ML20069J312

Safety Evaluation Supporting Tech Spec Changes for Licenses DPR-24 & DPR-27,allowing Operation at Reduced Power Level & Reactor Coolant Thermal Design Flow

ML20069J312
Person / Time
Site:	Point Beach
Issue date:	04/08/1983
From:	NRC
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ML20069J314	List: ML20069J312
References
TAC-48853, TAC-48938, NUDOCS 8304220127
Download: ML20069J312 (14)

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UNITED STATES

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j NUCLEAR REGUtATORY COMMIS0lON j

W ASmG TON, D. C. 20555

~j SAFETY EVALUATION (SE)

POINT BEACH NUCLEAR PLANT UNIT 1' OPERATION WITH REDUCED THERMAL DESIGN FLOW (TDF)

Introduction By letter dated September 17, 1982, Wisconsin El'ectric Power Company (licensee) requested changes to the Technical Specifications (TS) for Point Beach Nuclear Plant. Unit 1.

These proposed changes would allow operation at reduced power level (91%), reduced thermal design flow (TDF)

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(95%) and with an increased percentage of steam generator tube plugging (SGTP) (24%). These charig~ss were prompted-by the results of the licensee's previous calorimetric flow test (178,900 GPM or 100.5% of the TS limit of 178,000 GPM at 100% rated power) and the anticipation that further SGTP might occur as a r'esult cf the then forthcoming Unit 1 steam generator

. eddy current inspection.

Discussion The purpose of this SE is to present the NRC staff's evaluaticn of the Point 5each Unit 1 Safety Analysis for operation at reduced TDF presented in Attachment A of reference 2.

This SE also presents the NRC staff's evaluation of licensee-submitted sensitivity study results related to the licensee's previously approved large break loss of coolant accident (LOCA) analysis. Reference 2 proposts changes to the TS to enable operation of Point Beach Unit 1 at 91% rated power and a. minimum primary flow rate of 169,000 GPM or 95% of rated TDF, This is'the predicted primary flow rate 0304220127 830400 PDR ADOCK 05000266 P

PDR

2-if 24% of the steam generator tubes are plugged.

Attachment A to refer-ence 2 presents the non-LOCA accident and transient analyses for operation

' at reduced TDF. The following assumptions were utilized-Maximum core thernal power 1382 !!wt (91%)

TDF 169,000 gpm (95%)

Stean Generator Plugging Level 24%

T 572.9 F average A T 55.5 F Pr'inary Pressure E000 psia

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This SE includes our evaluation of transients and accidents that could be significantly affected by the above cperating conditions.

These include itss of external load, loss of normal feedwater, locked rotcr 2nd stetn line break. The folleuing transients are net adversely ano/cr sicnificantly affected by the above conditions and are therefore not further discussed: CVCS nalfunction, startup of a,n inactive reactor

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ccolant loop, reduction in feedwater enthalpy, excessive load increase,--

loss of reactor coolant flow.

Evaluation of Transient and Accideats 1.

Loss of External Electrical Load The FSAR analyses for the loss of external electrical load were perforued for four cases, i.e., with automatic reactor control and credit taken for pressurizer relief and spray, at both beginning of core O

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. life (BOL) and end of core life (EOL), end with manusl reactor control, no credit for pressurizer relief valve actuation and spray, at both BOL and E0L.

Initial conditions were assumed to be 102% power, 581*F T,yg, and 2250 psia.

No credit is taken for direct reactor trip due to loss of load, and it is assumed that the reactor trips on high pressure at 8.5 seconds.

For each case analyzed the DNBR increases during the transient. Tne most severe peak pressure is 2514 psia for the manual control case at BOL. The primary safety valves lift but no water, relief.

occurs.

In reference 2 the licensee compares this transient at reduced TDF cor.ditions with the FSAR analysis and indicates that the pressure rise will be slightly nore rapid because of reduced TDF and extensive stean generator plugging. The tine to reach the high pressure trip set point would be less than for the FSAR case and therefore the total energy input to the coolant would be less.

However, this is no't a good comparison sin.ce the ESAR anelysis was perforned at 2250 psia, while creration at reduced TDF will be at 2000 psia.

Since the high pressure '

trip setpoint is the same for both operating pressures (i.e., 2400 psia)-

the tire to trip may actually be longer for red'uced pressure operation.

The staff questioned the licensee's assumption that reactor trip due to high pressure would be more rapid for the. reduced TDF case than during conditions described in the FSAR analysis and requested additional ~confirma-tory information just.ifying the analysis.

In Reference (13), the licensee 6

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indicated that for operatiori at reduced pressure, DUB is limi+ing, while peak pressure is limiting for operation at rated pressure. The reactor would trip on overtemperature delta T at reduced presisure. The consequences 1

of this transient with regard to ONBR would be bounded by the " uncontrolled rod withdrawal at power" (URWAP) analysis. The URWAP analysis at reduced TDF, presented in Reference (2), indicates that minimum DNBR does not fall below 1.3.

We find, based on our review of previous analyses and the additional information provided by 5e' licensee, that'the consequences of loss of load transient at reduced TDF will not result in taacceptable fuel per-formance and that the prirenry system pressure will not exceed allowablea values. The licensee's loss of. external load analysis is, therefore, ~

acceptable.

2.

Less of Nornal Feedwater The FSAR analysis for the loss of normal feedvater. transient assumed this esent to occur at 102% power, at minimum ncrmal team generator level, and loss of the reactor coolant pumps.

The reactor trips on low-low steam generator level. One auxiliary feedwater pump starts one minute after the low-low stean generator level signal, delivering flow to one steam generator.

Secondary steam relief is via the steam generator safety valve.

The tube sheet of the steam generator receiving eexiliary feedwater flow is always covered. The capacity of one auxilicry feedwater pump is sufficient to prevent water-relief frcm the primary relief ard safety valve. The peak Tavg. is 609*F at about i ~

hcur af.ter transient start. The peak pressurizer lig;id volume is '790 I

ft.

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. Ir reference 2 the licensee ir.dicates that at reduced TDF the maximun 3

pressurizer liquid volume could be 905 ft, which is less than the 1000 ft capacity of the pressurizer and therefore no reanalysis was necessary. This is based on an assumption "that the average temperature would increase 50 percant due to flow reductions", which we interpret to mean that the primary tenperature rise during this transient is 1.5 tites the temperature riss at rated conditions. We conclude,that this is a censervative assumption since the total primary rass reduction due te E45 steam generator tube plugging is 8%.

Hceever, the licensee did not address in Reference 2 the effect of 24%

reduction in heat transfer area on the capability for shutdown without primary water relief utilizing one steam generator and one auxiliary

'feedwater pump. We then requested that the licensee provide additional information regardirg the effect of reduction in steam generator heat transfer area.

In Reference (13), the licensee indicated that the decrease in heat transfer area is offset by the decreased decay heat since operation is'at reduced power level. Therefore, the pressurizer will not be filled, the RCS pressure limit would not be reached, and the tube sheet would remain covered, with only one steam generator and one AFW pump available.

We conclude, based on our review of previous analyses and the additional information provided by the licensee, that the consequences of loss of ncrmal feedwater transient will not result in unacceptable fuel performance ar.d that the primary system pressure will not exceed allowable values.

The 1icensee's analysis is, therefore, acceptable.

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Lccked Roter The FSAR analysis for the locked rotor accident assumes that seizure of

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one reactor coolant pump (RCP) shaft occurs at 102% power. React 6r trip j

occurs on a low flow signal.

Upon reactor trip, it is assumed that the most reactive RCCA is stuck in its fully withdrawn position. The time from pump seizure to initiation of control rod motion was assumed to be.

0.9 secords, The licensee has stated that test data indicates a

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neasured tirne interval of 0.45 seconds from the tine the low flow trip setting is reached until the instant the rods are released. Another 0.1 second is assumed for the nterval between piamp seizure and reaching the l

low flow trip set point, for a total of 0.55 seconds. Thus 0.9 seconds is conservative (Reference 6).

No credit was taken for the pressurizer relief valves, pressurizer spray and steam durp. The licensee assured offsite pcuer to be available and continued oceration of one RCP.

This is further discusse.d below.

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The FSAR Enalysis shcwed the peak pressure to be 2778 psia.

We consider j

this value acceptable, since it is below 120% of design pressure l

(service limit "C" of the ASI'E code), and thus naets the acceptance criteria of the June 15, 1982 revision of Staccard Review Plan (SRP)

Section 15.3.3-15.3.4 for peak pressure. The results of this analysis further indicate that about 221; of the fuel rods reach a DNBR less than 1.3 crd about 15% of the fuel reds reach a CUBR less than 1.0.

This occurs for a very short tire period (about 2 secords).

Peak clad surface ten perature is 1522 F.

The licensee indicates that the peak l

i clad surtece ' temperatures are below the threshold for metal-water _

rsction, and therefore, the results are not unacceptable.

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Re#ererce 5 contains the licerste's analysis of this event at reduced operating pressure. While the resulting peak pressuie is lower than in the FSAR analysis, the results of the DNB calculations are more s'evere, 1

predicting that 63% of the fuel rods reach a DNBR of less than 1.3.

The licensee indicates that this analysis was performed on a highly ccrservative basis, since the coolant pressure increase as a result of e

the transient was ignored, and rods for which the fluid conditions are

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beyond the range of the DNB correlation were assigned DNB ratios less than 1.3.

In view of the high percentage of potentially danaged fuel as aresultofthispostulatehaccident,theslaffhasperforned independent site boundary calculations to determine whether the radiological consequent 2s of the postulated accident meet the guideliner of 10 CFR Part 100. The licensee is adopting standard technical specification (STS) limits for primar coolant iodine.

Assuming prinary coolant STS limits.and 63% fuel cladding damage, the radiological consequences at the site bcundary would be less than a smal1 fraction of The licen'ee's analysis did not assune the 10 CFR Part 100 guidelines.

s loss of offsite poter (LOOP) and thus the radiological consequences could conceivably be higher if LOOP occurred. Therefore, a liniting calculation was also performed assuming that all-the fuel cladding is damaged.

The resulting site boundary dose is still less than the 10 CFR

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Part 300 guidelines, and thus meets the acceptance criteria of the June 15, 1982 revision of SP.P Section 15.3.3-15.3.4 for site bcundary dose.

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. With regard to the effect of operation at reduced TDF and pcwer, references 2 and 4 indicate that the expected fuel and clad temperatures would remain about the same as at rated conditions, since the effect of reduced flow would be offset by the lower power _ level.

The effect of reduced flow and primary mass would not be detected by the core in the time frame of interest since th'e peak values are reached in considerably-less than one loop transport time constant. We concur with this assumption and find that the locked rotar analysis is acceptable.

4.

Stean Line Break (SLB)

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The FSAR steam line break analysis was performed using 7 combinations of break sizes ar.d initial plant conditions, including large breaks upstrean cro downstream of the flow limiting nozzle, one and two-loop operation, offsite power available and enavailab'a, and a break equ'ivalent to steam release through one steam generator safety valve.

The analyses were performed assuning end of core life, hot shutdown with the most reactive rod stuck in its fully withdrawn pos,ition, 2.77%

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shutdcun reactivity, and one safety injection pump failing to function. ~"

The most severe cese involves a break upstream of the flev limiting nozzle, two loops in operation, and loss of offsite power, and results in a peak power after return to criticality of 2"'.

For the break downstrean of the ficw limiting nozzle, peak power after return to criticality was of the order of 10%.

Utilizing the MacBeth critical heat flux correlation provided acceptable DNBR values for all the transients analyzed.

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For operation at reducec pressure Erd temperature, Reference 5 indicetes that, os a result of slightly less stored energy in the coolant system, o

cooldown is slightly faster and the resulting thermal' power is about 1%

higher. itinimum DNBR is still above 1.3.

Reference 2 provides reanalyses at reduced TDF for the following cares:.

large SLB inside containment with and without offsite power,.large SLB outside containment with and without offsite power, and a break size equivalent to one open safety valve.

The assunptions for these analyses are:

end of core life, t most reactive rcid stuck in its fully withdrawn position, one safety injection train not functioning. While the text of reference 2 states that the initial shutdown rargin is 2.77%

fcr all cases, figures 4 through 8 indicate the initial reactivity to be

'O for the first 4 cases. As noted above, the FSAR analyses were all perferred with en initial reactivity of.0277. We consider that this.

may be a nore conservative assumption for the SLB initici conditions.

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The licensee was requested to clarify these apparent discrepancies and jus'tify the assumptions utilized or submit new analyses.

In Reference (13), the licensee provided additional information including better figures, and indicated that the assumptions utilized are consistent with Reference (14). The minimum DNBR for the postulat'ed breaks was greater than the 1.3 limit. Reference (2) indicates that the increased level.of steam generator tube plugging would, because of reduced heat transfer coefficient and flow, result in slightily lower peak power levels when compared with the FSAR analysis.

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- The res;1:s of the above SLB analysis indicate that the largest power

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excursicn occurred for SLB inside containment with outside power available.

DflBR remained above 1.3 for all runs. We conclude that, based on pre-vious analyses and additional information provid.ed by the licensee,_the consequences of an SLB at reduced TDF will not result in unacceptable i

fuel performance. The licensee s SLB analysis is, therefore acceptable.~

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5.

Large Break Loss of Coolant Accident. (LOCA) Analysis The licensee has indicatqd_that the most applicable existing large-break LOCA analysis to be used for operation with reduced TDF was performed with IS% steam generator tube plugging and peaking factors (F ) equal to 2.32.

g References 10 and 11 contain such analyses for operating pressures of _

2250 psia and 2000 psia, respectively.

Reference 10 indicates that 100%

TDF would be obtained even with 18% tube plugging.

Reference 8 contains cur evaluation of the LOCA analysis submitted in Reference 10. The staff concluded that a large-break LOCA during operation at Point Beach Unit 1-while a: a primary pressure of 2250 psia and with up to 18% tube plugging would result in a peak clad temperature (PCT) of 20530F and would be in ~"

conforcance with 10 CFR Part 50.46 criteria. Reference 8 provides a LOCA analysis for reduced pressure operation. PCT is calculated to be 20520F.

A correction factor of 600F is applied to these numbers to account for the effects of upper plenum injection.

(Ref. 8) The criteria of 10 CFR rart 50.46 are still met.

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. The licensee has not performec a detailed calculation of PCT for the large break LOCA at reduced TDF and pressure operation. The licensee has, however, submitted the results of sensitivity calculations for PCT-at 91% rated power, 95% TDF, 24% tube plugging and 2000 psi RCS pressure. Assuming an F f 2.52, the resulting PCT, when corrected for 0

upper plenum injection, was 2188 F.

This is close to the allowable

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limitsof2200 Fin 10CFikPart50.46.

In Reference 12, we questioned._,

and the use of sensitivity analyses to correct for an increased F0 indicated concern about the.small cargin to PCT limit of 2200*F. We also questioned assumptions regarding linearity ano superposition of

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sensitivity analyses and requested clarification regarding apparent inconsistencies in the analysis.

In subsequent conversations with the 1

licensee we ir.dicated that our major concern is the utilization of F of q

2.52, which apparently increases PCT by 200 F over.a utilization of Fq of 2.3?..

In Reference (13), the licensee indicated that 18% steam generator tube -

plugging is the maximum expected. Reference (13) clso indicates that, '

while operation at reduced power and TDF would involve higher ratios of peak to average linear power, the peak kw/ft value for 91% power and 95% TDF would be bounded by the full power case. F6H (enthalpy rise' hot channel factor) is also slightly lower for the reduced TDF case. The submittal also indicates that the effect.of small flow or coolant tem-perature changes.on PCT is small when compared to the effect of the power level. _For the large cold leg break LOCA, the core flow reverses direction o

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within 0.1 seconds of the LOCA transient,-so the initial flow rate through the core is of relatively little importance. The licensee concludes that the LOCA analysis @ 100% power and TDF bounds the 91% power, 95% TDF, F q 2.52 case and, therefore, the latter meets 10 CFR Part 50.46 criteria.

l In Reference (13), the licensee'also proposes to administratively limit

• power to 84%. This would'be equivalent to an F LOCA limit of 2.32 for,_,

g 91% power. Such power limitation would reduce linear kw/ft by about 7.5%.

We conclude that operation at a maximum power level of 84%, a minimum TDF of 95%, and 18% SGTP would not result in. values exceeding.10 CFR I

j Part 50.46 acceptance criteria in the event of the design base LOCA.

Subsequent to the licensee's submittal of Reference 13, a calorimetric flow test was perfctmed at Unit I which indicated that TDF was slightly.

above 100% with 1/4 SGTP. However, there is a concern th.at if additional tube plugging is required, TDF may be reduced to less than 100%.

l Conclusion Based on our above evaluation, if future calorimetrics indicate-that TDF is less than 100% but not less than 95%, SGTP is hela to a maximum of IS%, and the power level is administratively limited to a maximum of 845,thelicenseeneednotsubmitadetailedLOCAanalysis. However, if operation at lower TDF than 95%, higher power levels than 84%, and higher SGTP than 18% is contemplated, the licensee must furnish to NRC for-approval a LOCA analysis for the new operating conditions.

Principal Contributors:

i T. G. Colburn B. Sann L

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RtFERENC_E5 l

1.

Point Beach Nuclear Plant Units 1 and 2 Final Safety Analysis Report (updated) 2.

WEPCo. letter of September 17, 1982, forwarding Technical

, Specification Change Request No. 85 for Reduced TDF 3.

WEPCo. letter of October 15, 1982 forwarding additional'information-for Technical Change Request No. 85 4.

Point Beach Nuclear Plant Unit 1 " Steam' Generator Repair Report" August 1982 5.

WCAP 8151, Point Beach Unit 2 Low Pressure Analysis, June 1973

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6.

UEPCo. letter of November 9,1982, forwarding responses to cur request for additional infornation regarding the locked rotor analyses.

7.

NRC Confirmatory Order for liodification of Point Beach 1 License, Novemoer 30, 1979.

3.

NP.C Podifying Confirmatory Order of November 30, 1979, dated January 3, 1980.

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9.

ImC letter to LEPCo. dated April 29, 1980, forwarding Amendner.ts

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No. 44 to Poin.t Beach Unit 1 Facility Operating License.

10. WEPCo. letter of November 19, 1979, forwarding ECCS Reanalysis for 18% Steam Generator Tube Plugging Limit, Point Beach Unit 1.

11.

WEPCo. letter of flovember 27, 1979, forwarding Low Pressure ECCS

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Evaluation for 18% Steam Generator Tube Plugging, Point Beach Units 1 and 2.

12.

NRC letter to UEPCo.-dated November 19,_1982, _ forwarding request for additicr.a1 information regarding the' LOCA analysis.

13. WEPCo. letter of December 10, 1982 forwarding additional information for Technical Specification Change Request tio. 85.

14. WCAP 9226 " Reactor Core Response to Excessive Secondary Steam Releases" I

January 1973 l

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