ML20195J136

Summary of 981026 Meeting with Util in Rockville,Md to Discuss Plant Analysis of Replacement SG & Evaluation of Other Small Changes for FNP Units 1 & 2 Best Etimate Large Break loss-of-coolant Accident

ML20195J136
Person / Time
Site:	Farley
Issue date:	11/17/1998
From:	Jacob Zimmerman NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9811240204
Download: ML20195J136 (35)

Text

. _ __ . _ _ _ __. _._ _ _ _ _. _~ _

, November 17, 1998 LICENSEE: Southern Nuclear Operating Company, Inc.

FACILITY: Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2

SUBJECT:

MEETING

SUMMARY

OF OCTOBER 14,1998 l l

l The U.S. Nuclear Regulatory Commission (NRC) held a meeting with Southern Nuclear

_ Operating Company, Inc. (SNC), on October 26,1998, at NRC Headquarters in Rockville,  !

Maryland. The purpose of the meeting was to discuss SNC's analysis of replacement steam '

generators And evaluation of other small changes for FNP Units 1 and 2 Best-Estimate Large j Break Loss-of-Coolant Accident. l Enclosed is a list of attendees and material that was distributed during the meeting.

4 i in accordance with Section 2.790 of NRC's " Rules of Practice," Part 2, Tiiie 10 of the Code of  :

Federal Reaulations, a copy of this meeting summary and its enclosure will be placed in the NRC's Public Document Room.

ORIGINAL SIGNED BY:

Jacob 1. Zimmerman, Project Manager Project Directorate 11-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosure:

As stated cc w/ encl: See next page Distribution:

See next page q ;c ~~ , .

ICA; biLZ gL 31/

9811240204 981117 PDR ADOCK 05000348 P PDR To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy In OFFICE PM:Pjl) ld LA:PDil-2irl y D:PD)l4 3 7 l l l NAME JZim'nWe'rman:cn LBerry V((} HBerkcW DATE nl/298 h/O/98 ll/()/9ff / l98 I 198 / 197 DOCUMENT NAME: G:\FARLEY\MA0256MT. SUM ' OFFICIAL RECORD COPY

. c# %

[  % UNITED STATES g

a j NUCLEAR REGULATORY COMMISSION

! WASHINGTON, D.C. 20666 0001

%'OV/'

November 17, 1998 LICENSEE: Southern Nuclear Operating Company, Inc.

FACILITY: Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2

SUBJECT:

MEETING

SUMMARY

OF OCTOBER 26,1998 1

The U.S. Nuclear Regulatory Commission (NRC) held a meeting with Southern Nuclear Operating Company, Inc. (SNC), on October 26,1998, at NRC Headquarters in Rockville,  ;

Maryland. The purpose of the meeting was to discuss SNC's analysis of replacement steam  ;

generators and, evaluation'of other small changes for FNP, Units 1 and 2, Best-Estimate Large l Break Loss-of-Coolant Accident.

Enclosed is a list of attendees and material that was distributed during the meeting. .

in accordance with Section 2.790 of NRC's " Rules of Practice," Part 2, Title 10 of the Code of Federal Reaulations. a copy of this rneeting summary and its enclosure will be placed in the NRC's Public Document Room.

rvw J acob 1. Zimmerman, Project Manager Project Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364 l l

Enclosure:

As stated cc w/ encl: See next page

. Joseph M. Farley Nuclear Plant cc:

Mr. L. M. Stinson Rebecca V. Badham General Manager- . SAER Supervisor Southem Nuclear Operating Company Southem Nuclear Operating Company Post Office Box 470 - P. O. Box 470 Ashford, Alabama 36312 Ashford, Alabama 36312 Mr. Mark Ajluni, Licensing Manager Mr. D. N. Morey Southem Nuclear Operating Company Vice President - Farley Project Post Office Box 1295 Southem Nuclear Operating Birmingham, Alabama 35201-1295 Company, Inc.

Post Office Box 1295 Mr. M. Stanford Blanton Birmingham, Alabama 35201-1295 Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201, Mr. J. D. Woodard -

Executive Vice President Southem Nuclear Operating'Compa'ny Post Office Box 1295 Birmingham, Alabama 35201 J

State Health Officer

Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701 Chairman -

Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Regional Administrator, Region 11

. U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 Atlanta, Georgia 30303 Resident inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 '

Columbia, Alabama 36319 ,

4

.. . . . .. .- . .. ... . - ~ - - . -- . .

6 2.

Distribution for Meetina Summary dated November '17,1998 HARD COPY Docket File '

PUBLIC PD ll-2 Rdg.

J. Zimmerman

• OGC

-ACRS "

E-MAIL S. Collins /F. Miraglia (SJC1/FJM)'

B. Boger (BAB2)

J. Zwolinski (JAZ)'

' H. Berkow (HNB)

T. Martin (e-mail to SLM3)

E. Weiss R. Caruso F. Orr J. Staudenmeier R. Landry

D. Rasmuson M. Tschiltz (MDT)

L. Plisco, Rll '

P. Skinner, Ril '

l l

l l-4 t

LIST OF ATTENDEES U.S. Nuclear Reaulatory Commission .

E. Weiss, Omce of Nuclear Reactor Regulation (NRR), Division of Systems Safety and Analysis (DSSA), Reactor Systems Branch (SRXB)

' R. Caruso, NRR/DSSA/SRXB F. Orr, NRR/DSSA/SRXB

' J. Staudenmeier, NRR/DSSA/SRXB R. Landry, NRR/DSSA/SRXB D. Rasmuson, Office for Analysis and Evaluation of Operational Data, Reliability and Risk Assessment Branch Jacob Zimmerman, Farley Project Manager, Projecta Directorate ll-2, NRR -

Southem Nuclear Ooeratino Comoanv. Inc. I M. Ajluni, Farley Licensing Manager D. McComb, Steam Generator Replacement (SGR) Project Engineering Manager A. Alapour, Nuclear Fuel H. Mahan, Farley SGR Licensing Engineer .

Westinahouse M. Watson, Farley SGR Lead h

M. Nissley i S. Dederer Enclosure

. .s I

I i

j Analysis of Replacement Steam l Generators and Evaluation of l Other Small Changes for l . Farley Cnits 1 and 2 Best Esti~ mate LBLOCA l

4 Sue Dederer and Mitch Nissley 4

LObA Integrated Services i Westinghouse Electric Company i

l 1 .

~ ~

p Introduction m Farley Units 1 and 2 received approval for operation at.uprated power using the approved BE LBLOCA Evaluation Model a SNC has requested evaluation of two types of plant changes:

~

+ A major change to the primary system (RSG)

+ Changes affecting the containment pressure e

4 60

. I

Irrroduc-ion m Westinghouse developed a method to address these changes following approved WCAP-12945-P-A (CQD) guidelines i

+ Section 28-3 allows for evaluation of small changes to a plant analysis utilizing a subset of the original rtin matrix to determine effect on

~

PCT

< Definition of the "small change" and the size of the " subset" was not strictly defined a Purpose of this meeting is to illustrate application of Section 28-3 guidelines

. 2

Re olacemen-- S- eam Genera- or m Farley will replace Model 51 steam generators with Model 54F m Main difference between these SGs

. impacting LOCA is an increase in flow area and reduction in SG hydraulic resistance m Difference is comparable to reduced steam generator tube plugging

+ Effect is expected to be small W

3

I

!~ Re o acemen-- S- eam Genercror i

! m Change in hydraulic resistance is within the

range of flow path resistance ratio (RB) already analyzed -

1

+ Global model response surfaces remain valid i for RSG I

t 4

o ed

~

i Reo acemerr Steam Generator

~

m No change in operating parameters utilized in the Farley uprate analysis

+ Initial Condition uncertainty distributions remain valid for RSG m No change in power distribution parameters

+ Power Distribution response surface remains -

valid for RSG

\

~

Reo acement Steam Genera or a To assess impact of RSG, will incorporate RSG into reference transient to see impact s Farley BE analysis reference transient is a SPLIT break (C o =1.0) m Figures 1 and 2 compare RSG reference SPLIT transient with Model 51

+ Little difference in thermal-hydraulic characteristics

+ Additional 0,% SGTP case and break sizes were run to confirm that limiting tube plugging and break size are unchanged due to RSG <

g 7

Westinghouse Proprietary Class 3 4

Farley Reference Split Break--CD=1.0 Peak Clodding Temperature--Hot Rod RSC--Wodel 54F

---

O r ig t s e l SG--Wodel 51 2000 1800 '

l i 4 ,/

{hd .

e h 1600 ' *- l

,3: I n

uu t' .

1 I

v a

, 1400 -

u -

3 -

O .

e 1200 a .

E e -

+ _

1000 800 600 '' ' ' ' ' ' '

V 50 100 100 200 200 300 TIME (seconds)

Figure 1 Hot Rod Peak Cladding Temperature 7

i . 9 Westinghouse Proprietary Class 3 4

Farley Reference Split Break--CD=1.0 Average Core Channel LJquid Level--ch 10.11.12.13 RSC--Wodel 54F

---

O rig l e s i SC--Wodel 51 12 4

10 1

Oe u_ _

v -

e 6

. 8 m 8 l

4 e 2 .

ll l' I. e lh.. ,

&{ M 16 N

!}y ,

, , i 2 , d' I

L lIl' I

a.

l 0

O 50 100 ISO 200 250 300 TIME (seconds)

Figure 2 Average Core Channel Liquid Level S

i l~ Re o acemen-- Steam Genera- or u Thus far RSG has proved to be a small

impact l

m Further tested of this assumption by l . incorporating RSG into Superposition cases i

+ Used same parameters as utilized in uprate

. analysis

! + RSG Superposition and Reference Transient

incorporated into final Monte Carlo evaluation l- .

j. .

i 9

w. ---- -, e ,w. -m---ve-- 3 w ,,-i,.% ,e 3 -s - , evtrw* ,*,i- = ,:- z_

I acemen-- Steam Generator

~

Re o

! + Determine new 95th percentile PCT of 20560F

+ Compare this to equivalent Model 51 SG i PCT of 20600F

~

+ Confirms that incorporation of RSG was a i

small effect 9

9 1  %

I O OO

/0

-y- , , --y--- 9--%--p -

y,- , .,,sy- 3.-s -,,a. . mo s , . , , . , , ,,-w..,,w-.=_ , . - , -, + - . - , , yn - sy,. - - . . - - . .,

O Technica Basis for Memocology

~

m BE Methodology includes several assumptions, as c escribed in CQD, Section 26-5

m. Superposition accounts for uncertainties associated with

+ Assumption 2 and 4 for all plants

+ Assumption 7 for split break limiting plants such as Farley il

I i

I l

• l~ Tec1nica Basis "or Methodology l .

! m Superposition step increases the estimated j value of 95th percentile PCT as shown in

! Table 1

m. Indicates that Assumption 7 does not significantly affect overall PCT uncertainty l

+ Assessment of only the reference transient and superposition runs (all split breaks) is valid

\

12-

. . _ . _ . _ . . _ _ _ . . - . ~ . - - . . . . . . . ~ . _ _ .

Tec1nica Basis "or Ve- loc o ogy '

1 TABLE 1

Increase in 95"' Percentile PCT Due to Superposition

. BREAK TYPE Delta PCT DECLG +40 DECLG +84 DECLG +144 SPLIT +108 SPLIT +140 O

n

/3

-.e--n -,% , , ,- , -.,- , --,- v -. wv, ,-,- , - , - , -

l~

E Tec1nical Basis for Ve--hoc o ogy i

l m Additional assumption that all response j surfaces and uncertainty distributions remain valid for RSG -

! + Validity confirmed by demonstration that l thermal-hydraulic transient is not significantly affected

+ Change in break flow path resistance (RB) is minor I

l + Uncertainty in this assumption is accounted for by Superposition validation / correction step

/f

s i

'~

RSG Conclusions l m Incorporation of RSG had little effect on l Refere'n ce Transient PCT (5oF) l m Also had little effect on 95th percentile PCT i (4oF) l m Reference transient provides important

! . indication of.the effect of change on results

! a Reference transient PCT result can be used l to estimate'95th percentile PCT effect l

IS

l~ Otler Sma' C7anges j u A number of small changes made to the j plant which had to be evaluated

! + Increase in Containment Spray and reduction in j .

spray time impact the containment pressure i response calculated with COCO j .

m Containment pressure gradually reduced up l to a maximum of 1 psi l m Impact on BE LBLOCA must be evaluated es e

t l~ O~ler Sma C7anges i

l m Locate a sensitivity available for a similar

plant type and similar reference transient l m Compare thermal-hydraulic behavior of

! . reference transients i

! m If transients are similar, can apply l

sensitivity from one plant to another

! m Consistent with 10 CFR50.46 requirements l and current practice i

IT i

OtlerSmal C7anges

~

m A sensitivity study meeting these l requirements was available to apply to Farley

. + Similar plant (3-loop, SPLIT break limited)

+ Reference transients were similar (Figure 3)

+ Application of sensitivity study to Farley resulted in a 9oF penalty u Final PCT for Farley RSG

+ 2056oF + 9 = 2065oF

.e IS

t l

Westinghouse Proprietary Class 3 1

\

Comparison of Reforence Split Breaks Peak Cladding Temperature--Hot Rod R50 Reference Tre'nelent

---

3-l oop Re f e re nc e Tronalent 2000

' "a 1800 ' -"

s' " '"% '

e (t ,i ,

~

I 1,800

^

u. ,* /

v l ,

~ '

, 1400 ' l

3 .

l o -

L e 1200 a .

E 1000 800 800 ' ' ' ' ' '

U 50 100 100 200 2SD 300 TIME (seconds)

Figure 3: Peak Cladding Temperature F

l'

i t

Conc usions

~

j m Method for assessment of small changes is l dependent on magnitude of change and its

im. pact on the thermal-hydraulics

+ First step is to estimate the impact of Reference

Transient -

i

. + Incorporate the change into Reference Transient i

to confirm similarity in transients l + If effect is small, can apply reference transient

impact as an estimate of the effect on the 95th

!. percentile PCT l.

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

!~ Conclusions l .

l + If a similar plant sensitivity study is available, l could estimate the impact on the 95th percentile l PCT in the same fashion l m Method outlined in Section 28-3 of WCAP-12945-P-A (CQD) provides an l

a appropriate method for assessing impact of

small changes

e i'

2/

U, ,.

r .

Westinghouse Proprietary Class 3 Analysis of Replacement Steam Generators and Evaluation  !

of Other Small Changes for Farley Units I and 2 Best Estimate Large Break Loss-of-Coolant Accidents 1 Introduction The approved Westinghouse Best-Estimate Large Break LOCA Evaluation Model (BE LBLOCA EM) (References 1 and 2) has been used to analyze Farley Units 1 and 2 at uprated power (2775 MWt) over a range of operating conditions. A plant specific _WCOBRA/ TRAC model was developed and explicit treatment of the uncertainties in the required reactor parameters was considered. From these sensitivity studies, response surfaces were fit and were used to obtain predictions of the PWR response to variations in the dominant model and plant inputs. A Monte Carlo technique was used to sample from the uncertainty distributions for the various model and plant parameters. A large number of samplings was used to generate an overall PCT uncertainty distribution, from which the 50th and 95th percentile PCT values were obtained (Reference 4).

Section 28-3 of Reference 2 provides a general discussion of the strategy which will be used to assess the effects of plant or evaluation model changes 'on Large Break LOCA analyses performed with the BE LBLOCA methodology. Several potential changes and the method which would be used to assess them are discussed. The intent of the following discussion is to present the methodology used to assess the effects of a Replacement Steam Generator (RSG) program and other small changes for the Joseph M. Farley Nuclear Plant Units 1 and 2.

2 Description of Methodology The RSG program involves replacement of the existing Model 51 steam generators (SG) with new Model 54F steam generators. The main difference between the steam generators from a y LBLOCA perspective is an increase in the tube flow area, due to an increased number of tubes.

This results in a decrease in the steam gencator hydraulic resistance. Variations in the break flow path resistance ratio (RB) are explicitly modeled in the BELOCA EM, as described in

. Sections 25-3 and 26-4-4 of Reference 2. Applying Equation 25-3-21 of that reference to the Farley plants with the Model 51 and Model 54F steam generators shows that the RSG changes the nominal RB by only 0.01. This is small relative to the range already analyzed in the existing analysis. Therefore, it is expected that the RSG effect on the LOCA transient should be small, and the global model response surfaces previously generated for the Model 51 SG remain valid.

Due to the similarities in hydraulic factors, the RSG is accomplished with no change to the operating conditions set in the BE uprate analysis for Farley Units 1 and 2. Thus, the initial

9 l

Westinghouse Proprietary Class 3 i

condition uncertainty distributions previously generated from the uprate analysis remain valid for l the RSG application. , 1 I

There are also no changes to the parameters varied for the power distribution bias and uncertainty, so the power distribution response surfaces generated for the uprate program remain valid for the RSG application.

The Farley BELOCA uprate analysis indicates that the split break is more limiting than the double-ended guillotine (DECLG) break. As the first step in the assessment of the RSO. the limiting split break size (break area equal to 1.0 times the cold leg area) was reanalyzed for the Model 51 and Model 54F steam generators, using the approved version of WCOBRA/ TRAC with error corrections as reported in Reference 3. A maximum expected steam generator tube plugging level of 20% was assumed, based on previous Farley sensitivity studies which showed I maximum tube plugging to be more conservative.

Figure 1 compares the PCT transient between the reference Split break transients for the RSG and the Model 51 SG. The RSG transient shows little difference in. thermal-hydra,ulic characteristics from the Model 51 SG. This is confirmed in Figure 2, which shows the average core liquid level during the transient. The core liquid levels show that the timing of the core draining and refilling are virtually identical. The resulting limiting first reflood PCT for the RSG is a 5*F reduction compared to the Model 51 SG transient.

Additional cases were analyzed to ensure that maximum tube plugging remains limiting, and that the limiting split break size has not changed. Accordingly, single effect sensitivity studies with 0% SGTP, and break areas equal to 0.8 and 1.2 times the cold leg area were performed. These studies confirmed that the maximum tube plugging level is more conservative, and that the limiting split break size is unchanged.

3 Uncertainty Assessment The next step in the RSG assessment involves reanalysis of all of the uprate superposition cases, incorporating the RSG. These cases include parameters which were not varied in combination in the initial steps; for example a W_COBRAfrRAC calculation with PS3 and low average RCS temperature could be performed. The same parameters used in the original analysis are used for these RSG superposition runs, since all of the sensitivity studies used in the selection of these parameters have not been changed for the RSG.

l The results of these calculations are used in the fina. Monte Carlo evaluation to determine the estimate of the 95th percentile PCT.

2-

e Westinghouse Proprietary Class 3 I

The justification for using a reanalysis of the reference Split break transient and the superposinon cases to establish the RSG effect on peak cladding temperature is as follow.s. Sections 26 5-1 and 26-5-2 of Reference 2 describe the major assumptions considered in the BELOCA EM, and the methods used to justify or account for them. The superposition validaticn/ correction step accounts for the uncertainties associated with Assumptions 2 and 4 for all plan.s and also for Assumption 7 for those plants which are split break limited. A review of analyses completed to date indicates that the superposition validation and correction step has increased the estimated value of the 95th percentile peak cladding temperature by 40 to 144 F for plants that are guillotine break limited (three analyses), and by 108 and 140 F for plants which are split break l limited (two analyses). These values are relative to the 95th percentile PCT values estimated prior to the application of the superposition validation / correction step. These results indicate that j Assumption 7 does not appear to significantly affect the overall PCT uncertainty results. In the '

Farley RSG assessment, an additional assumption is made that the global model response surfaces, powe'r distribution response surfaces, and the initial condition uncertainty distribution determined with the Model 51 steam generators are also applicable with the Model 54F steam genciators. This validity of this assumption is confirmed by demonstrating that the change in the break flow path resistance ratio is minor, and that the thermal-hydraulic transient for the reference split break transient is not, significantly affected. Furthermore, any uncertainties in this assumption are accounted for by the application of Equation 26-5-2 of Reference 2, which is explicitly considered in the Monte Carlo simulation (Steps 30 and 31 of Section 26-5-2). Note that the uncertainty in the superposition correction factor is dependent on the number of test cases used. Figure 3 shows the multiplier on oy in Equation 26-5-2, when the sampled T,is equal to the average T,. It can be seen that the dependence of the predictive standard deviation on sample size decreases as the sample size increases.

4 Results of the RSG Analysis Incorporation of the RSG reference transient and superposition results into the final Monte Carlo simulation resulted in a 95th percentile PCT of 2056 F. Comparison of the RSG final Monte Carlo simulation results with the Model 51 SG analysis using the approved version of IV. COBRA / TRAC with error corrections as reported in Reference 3, shows a 4*F decrease in the 95th percentile PCT, from 2060*F to 2056 F.

The results of this analysis indicate that incorporation of the Replacement Steam Generator into the Farley Units 1 and 2 BE LBLOCA analysis has little effect on the thermal-hydraulics of the transients and thus can be evaluated following the guidelines approved in Section 28-3 of Reference 2. The assessment of the RSG impact included analysis of the reference split break transient (Co=1.0), several confirmatory single effect sensitivity studies, along with the eight superposition cases. This resulted in a 4 F decrease in PCF from the 95th percentile PCT for 3-

l-', * .

i l, Westinghouse Proprietary Class 3 l Farley with the Model 51 steam generators. Thus, the original assumption that the RSG would ,

be a small impact on the BE LBLOCA results has been shown to be valid.

5 Evaluation of Other Small Changes Examination of results of the RSG analysis shows that the relative effect of the RSG on the limiting PCT was closely predicted by the initial incorporation of the steam generator changes into the reference split transient. The limiting reference split break PCT was 5'F lower for the i RSG than the Model 51 SG results. Likewise, after incorporation of the RSG into the I

! superposition correction, the limiting 95th percentile PG was 4*F lower for the RSG than the Model 51 SG results. This similarity in results illustrates a simpler method for evaluating small changes in conditions used in the best estimate LBLOCA methodology.

For changes in conditions that are expected to not significantly affect the thermal hydraulic transient, incorporation of the change into the limiting break type reference transient (DECLG or Split break) yields an imponant indication of the effect on the results. The desired change in conditions can be incorporated into the reference transient, and a comparison made between the new case and the original transient. If the new transient shows similar thermal hydraulic response, an estimate of the effect of this change on the 95th percentile PCT could be issued based on the reference transient PC results.

Likewise, if a sensitivity to a change in a given condition has previously been evaluated for a

- similar plant type (3- or 4-loop), and similar reference transient (DECLG or split), the PG effect of the change to the reference transient could be used to issue an estimated PG impact on the 95th percentile PC for another plant.

An example of this type of small change for Farley Units 1 and 2 is an increase in containment spray flows. An evaluation of the impact of the increase in spray flows on the containment pressure transient for Farley resulted in a maximum reduction in pressure of 1 psi. To evaluate the impact of this change on the reference split break transient, an appropriate sensitivity study was required. A sensitivity study to a 2.7 psi reduction in containment pressure performed with ECOBRA/ TRAC on another three loop plant which has the split break as the limiting reference l

transient was found. This sensitivity to containment pressure change showed no significant effect on the thermal hydraulic transient. Using this sensitivity study, the impact of the gradual reduction in containment pressure is estimated to yield a 9 F increase in reference transient PCT for Farley. This 9'F penalty is then applied to the limiting 95th percentile PC as an estimate of the impact of this change. Thus, the limiting 95th percentile PCT for the BE LBLOCA ECOBRA/ TRAC analysis for the RSG including the estimated impact of the increase in containment spray flows becomes 2056 + 9 = 2065'F.

i

)

l .

I -

Westinghouse Proprietary Class 3 1 l l 6 Conclusions l Results of these assessments indicate that the approach outlined in Section 28-3 of Reference 2 is an appropriate method for determining the impacts of small changes in plant configuration.

Assessment of the impact of the change on the reference transient provides an indication of the significance of the chang:, which then determines the extent to which further evaluation becomes l necessary.

l l

l l

7 References

]

i

1. Letter, R. C. Jones (USNRC) to N. J. Lipamlo M), " Acceptance for Referencing of the  ;

Topical Report WCAP-12945 (P), Westinghouse Code Qualification Document for Best )

Estimate Loss-of-Coolant Analysis," June 28,1996.

l l

2. " Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Accident Analysis," WCAP-12945-P-A (Proprietary), Volume 1 (Revision 2) .and Volumes Il-V (Revision 1), March,1998.

3. Letter, H. A. Sepp GO to T. E. Collins (USNRC), "1997 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation l Models, Pursuant to 10 CFR 50.46 (a)(3)(ii)," NSD-NRC-98-5575, April 8,1998.

4. "Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Licensing Report,"

WCAP-14723, January,1997.

l l

l 2

l l

. . \

e.

Westinghouse Proprietary Class 3 Farley Refarence. Split Break--CD=1.0 Peak Clodding Temperature--Hot Rod

- RSC--Wodel 54F

---

cri s t no t sc--Wodel at 2000 1800 , '

, e d' \%d

- tt, 1600 '

, . i

'n

\ .

1 v

, 1400 u

s i Q i.

1200

a. -

l E {'

e -

s .

1000 800 i

,,,, ,,,, iiii iiii iiii iiii o so t90 1u0 290 200 300 TIME (seconds)

Figure 1 Hot Rod Peak Cladding Temperature

l.,.. \

t Westinghouse Proprietary Class 3 l

Forley Reference. Split Break--CD=1 0 Average Core Channel Liquid Level--ch 10.11.12.13 RSC--Wodel 54F

---

oris t se l SC--Wodel 51 12 -

10 C8 "

u. _

v e -

e 6

."._ - i I I

3 -l r, l

e -

l , l l

4 e!{ il l' ,

' : > }4l r. L dl A L l

l . 1 \

l' I l! I '

I

1 J. l 0

' l '' ' ' ' '

O 50 100 100 290 200 300 TIME (seconds)

Figure 2 Average Core Channel Liquid Level Westinghouse Proprietary Class 3 l

8 . ,-

\

'd.

'la3 .. ,

.. t . 4 , ,

1 . .. ,

, ; pt. .

3.20 -

3.00 -

2.80 -

n e

Q 2.60 z._

3 2.40 -

2 .

tt 2.20 -

2.00 -

1.80 -

1.60 - ..-------> n -

t 1 l I 4 6 8 10 Sample Size n Figure 3 Superposition Uncertainty Multiplier as a Function of Sample Size

_ _ _ _ _ _ _ _ .