ML17206A824

Forwards Responses to Reload Safety Evaluation Questions

ML17206A824
Person / Time
Site:	Saint Lucie
Issue date:	05/23/1979
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
L-79-136, NUDOCS 7905300014
Download: ML17206A824 (11)

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. BOX 529100 MIAMI,FLA 33152 FLORIDA POWER 8( LIGHT COMPANY Nay 23, 1979 L-79-136 Office of Nuclear Reactor Regulation Attention: Hr. R. ll. Reid, Director Operating Reactors Branch P4 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear ter. Reid:

Re: St. Luci e Unit 81 Docket No. 50-335 RS~ (Suessions Attached are responses to several questions generated by the NRC staff relative to the St. Lucie Unit 1 Cycle 3 Reload Safety Evaluation (RSE).

This submittal (questions 2.15, 2.18, and 2.21 through 2.28) completes our response to all questions in this area received to date from the NRC.

Please note that a revised response to question 1,.11 is also attached.

.Our earlier response to this question inadvertently omitted information we intended to transmit.

Very truly yours, obert E. Uhri g Vi ce Presi dent Advanced Systems & Technology REU/DKJ/cph Attachment cc: Nr. James P. O'Reilly, Region II Harold F. Reis, Esquire peal s9 V905.3 00 0/P PEOPLE... SERVING PEOPLE

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Question 1.11 Xn the response to Question 1.6 you state that the implied peaking factors F anQ Fxy are higher than the Tech Spec values by the uncertainty in these factors. Xf this is the case, then the Safety Analysis value should be greater than the Tech Spec value by this uncerLainty. Explain why the Tech Spec and Safety Analysis values are equal in your reloaQ application.

Res onse 1.11 The measurement uncertainties are accounted for in the safety analyses.

The measurement, uncertainty factor for Pz is listed in Table 6-1 of the Reload Application (Statistical Component of FNz 9 95/95 Confidence Level) along with other calculational factors which are also useQ in the sageLy analyses. These factors are not zepeateQ in the tables'resented in Section

'7.0 of the Reload Application, since these tables only core'parameters anQ initial conditions for list the various tran-sients, consistent with 'the format useQ i,n the Reference Cycle Application anQ the PSAR. Even though Table 6-1 lists the uncertainty factor on Pz as 1,0513 (CL'alculated value), a 1.06 uncertainty factor on Fz was'conservatively used along with the other factors listeQ in Table 6-1.

As indicated in Table 7-2, the 'initial peak linear heat rate assumed for NON-LOCA safety analyses is 16.KW/PT. The peak linear heat rate for LOCA is 14.8 KW/PT (Chapter 8, page 67).

Xn order to account for the uncertainties on Pq, the Tech Spec ex-core monitoring band on ASX has been based on the LOCA peak linear heat rate limit of 14.8 KW/PT and also incorporates the 7-o uncertainty on Fq as described in Section 9.1 of CHNPD-199. When monitoring on In-coze Instrumentation, the'Fq un-certainties are applied as indicateQ in Section 4.2.1,4 of the Tech Spec. Because the initial peak linear heat rate used in the safety analyses is higher and therefore more deleterious than allowed per Tech Spec by at, least the uncertainty factor, the uncertainties on Pq are accounteQ for in the safety analyses.

Question 2.15 The Cycle 2 parameters were supposedly computed with fine mesh 2D PDQ. Elow were you able to compute different F and P

xy with only a two dimensional model?

Res onse 2.15 Our estimates of the integrated radial peaking factors are "

obtained. by weighting planar radial peaking facLors calcula-ted with fine mesh 2-D PDQ for different planar regions of the core. The values used in the safety 'analy es are always higher than the esLimated inLegrated radials.

f - - p 1

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Question 2.18 (Cycle 3 Application Page 36) The steady state LAIR to fuel centerline melt is given as 21.0 kw/ft. Xs this a best estimate value, or is estimate. Xf it conservative relative to the best it is conservative, what is the best estimate value?

Response 2.18 The power given for fuel -centerline melt of 21.0 kw/ft. is conservative.

Power to fuel centerline melt is calculated by the FATES fuel performance code as described in CHNPD-139-P-A. The FATES code was developed for the purpose of providing conservatively high fuel temperatures for a given power level for use in safety calculations, where conservatism is essential. Best estimate values of power to melt: are not required and therefore they are not addressed.

Question 2.21 (Cycle 3 Application Pages 34 and 42)

On page 34, the TAB for CHA Withdrawal Event. is 52 psia and on page 42 the TAB for RCS Depressurization Event, is 30 psia.

What are the previously calculated values for these two parameters?

Response 2.21 The pressuxe bias factors of 52 psia for CEA Withdrawal event and 30 psia for the RCS Depressurization event were calculated for. Cycle 2. Since the values are con,ervative for Cycle 3, these values were unchanged from Cycle 2 to Cycle 3.

Question 2.22 (Cycle 3 Application Page 34)

For th'e CEA'..Withdrawal Event it is indic:ates that the approach of the L1IR to KW/FT limit is considered. Xs this a consideration for both the HFP and EIZP cases? What criteria are used to determine that the LHR is acceptable? State the coxe parameter values used to determine that the LHR SAFDL (No centerline melt= ) is not violated.

l For the CEA Nithdrawal event, the approach to the fuel centerline mel.t was considered for both the HFP and the 1IZP cases. Xn Table 7.2 of the licensing submittal, 21 KN/FT was conservatively established as'the steady state linear heat,. rat>> for centerline melt. For the HFP case, the event is terminated by the Variable High Power trip or the local power density trip prior to reaching 21 EN/FT. For the HZP case, since reactor power rises rapidly foz a vexy short period of time befoxe the powex. transient is the integral of the heat generation rate is the con- 'erminated, trolling parameter. Therefore, for the HZP case the total energy generaLed and the corresponding temperature rise at the hottest spot of the coze was calculated for the duration 'of the transient to demonstrate that the fuel centerline melt was not violated.

Coze parameter values used for the CEA Nithdrawal event are listed in Table 7.2 of the Cycle 3 license submittal..

Question 2.23 (Cycle 3 Application Page 16)

Heze the predicted boxon worthsare given as 90 PPM/-"o- .hp fox.

BOC and 80 PPM/-"o- hp for EOC. Xn Cycle 2 these items were 88 and 77 respectively. Does the boron worth play a significant,

role in the Safety Analyses. (For example, the Steam Line Break) Xf so, w'hat values of boron worth were, used in the safety analyses f'r Cycles 2 and 3?

Response 2.23 The inverse boron worth has no impacL- on the design basis events (DBE's) in the safeLy analysis, except for steam line break and boxon dilution. For the steam line break event, EOC conditions produce the limiting case. For both Cycle 2 and Cycle 3,, a high EOC inverse. boron worth of 82 ppm/-..Pp was assumed. This is conservative in comparison to the EOC values of 80 ppm/":hp for Cycle 3 and the 77 ppm/Shp for Cycle 2. For boron dilution it is conservative to use the minimum inverse boron worth. Since those values assumed for the Cycle 2 analysis (Table 7.2-1 of Cycle 2 submittal) bound those values for Cycle 3, the boron dilution event.was not reanalyzed.

Ques. Lion 2.24 (Cycle 3 Application Page 55)

Here the HFP F is given as 1. 39 and a HZP F is given a's 1.59.

7A Are these best, estimates or conservative values? Xf they are conservative relative to the best, estimates, what are the best estimates values?

Response 2.24 These are conservative values assumed in the safety analysis for t: he CEA Ejection event and they are based on axial power shape at the limit:ing T allowed by the LCO's. Only at full power can a best esLimate F. be determined, since one can define the operational mode. The &est estimate for F at full power, ARO, is 1.12. The actual. F. at HZP is a function .7 of many variable such as the mode of operation, time at zero power, past operating hist:ory and xenon disLribution. Therefore, is not. possible to quote one single best estimaLe value.

it EIowever, the F will always be less than 1.59.

Quest:ion 2.25 (Cycle 3 Application, Page 16)

Here the neutron generation time R"'s given as 33xE-6 sec at EOC and 28XH-6 sec at BOC. should play a role in the CEA EjecLion Event, but it does not appear in the for this event. What value of R* was used in the Safety list of parameters Analysis?

Based on parametric studies where R* was varied over the range of BOC and EOC values, there is a 0.05 cal/gm difference in total energy deposit:ed for the CHA Ejection event. For Cycle 3 the nominal val.ue of 30 x 10 sec was assumed in the safety analy. is.

Question 2.26 (Cycle 3 Application Page 55)

Here a HZP Azimuthal Power Tilt of 1.10 is used in the Safety Analysis. What is t: he best estimate of the 'illP Azimuthal Power Tl.lt-Res onse 2.26 Our best est.imate HZP azimuLhal power to 1.10 used in the safety analyses.

tilt is 1.03 as compared Question 2.27 (Cycle 3 Appl.ication Page 33)

I The Three Pump Plenum Factor used in the Safety Analysis is 1.09. What is the best est.imate value for this parameLer?

Response 2.27 The.use of the plenum factors in the thermal-hydraulic. code, COSI10, is discussed on Page 7-12 of CHNPD-161-P and in Section 3.2.4.3 of CHNPD-199. The 1.09 three pump plenum factor is a conservative value used in the safety calculations where con-servatism is essential. Best estimates are not required and therefore not addressed.

Question 2.28 The malfunction of one-steam generator events are not re-analyzed in either Cycle 2 or Cycle 3 Reload Application.

Thus, the only analysis must be in the FSAR. We are unable to find the malfunction of one-steam generator events in the FSAR Chapter. 15 Table of Contents. Xndicate where these events are discussed in the FSAR.

Res onse 2.28 The Asymmetric Steam Generator transients (Loss of Load to One Steam Generator, Excess Load to One Steam Generator, Loss of Feedwater and Excess Feedwater to One Steam Generator) were not in the set of design basis events analyzed for the FSAR.

Subsequent to completing the FSAR, but prior to generating the Cycle 1 setpoints, these events were incorporated into the list of the Design Basis Events as documented in CENPD-199-P. Analyses were done at that time to demonstrate that the required overpower margins for these. events were less than for other AOO's, such as the Loss of Flow. The input parameters for this original analysis bound those for Cycle 3 as indicated in Table 1. This is the basis for the statement in the Cycle 3 license submittal that these events were not. reanalyzed.

TABLE 1 Key Parameter s Assumed in the Analysis of the Malfunction of One Steam Generator Parameter Units C~cle l ~CC l8 3 Xnitial Core Power MW 2621 2621 0

Xnitial Core Xnlet Temperature F 544 544 Initial Reactor Coolant psia 2200 2200 System Pressure Moderator Temperature x10 h,p/'7 '. 5 -2.5 Coefficient Doppler Coefficient 0. 85 0.85 Multiplier