual bulk Tu because of fluid streaming, and that this condition does not compromise safe full power operation. Westinghouse has reviewed associated data and has concurred 51th these determinations. A variety of options are being evaluated bv Westinghouse and TVA for long-term resolution of associated issues.

If you have any questions concerning this submittal, please telephone M. A. Cooper at (615) 843-6651.

Very truly yours, TENNESSEE VALLEY AUTHORITY N

Mark O. Medford, Vice President Nuclear Technology and Licensing Enclosure cc-(Enclosure):

Ms. S.-C.

Black, Project Chief I-IV U.S. Nuclear Regulatory Commissica One White Flint, North 11555 Rockville Pike, MS 13H2 Rockville, Maryland 20852 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy. Tennessee 37379 Mr. B. A. Wilson, Chief of TVA Projects U.S Nuclear Regulatory Commission

-Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323

06/18/1990- 15:19..SGH SITE DIRNSITE Llc 615 043 71 9 P,. 23 '

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L Tennessee Valley Authority Sequoyah Nuclear Plant Unit 1 Reduced RCS Flow-l

.M Justification for continued operations

. Detr.Mr. Trudel '

In re nse to operatNnsofUkourrequest,attachedisajustificationforcontinued it l'at-a reduced RCS flow. Thisjustificationshowsthtt a-more rigorous safety evaluation would support a no significant hazards L

consideration pursuant to 10CFR 50.92 criteria.

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, If you have any coments or questions, please contact the undersigned.

p Very truly yours, "l

U-WESTINGHOUSE ELECTRIC CORP 0RATI0ld i

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Attachment to TVA 00 8%f page 1 of 11 Tennessee Valley Authority Semacyth Unit 1 Recuced RCS Flow Justification For Continued Operation 8@elARY' This justification for continued operation of Unit 1 reduced flow of 369 000 ops less

. uncertainties, and a,t 100 %, power,3.5 % Technical 8pe. Cycle 8 at a cification would support a no significant hazards consideration pursuant to 1 So.gt criteria.

The analys adjustment,is 356,000 gps.is flow value addressed, after the uncertainty This desument addresses the F5AR Chaptera 6 and 15 accident anal NSSS System compone,nts design trans,ienta, This upon the consideration that the Sequoyah Unit 1 These modifications include VlH Fuel, RTDE/tagle/N8 Removal and RWST,8eron Concentration Increase.

JUSTIFICAT!0N N888 System And 84uipment The potential impact of operation with the evaluated RCS coolant flow reactor coolant system components was addressed. A Peview of the therm desig arameters for the evaluated flow condition at 5% steam generator tuba 1 gging, indicate that the primary system temperatures remain assen i 11y unchanged plu gin thermal design (less than one degree change from the previous 65 of kha dS$$ sustest and equipment influenc)e,d by pr,ima parameters values Thus the ort inal analysis are not expec4ed to be affected by the' assumed flow condition.

For the secondary side of the steam generator, the steam pressure and temperature also show very little change for the reduced flow condition and thus the conclusions of the structural analyses would be expected to remain unaffected.

the analysis was based on a conservative set of operating co envelop the Sequoyah operating conditions by a large margin.

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- son SITE Dlh% SITE LIC tsiae 615 843 7109,,P c5 Attachment to TVA 90e847 page a of 11-by the thermal desisn parameters computations, the curre analysis would still bound the steam conditions coincident with the assumed reduction in primary system flow.

LOCA Accidents Alteration of the design basis reactor coolant system (RCS)l t not affect the following Less of Coolant Accident (t.0CA) re a e analyses! flow rate.do het log switchover to preclude boron precipitation, the post LOCA long term core cooling subcriticalit minimum safety injection flow. y, and post LOCA long term core cooling germane to these long term requirements. Steady stkte RCS flow rates are not analysis performed with the NOTRLMp Evaluation Mod,el in supoort ofFurther the assumed an RCS flow rate of 355800 gpm. removing upper head injecti rate therefore remains bounding. even at the evaluated low flowThe small break LO condition.

The above LOCA related accidents are not adversely affectoc, at 100% power by the evaluated low flow situation $on in the t d for lower power operation a'nd they are conservative rates is judged to have no. Moreover,t effect upon the reactor vessel a a reduct algnifican s ea y state flow loop LOCA blowdown forcing functions.

The large break LOCA 10CFR50.46 analysis for sequoyah I has also been considered.

To support upper head injection removal and Cycle 6 operation, Model at 105 steam generator tube plugging and a R Evaluation of 362000 opm.

for Cycle 5 has achieved little burnup and possesses very litt ow rate-heat relative to the 100% power basis of the large break LOCA analysis.

While the analysis of the freshly leaded fuel is therefore very conservative for the current situation previous analysis of Sequo

-1 Cycle 5 has demonstrated that once bu,rned fuel assemblies being yah U' tit temperature (PCT)thanthefreshassemblies. reinserted into the core ar ThegFtbyconsideringtheexpectedactualcorepeakingfactalcula 2013 beneficial reductions in both het assembly and core average seeking and in ors. -

i initial fuel pellet temperature are achieved.. Conversely steady. state flow rate will exact a penaltv in PCT on the, existing tie reduced anal sis result.

simiiarinvesse A previous sensitivity shudy performed for a plant flow causes a 17}F increase in calcu{ated PCT for a UN! im case.

8 Although Sequoyah now operates with UHI removed, the UHI imperfect mixin transient is similar enough to a non.UH! case to apply this sensitivity for JC0 purposes.

The 386000 1.66% from the analysis value and over a limited range it isgpm flow rate represe reasonable to extrapolate the, pertinent sersitivity linearly. judged to be -

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value of 8041 F or the 8equoyah Unit 1 CItyinPCTthereforeis26'F,81ar giving a new not PCT 1e

'limittar case. -Since no credit has been aken for the identified

. benefic' al aspects of th real life core peaking factors, 204PF is a suitably conservative va ue for the la e break LOCA PCT at the evaluated f1 and substantial ma in exists to h' it00 F regulatory limit.

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$equoyah Unit I with an accident anal sis basisconc uded above for the o her

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during Cycle 5 is Judged to be accept ble.

356066 gpa ACS flow rett

-i" Containment Related Analyses "Short Tera subcompartment Analyses The thort term eubecopartment analysis was performed at 10t% power with a

-thermal design flew of 365-000 tpM.

The analyses would not belm> acted by this flow chan a because a 2.8%This i reduction in RC8 flow would lave no significant effe t on the initial i

. system temperatures, so the initial system energy would remain he mass and energy releases would remain unchanged.i.e.1 8 sec.)..

Therefore the peak calculated differential pressures would remain unchanged from tbe current -

6esign basis subcompartment analyses for a t.6% decrease in RCS flow rate.

Long Term Containment Analyses

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LOCA Containment Integrity Analysis The most recent analysis-that was erformed for lon ters LOCA containment int r ty (CIA was dor a reductio in ice weight eference 2 This anal s a=Was a so performed et 102% ower with a t oraal design).

flow of Sol. 00 QPM.

flow reduction of 2.8% is an insi ni icant change.For the long term dosi n ba 9'

There would no

. chance to the initial system stored energy.'offect on the initial average sys O

forthelongtermLOCAtransienthavefour( The mass:and energy releases distinct phasest Slowdown, l

Refill. Reflood, and Post-affect the blewdown phase.Reflood Froth.._e initial system conditions Once t e-blo initial RCS conditions do not control the last three phases.of theown has been comp transient.

Since the initial conditions would be the same, the mass and the same as the current design basis analysis. Therefore. the pea R~'

eniculated containment pressure and temperature for the LOCA analysis would remain unchanged for the current design basis analysis.

Main Steam 11ne' Break (MSLB) tentainment Analysis The long term M81.8 containment analysis yields containment pressure and temperature profilea that are used to evaluate equipment qualification (EQ)forSequoyahUnits1&t.

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.,,,, a Attachment to TVA.ge 857 Page 4 of 11 8 team 11ne Break Outside Containment An analysis for steamline break outside containment was recently performed as part of the tagle 81 program.

The results of this analysis were used for EQ purposes in the steemitne valve vault and the auxiliary building.

58TR Accident l

The steam Generator Tube Rupture ($47R) accident is analyaed at a flow of 884,000 GPM. Therefore, the current SGTR FSAR analysis bounds operation at a flow of 886,000 SpM.

NON LOCA Accidents

., DN8 Considerations In order to determine the effects of the evaluated flow condition on the DHl yelated transients the core thermal limits and subsequent AT

• setpoint calculations w,ere examined. To accomodate the 2.5% decrease in tho' thermal design flow assumption used in the thermal hydraulic design of the fuel and the non-LOCA safety analyses, sensitivity studies as well as a Sequoyah specific evaluation have shown that 3.9% DNBR margin must be allocated to offset the Unit 1 Cycle 5 evaluated flow condition.

Allocation of this marcin-ensures that the DN8 segments of the Core thermal limits will noi, change.

However, the vessel exit boiling limit segments of the core thermal liinits do change.- These core limits are used in the calculation of the overtemperature and Overp6wer AT setpoint equation coefficients. The overtemperature and overpower AT setpointa protect against DNB and fuel centerline melting. respectively, for pressures as low as the Low Pressurizer Pressure reactor trip. The change to the vessel exit boiling 1

segments of the core limits impacts the Overtemperature setpoint equation such that the teefficienta used in the safety analyses do not remain valid

. at the lower flow condition. However, if the safety. analysis limit for the Low Pressurizar Pressure reactor trip is increased safety analysis equation will pecification equation as. then the Overtemperature &T technical s well as the remain valid.

Sufficient maelin exists between the current Technical Specification setpoint for tat Low Pressuriger Pressure reactor trip and the safety analysis limit such that the vessel exit-hoiling segments are adequately protested and the Technical Specification setpoint does not need to be changed.

The flow tendition also results in portions of the core thermal limits not being protected by the Overpower setpoint equation used in the safety analyses. Therefore, the safety analysis limit value for the K4 coefficient used in the Overpower AT equation was also reduced.

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Pete 8 of 11-As with the Low Pressuriser Pressure reactor trip safety analysis set!oint change, this change is covered by existing margin (ification value of K4 i.e., the magnitud of this change is small enough that the Technical Spec wouldnotbeimpacted).

Considering the previous dise.ussion, it'can be concluded that the DN8 design basis is met for the following FSAR Chapter 18 non LOCA safety-analyses

- Uncontrolled RCCA Bank Withdrawal from a Suberttical Condition

- Uncontrolled RCCA Bank Withdrawal e,t Power

- RCCA Hisalignment Partial Less of Flow

- Startup of an inactive Reactor Coolant Loop

- Loss of External Electrical Load / Turbine Trip

- Excessive West Removal Due to Feedwater system Malfunction Excessive Load increase Accidental-Depressurization of the Rekctor Coolant System a

- Accidental Depressurization of the11ain $ team System Inadvertent Operation of the Safety injection System at Power C

inte Loss of Forced Reactor Coolant Flow Sin le RCCA Withdrawal at Power e Mai Steaaline Rupture iLockedRoter(RodeInDNB)identRodWithdrawalatPower

. Steamline Break With Coinc Nnn DNR canaiderations In addition to'the DNB concerns discussed earlier the following evaluations are presented for:those licensing basla events which are not DNB rc.ated or for which DNS is not the only safety criterion to be met.

Uncontrolled Rod Withdrawal From a Soberitical Condition An. uncontrolled rod withdrawal from subcritical event results in't rapid i

uncontrolled-addition of reactivity leading to a power excursion (by a Section 15.2.1 of the FSAR. The nuclear power response is characterised t

very fast rise term)inated by the reactivity feedback of the fuel- (Dopp?ar temperature coefficient. The power excursion also causes a heatup of the moderator / coolant.

However, since the power rise is extremely rapid and short lived and reactor trip quickly terminates any tdditional power generation the

. thermal lag of the fuel pellet limits the moderator temperature, rise tc. a B

smail:value after reactor trip has occurred.

Thus, the nuclear power response is essentially e function of the Doppler feedback.

A 2.8% reduction in RCS flow would result in a slight increase in the calculated moderator tempefature rise.

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06/18/1990 15:24 -SQH SITE DIRNSITE LIC 615 843 71:9 P 29 AttachmenttoTVA>90dif Page 8 of 11 mere Do pler feedback due to hardening of the neutr Mducin the power excursion from that calcuitted in the F8AR.

The FSAR analysis shows that for e M 6ctivity insertion rate of 57pcm/sec,thehetspotpeakfuelaveraggFand665gF and clad varage temperatures en conservatively calculated to be 1818 A 2.5% flow reduction would degende the fuel.to coolant heat transfer by

, respectively.

4t most 2.85.

in the calculated fuel and clad temperatures when compared This would yield hot spot peak fuel averact and clad averepe temperatures which are s9111 well below fuel melt and lirc Hg0 reaction limits.

Note that in addition to the impact on the fuel / clad temperatures an

' of the RCS due to the primary to secondary power mismatch pressurization which results from this event is bounded by theHowever, the pressurisation experienced during the loss of load event discussed later.;

.ef the fue,l/ clad temperature or peak RCS pressure crite i

control-rod withdrawal from subcritical event.

l Boron Dilution The results of the boron dilution analysis would remain unchan modes of operation due to a reduction in reactor coolant flow.ged for all.

maximum dilution flow ratti RCS active volumes, and RCS boronThe concentrations are not impacted by a 2.6% reduction in RCS flow.

terminate the dilution event, the results presented in the Since remain-unchanged.

.of the licensing basis criteria following a boron-dilution e Loss of Lead I

The loss of load event presented in Section 16.2.7 of the FSAR may result from either a loss of external electrical load or a turbine trip.

result of a loss of load 14 4 rapid decrease in the secondary side heat The-removal, causing a rapid primary side hettup and pressurization.

i with and without pressure control. case are analyzed, beginning and end of-li Four 8

(bejlinning of-life, with pressure control) trips on low-low SG 1eveOf the fo whi e the remaining three cases trip on high pressuriser pressure. l, 0

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Atttchment to TVA 90 457 page 7 of 11 A reduction in the RC8 flow will result in 4 more relld primary side bettup and pressurisation than that shown in the FSM. However the effect Will be minor.

For those cases which trip on high pressu,rizer pressure,l energy input to the RC$.

the time to trip will be slight 1.y Mduced which will Msult in.

less tota the reduction in Ac8 flew would not be ex>ected to change

$4 level,hich the low low SG 1evel astpoint is time at w there is substantial margin to the primary nacaed.

In all four cases aswellassignificantmergintotheminim/secondarysidepnseurelimits um DNBR.

pressurizer will not fill asAn additional concern during the loss of load e insurge into the pressurizer.pmssuritation of the RCS results in an An RCS flow reduction leading to a rapid pressurization may result in a greater pressuriter insurge. men Howeveri as shown in the FSAR there is approximate of total pressuriser volume of margin to filling. 400cubicfeet(~2f%

than' sufficient margin to pr)essurizer filling to accommo,date e n is moro flow. reduction.

It is wor high pressurizer pressure th noting, that for the 3 cases which trip on into the pressuriser since there is less energy input to the RCS.a m Thus e 2.8% reduction in RCS flow would not result in th

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the licensing basis criterin following a loss of load event.

i Loss of Normal Feedwate'r/ Loss of AC Power to the Statio i

The loss of normal feedwater analysis in Section 15.2.8 of the FSAR presents the consequences of a complete loss of normal feedwater flow simultaneous to all four steam generators.

reactor coolant pumps (RCPs consting downsimilar except t todemonstratethatneither)theprimaryor. Tnese transients are entlyaed ucondary sides are pressuriser does, not fill.overpressurized thet the core is not adversely affect until-due to the rapid loss of steam generator invento continued heat. transfer to the secondary sidel,ed that L.

it is tripped on a low. low steam generator level signal.

It is anticipa E

.RCs flow would have little or no impact on the time of trip on low low n in steam generator level.

increase in the heatup of the RCS during the Inital phase of theT b

L transient.

i The increased heatup results in a decrease in the coolant heatup. However considerable margin exists to fillin t

-during this initial portion of the transient so that fg111ng would not occur.

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06/18/1990- 15:26 SQH SITE DIRNSITE Llc 615 843 71;9, P.11

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Attaehment to TVA 50-457 Pete e of Il During the long term portion of the transient, the peak RCS temperature (and resultant peak pressuriser water vo'lume) is reached ~ when tse heat removal capability of the auxiliary feedwater system matches the core decay heat entration, than antici sted loop flow Msistances,. the natural airculation be reduced y an amount proportional to the 2.9% thermal design flow reduction.

This slicht reduction in natural circulation flow at the peat RCS temperature condition would not significantly impair the heat transfer across the steam generator tubes thus,resulti temperature and peak pressuriser, water volume. ng in a similar hot leg

.Therefore, a 2.55 reduction in RCS flow would not result in the violation

' f the licensing basis criteria following a less of normal feldwater cr o

loss of AC power event.

part power loss of normal feedwater analyses completed for Delay.

1 Rupture of a Main Feedwater 1.ine The analysis in Section 15.4.t.2 of the F8AR presents th a double ended. rupture of a main feedline at full power.e consequences of Initially the RCS is cooled as the faulted steam generator blows down removing hea,t from the corres>onding RCS loop. However after the faulted steam generator empties, the reduction in secondary s,ide inventory results in inadequate' heat removal from the primary which i without offsite power available. RC$.n turn, increases primary system temperatures and pressurines the Two cases are examined: with an The case with offsite power is more :1 primary system. limiting since tie operating RCPs increase the energy addition The F$AR analysis demonstrates that sufficient auxiliary feedwater (AFN) is available to prevent overaressuriaation of the primary and secondar.y syatoms and to ensure that tie core remains intact and in:a coolable geometry.

does non occur in the RCS hot leg prior to AFW turnaround,This n

A 2.6% reduction in the RCS flow would result in a slightly more ranid heatup-of the RCS following the initial steam generator blowdown,the lower RCS flow would also result in.a-slightly higher hot leg temperature.

margin (126'F) to hot leg saturation throughout the transient.Howe i

L flow reduction would'not significantly degrade the heat transfer acrosaA 2.61i the steam centrator tubest thus, the long term RCS heatup calculated in the FSAR wl11 not be significantly impactsd.

Therefore, any increase in

to be negligible with respect to the available margin.the hot le l

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-Furthermore, the PSAR analyste demonstrates that there is a significant q

amount of margin to overpressurization of the primary and secondary systems.

the secondary side safety valves to relieve the pressure transient.A the primary side On margin to hot leg, the PORVs were conservatively modeled to minimize the boiling.

more than suffic'ent to limit the RC5 pressure transient.The analysis shaws tha A 2.8% flow reduction would not opspromise thia abilit,y. Thus, neither the primary or, secondary systems would overpressuriae.

Therefore, a t.5% reduction in RC5 flow will not result in the violation 1

of the licensing basis criteria following a feedline break event.

Thi analyses completed for the trip Time Delay.same discussion and conclu Locked Actor 8ection 15.4.4 of the F$AR presents the results of an instantaneous the seizure of a retoriseizure of an RCP rotor at full power with four RCps operating flow in the affected loop rapidly falls and theFollowing Rt3 temperature rises.

Reactor trip is prematly initiated on a low loop flow signal.

Analyses are done to predict tie peak RC8 pressure as Wall '

as the maximum metal to-water reaction and peak clad temperature.

Since the low flow setBotat is a fraction of initial loop flow, a t.6%

nuclear power and heat flux transients are unchanged. reduction in ow will not impact the time of trip, and thus, the RC8 flow will resuit in slightly higher system pressures than thoseHowever, the lower.

calculated in the F8AR.- The peak RCS pressure has been calculated to be stress limits are exceeded.8603 psia. Thia.value is well below the pressure-at wh reduce the available margin. A t.5% flow reduction would not significi.ntly The peak clad temperature abalysis performed-for the locked rotor evert calculates a value of 8026 F.

that DNS occurs upon the initiation of the event.This analysis conservatively assuma This assumption maximiaes the eehulated PCT and minimizes the impact of a flow reduction The calculated PCT of 1026 'F 15 Wellbelow the t'100since fut1-tc c shows that a slight increase in the temperature due to a 2.5%,RCS flow F limit and reduction can be accommodated. Thuarth locked rotor event will not exceed 1700 [F due to a 1.5% reduction inpeak clad the violation of the licensing basis criteria following a locke event.

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4 ess.enm.en, se m.ee.esq fa.e i.f n Rupture of a CentM1 Rod Drive Meehanism Housing (ACCA tjestion)

The RCCA ejection analysis is analysed at four conditions beginning and i

nd of life core physics characteristics, at het stro power and ful li teeSection15.4.0ofthePSAR). Theanalypisdemonstratesthatgross power L.

uti dama geometry,ge will not esturk will remain intact,that the ccM wi 1 remain in a coolable and that the RC in order to demonstrate that these criteria tre met Westinghouse app'its the followine, more restr' olive, criteria 1)

The ave e f.uel pellet enthalpy at tha hot spot is less than 200 etl (80Stu/lbe) t)

Fuel tt the het spot is limited to less then the innermost 10% e the fuel pe11et, 3)

Peak RC8 pressure is less than that which would cause stresses t6 e

exceed the Faulted Condition Stress Lluits.

l The red ejection event is characterited by a rapid power excursion teminated by Doppler feedback.

The reacter is tripped on high neutron flux (1ew setting for the sero power casesl result in a nductio high setting for the full powercases). A reduction in RCS flow wil fuel red te. coolant heat transfer.

This may Moult in an increase in the

. calculated fuel and 014d temperatuns as well as the ful stored energy duringanRCCAejection.

l As shown in the F8AR, the full power cans result in the highest fM1 pellet temperatures and a groacn criter'4 1 and I with the least amount of margin.

Exantnation of tsese cases reveals theti due to the rapid power and fuel temperature rise coupled with the thermal lag in the fuel pe11st itself the time at which the maximum pellet enthalpy and ful melt an calcula,ted to ecour is before any sijnificant amount of heat has Mached the coolant. Thus, a Mduction in the fuel to. coolant heat transfer due fuel mit calculated in the F8AR.to a 2.5% flow reduction should not impact th However it may be noted the. time at which the peak fuel temperature,s occur increase,that, should i

it is ex ected that sufficient margin is available to accommodate a 2.8% flod l

uction.

Yhe analysis of the pink pressure transient for the RCCA ejection event is discussed in WCAP.7655. Rev. 1.

A red 4ction in RCS flow could increass the primary side pressurisation by reducing the primary to-secondary side heat transfer.

However, due to the rapid nature of this event it is anticipated that any secondary side heat removal will las well behind the heat addition to the primary side.

I Thus it is judged that a t.$% flow reduction will have e minimal impact on I,he primary side peak pressure.

4 However ti WCAP peak RC$ pr)essun-7888, senral cases are presented which calculate the 3

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The most detailed of these cases calculatts a Ak ressuriser pressurti of

$400 psia.ThisismorethansufficientmarsitokheFaultedConditic.n' Stress Limits to accommodate e t.5% reduction in the RCS flow.

Therefore, s 2.8% reduction in RCs flow wtuld not neult in the vielstion' l

of the licensing basis criteria following a ACCA e.jection event.

$teamlineBreakMass\\tnergyReleaseInsidecontainment i

Genericsensitivitystudieshavtthewnths,tfourmajorfastersinf19 tat.o' the nionsa of mass and energy following a steamline breakt t

1.

Steam senerator inventory I.

Protectionsystemoperation 3.

State of the stoondary fluid blewdown l

4.

Primery to steendary heat transfer A t.6% reduction in RCS flow would not affect the first two factors anc!

t would have an insienificant impact on the last two factors. A decreast in RCs flow would tend to reduce the primary to secondary heat tnnsfer, thereby reducing.the steam pressure and toeparature during normal 094 4 tion.

Any reduction in the steendary side temperature and pressure wou d tend to minimite the mass and energy released during a steamline break event. As a result 4 2.5% reduct en in RC8 flow would not i

adver$41 affect the stet the break mass l

Chapter.t.1.3.11 of the uquoyah f$AR. / energy releases provided in steamiine enaa nass/iner, ati.ase outside Centainmens In order to address NRC concerns over the' effect of superheated steam l-release en the environmental qualification of equineont located outsida containment containment,were providedsteamlinebrokmass/energyreleassaforbreaksoutMs,e or Sequoyah in WCAP 10961. The impact of a a:

2.6% reduction in RCS flow on these mass energy releases wedd be insignificant for the same reasons as ci ed in the previous section for steamilne breaks inside containment.

The movement of the break locatien from inside to outside containment does~not invalidate any of the arguments made above.

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