ML20012D966

Forwards Annual ECCS Evaluation Model Changes Rept,Per Revised 10CFR50.46.Info Includes Effect of ECCS Evaluation Model Mods on Peak Cladding Temp Results & Summary of Plant Change Safety Evaluations

ML20012D966
Person / Time
Site:	Farley
Issue date:	03/22/1990
From:	Hairston W ALABAMA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9003290167
Download: ML20012D966 (11)

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10CFR50.46:  !

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Docket'Nos. 50-348- l

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.50-364 U. S. Nuclear Regulatory Commission  !

X ATTN: Document Control' Desk

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Washington, D. C. 20555 .

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l ' Gentlemen L- ,

Joseph H..Farley Nuclear Plant - Units 1 and 2 10CFR50.46 Annual ECCS Evaluation Model Changes Report 5 ,

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The October 17,~1988 revision to 10CFR50.46 required applicants and {

holders of operating. licenses or construction permits to. annually notify the Nuclear Regulatory Commission (NRC) of insignificant errors and. changes  :

in the ECCS-Evaluation Hodels. Enclosed is Alabama Power Company's report i l .in compliance with this requirement for Joseph M. Farley Nuclear Plant

. Units:1 and 2.-

a  ;

Attachment A provides information regarding the effect of.the ECCS

. Evaluation Model n.odifications on the peak cladding temperature (PCT)-

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results: reported in Chapter 15,LSections 3:and 4 of the Joseph M. Farley Nuclear Plant?Unitsil and-2 Final' Safety Analysis Report (FSAR).

. Attachment B provides a summary of the plant change safety evaluations performed to'date (i.e., since 1987 for large break-LOCA and since the t L' original plant analysis for small break LOCA) under the provisions of l 10CFR50;59 that' impact PCT. Please note'that the facility change safety evaluations included in Attachment B reflect.only those which result-in 7

?non-zero PCT impact assessments.- This information package' constitutes Alabama Power. Company's report to the NRC as part of annual' reporting

-required by 10CFR50.46(a)(3)(ii). q

,It has been determined that compliance with the requirements of

'10CFR50.46(b)(1) continues to be maintained when the effects of plant design changes performed under 10CFR50.59 are combined with the effects of the ECCS Evaluation Model modifications applicable to Farley Units l'and 2. .,

~ '. This. determination is based on the fact that the total large break and t x

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i c.c U. S. Nuclear Regulatory Commission Page-2 - ' March 22, 1990-9 small break resultant PCTs reported in Attachment B (i.e., including ECCS  ;

Evaluation Model modifications and all non-zero PCT penalties associated '

with the design change safety evaluations performed under 10CFR50.59) are vell belo'v the PCT limit.of.2200'F.

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It should be'noted that the Farley Nuclear Plant Vantage-5 fuel reanalysis.  !

by Westinghouse for both large break LOCA and small break LOCA is currently '

. scheduled to be completed in early 1991. This will redefine the reference analysis results.

-If there are any questions, please advise.

Respectfully submitted,  ;

g/J. /4&

V. G. Hairston, III 4

VGH,III/JARingd 15.13 Attachments cci Mr..S. D. Ebneter l Mr. E. A. Reeves D

Mr. G. F. Maxwell-l l

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ATFACEMENT A - 0 l

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.! EFFECT OF-VESTINGBOUSE ECCS EVALUATION ~NODEL NODIFICATIONS 4 -ON THE LOCA ANALYSIS RESULTS FOUND IN CHAPTER 15, SECTIONS 3 AND 4~0F THE l'

FARLET UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT =

li 1,

i: RACKGROUND ll 1 The October 17, 1988, revision to 10CFR50.46 required applicants and ,

holders'of operating licenses or construction permits to annually notify 'l l .

the Nuclear Regulatory Commission (NRC) of-insignificant errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models. Reference 1 i, - j H - defines n'significant error or change as one which results in a calculated t P fuel peak cladding temperature (PCT) different by more than 50'F'from the L temperature calculated.for the limiting transient using the last acceptable  ;

model,-or as'a cumulation of changes and errors such,that the sum of the r absolute magnitudes of.the respective temperature changes is greater than i l; 50*F.

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.In-References 2 and 3, information regarding modifications to the Westinghouse large break and small break Loss of Coolant Accident (LOCA)

ECCS Evaluation Models vas-submitted to the NRC. The following presents an i assessment of the effect of the modifications to the Westinghouse ECCS

- Evaluation Models on'the LOCA analysis results found in Chapter 15, ,

1 Sections 3 and 4 of the Farley Units 1 and 2 Final Safety Analysis Report l - (FSAR).

I' s ll LARGE BREAK LOCA 4 L

i ECCS' EVALUATION MODEL-The-large break LOCA analyses-for Farley Units 1 and 2 vere examined to assess the effect-of the applicable modifications to the Vestinghouse large break.LOCA ECCS Evaluation Model on PCT results reported in Chapter 15, Section 4 of the FSAR. The large break LOCA-analyses results vere.

calculated using the 1981 version of the Vestinghouse large break LOCA ECCS Evaluation Model incorporating the BASH analysis. technology. .The analysis

. assumed.the following information important to the'large break LOCA i analyses:

o Core Power'= 1.02

• 2652 HVT

[ o 17x17 Standard Fuel Assembly o- F, - 2.40  !

o F-delta-H - 1.62 ,

L -o Steam Generator Tube Plugging Level - 10% (Uniform) l o Upflow Downcomer Modification Assumed i

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• ATTACWmWT A

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Page 2 For Farley Units 1 and 2, the limiting break resulted from the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD_= 0.4. .The calculated peak cladding temperature was 2013'F.

! The modifications to the Vestinghouse ECCS Evaluation Models discussed in Reference 2 which could affect the large break LOCA analysis results documented in Chapter 15, Section 4 of the Farley Units 1 and 2 Final Safety Analysis Report are described below.

. MODIFICATIONS TO THE BASH ECCS EVALUATION MODEL 3- Several improvements were made to the BASH computer code to treat special-analysis' cases which are related to the tracking of fluid interfaces and i

g which could affect the plant analysis results. ,

L i

! 1)_ A modification to prevent the code from aborting was made to the ,

heat transfer model for the special situation when the quench front  :

region moves to the bottom of the BASH core channel. The quench '

heat supplied-to the fluid node below the bottom of the active fuel ,

was set to zero.

2) A modification to prevent the code from aborting was made to allov ,

negative initial movement of the liquid /tvo-phase and liquid-vapor ,

interfaces. The coding in these areas was generalized to prevent >

mass imbalance in the special case where the liquid /tvo-phase interfSce reaches the bottom of the BASH core channel.

3) Modifications to prevent the code from aborting were made to increase the dimensions of certain arrays for special applications.

4) A modification was made to write additional variables to the tape of information to be provided to LOCBART.

L 5)~ Typographical errors in the coding of some convective heat transfer '

s + terms were corrected, but the corrections have no effect on the BASH analysis results since the related terms are always set equal' l- to zero.

l b) A modification was made to the BASH coding to reset the cold leg conditions in a conservative manner when the accumulators empty. '

The BASH model is initialized at the bottom of core recovery with the intact cold legs and lower plenum full of liquid. Flov into the downcomer then equals the accumulator flow. The modification removed most of the intact cold leg water at the accumulator empty L time by resetting the intact cold leg conditions to a n4gh quality l- two phase mixture.

In a t'pical BASH calculation, the downcomer is nearly full when the accumulators empty. The delay time, prior to the intact cold l

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gyym Page 3 leg vater reaching saturation, is sufficient to allow the dovneomer to fill from the addition of safety injection fluid before the vater in the cold legs reaches saturation. When the intact cold leg water reaches saturation, it merely flows out of the brr.ak.

The cold leg vater, therefore, does not affect the reflood transient.

However, in a special case where a substantial time was required to fill the downconer after the accumulators emptied, the fluid in the intact cold legs reached saturation before the downcomer filled, which artificially perturbed the transient response by incorrectly altering the downcomer fluid conditions causing the code to abort.

o The Parley Units 1 and 2 LOCA analysis results could be affected by the

. modifications specified in items 1, 2, 3, 4, 5 and 6 above. While there is no adverse effect on the PCT calculation for the majority of the changes which apply to Farley Units 1 and 2 discussed above, a conservative estimate of 10'F vill be assessed and tracked for use in determining the

available margin to the limits of 10CFR50.46.

l MODIFICATIONS TO THE VREFLOOD COMPUTER CODE In Reference 2, modifications are reported for the 1981 ECCS Evaluation Model which form the fundamental framework for application of the BASH i

methodology. The modifications made to the VREFLOOD computer code l described for the Vestinghouse 1981 ECCS Evaluation Model were carried into the VREFLOOD computer code used for BASH analyses.

In the BASH methodology, the VREFLOOD code is only used to calculate the bottom of core recovery time. Therefore, this modification has no effect on the BASH ECCS Evaluation Model calculations.

l MODIFICATIONS TO THE LOCBART COMPUTER CODE 1

No modifications have been made since those outlined in Reference 4.

CONTAINMENT PURGE LINES OPEN EVALUATION L

' A safety evaluation of the effect of containment purge lines being open coincident with the large break LOCA event was performed. An estimate of the large break LOCA analysis PCT results was projected. The evaluation determined that the large break LOCA analysis PCT results could be affected by a 4'F increase.

RESULTANT LARGE BREAK LOCA PCT is discussed above, modifications to the Vestinghouse large break LOCA ECCS Evaluation Model could affect the large break LOCA analysis results by i

altering the PCT as shown below.

A. Analysis Calculated Result 2013'F B. Modifications to Vestinghouse ECCS Evaluation Model + 10'F C. Containment Purge Lines Open Evaluation + 4*F D. ECCS Evaluation Model Modifications Resultant PCT Ril7'F

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ATTACWGINT A Page 4 i

SMALL BREAK LOCA r

ECCS EVALUATION MODEL ,

The small break LOCA analyses for Farley Units 1 and 2 vere also examined to assess the effect of the applicable modifications to the Vestinghouse i ECCS Evaluation Models on PCT results reported in Chapter 15, Section 3 of i the FSAR. The small break LOCA analyses results vere calculated using the 1974 small break LOCA ECCS Evaluation Model incorporating the VFLASH  !

analysis technology. For Farley Units 1 and 2, the limiting size reall i break resulted from a six-inch equivalent diameter break in the cold leg.

The calculated PCT vas 1712'F. The analysis assumed the following information_important to the small break LOCA analyses  !

< t o Core Power . 1.02

• 2652 MVT o 17x17 Standard Fuel Assembly '

o F, - 2.32 o F-delta-H = 1.55 o Auxiliary Feedvater Flov . 1050 gpm (Total) ,

The modifications to the Vestinghouse ECCS Evaluation Models discussed in "

References 2 and 3 which could affect the small break LOCA analysis results  ;

found in Chapter 15, Section 3 in the Parley Units 1 and 2 FSAR are '

described belov.

VFLASH ECCS EVALUATION H0 DEL CODE l Following the accident at Three Mile Island Unit 2, additional attention was focused on the small break LOCA, and Vestinghouse submitted a report, VCAP-9600 (Reference 5), to the Nuclear Regulatory Commission (NRC) detailing the performance of the Vestinghouse small break LOCA Evaluation Model which utilized the VFLASH computer code. In NUREG-0611 (Reference 6), the NRC staff questioned the validity of certain models in the VFLASH computer code and required licensees to justify continued acceptance of the '

model. Section II.K.3.30 of NUREG-0737 (Reference 7) clarified the NRC post-THI requirements regarding small break LOCA modeling and required licensees to revise their small break LOCA ECCS models along the guidelines specified in NUREG-0611.

Following the issuance of NUREG-0737, Westinghouse and the Vestinghouse Owners Group decided to develop the NOTRUMP (Reference 8) computer code for use in a new small break LOCA ECCS Evaluation Model (Reference 9). The NRC approved the use of NOTRUMP for small break LOCA ECCS analyses in May 1985.

Since approval of the NOTRUMP small break LOCA ECCS Evaluation Model in b

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ATTACWIBrf A Page 5 I

1985, the VFLASH computer code has not been maintained as part of the Vestinghouse ECCS Evaluation Model computer code.

In section II.K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small break LOCA analysis using an NRC approved small break LOCA Evaluation Model which satisfied the requirements of NUREG-0737 section II.K.3.30. NRC Generic Letter 83-35 (Reference 10) relaxed the requirements of item II.K.3.31 by allowing a more generic response and providing a basis for retention of the existing small break LOCA analyses.

, Provided that the previously existing model results were demonstrated to be conservative with respect to the new small break LOCA model approved under the requirements of NUREG-0737 section II.K.3.30 (NOTRUMP), plant-specific analyses using the new small break LOCA Evaluation Model vould not be required. In VCAP-11145 (Reference 11), Vestinghouse and the Vestinghouse owners Group demonstrated that the results obtained from calculations with VFLASH vere conservative relative to those obtained with NOTRUMP.

Compliance with item II.K.3.31 of NUREG-0737 has been completed by referencing VCAP-11145.

Vestinghouse, therefore, has not been modifying, investigating, or evaluating proposed changes to the VFLASH small break LOCA ECCS evaluation model. There are no modifications to report.

SBLOCTA-IV COMPUTER CODE The following modifications to the LOCTA-IV computer code in the small break LOCA ECCS Evaluation Model have been made.

1) A test was added in the rod-to-steam radiation heat transfer coefficient calculation to preclude the use of the correlation when the vall-to-steam temperature differential dropped below the useful range of the correlation. This limit was derived based upon the l physical limitations of the radiation phenomena.

There is no effect of this modification on reported PCT's since the erroneous use of the correlation forced the calculations irito aborted conditions.

l L 2) An update was performed to allov'the use of fuel rod performance data from the revised Vestinghouse (PAD 3.3) model.

An evaluation indicated that there is an insignificant effect of this modification on reported PCT's.

3) Modifications supporting a general upgrade of the computer program were implemented as follovs:

A) removal of unused or reduadant coding; B) better coding organization to increase the efficiency of calculations; and C) improvements in user friendliness i) through defaulting of some input variables,

11) simplification of input, iii) input diagnostic checks, and iv) clarification of the output.

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{ 1 l Verification analyses calculations demonstrated that there was no i E

effect on the calculated output resulting from these changes. <

4) Three modifications improving the consistency between the  !

Vestinghouse fuel rod performance data (PAD) and the small break ,

-LOCTA-IV fuel rod models were implemented.

5 A) The form of the equation for the density of Uranium Dioxide was "

corrected to calculate thermal expansion only in two F-dimensions, which is consistent with the way in which the fuel  ;

g rod is modeled in the LOCTA codes.

B) The correlation for the specific heat of water vapor at tempera. ,

tures over 1590' was improved.

l C) An error in the equation for the pellet / clad contact pressure was  :

corrected.

i The Uranium Dioxide density correction is estimated to have a maximum PCT benefit of less than 2'F, while the contact resistance modification has no PCT effect since it is not used.  ;

SAFETY INJECTION BACK PRESSURE FIX EVALUATION I A safety evaluation of the effect of spilling broken loop safety injection I water to containment back pressure instead of to reactor coolant system back'  !

pressure was performed. An evaluation of the effect of this modeling change -

on the small break LOCA analysis PCT'results was performed as documented in

, section:15.3.1.2.2 of the Farley Units 1 and 2 FSAR. The evaluation

. determined that the LOCA analysis PCT results could be affected by a 46'F increase. This 46' increase has been previously reported in an Alabama Power Company to NRC letter dated January 14, 1988.

RESULTANT SMALL BREAK LOCA PCT y

E As discussed above, modifications to the Vestinghouse small break LOCA ECCS  !

Evaluation Model could affect the small break LOCA analysis results by altering the PCT as shown below, l- A. Analysis Calculated Result 1712'F B._ Modifications to Westinghouse ECCS Evaluation Model -

2'F C. Safety Injection Back Pressure Fix Evaluation + 46'F

• E D. ECCS Evaluation Model Modifications Resultant PCT 1756TF L

L CONCLUSIONS An evaluation of the effect of modifications to the Vestinghouse ECCS t Evaluation Model, as reported in Reference 2 and 3, was performed for both the large break LOCA and small break LOCA analysis results found in Chapter 15, Sections 3 and 4 of the Farley Units 1 and 2 FSAR. When the effects.of the ECCS model changes were combined with the current plant analysis results, it var letermined that compliance with the requirements of 10CFR50.46 vould be maintained. ,

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REFERENCES ,

1. " Emergency Core Cooling Systems Revisions to Acceptance Criteria,"

Federal Register, Vol. 53, No. 180, pp. 35996-36005, dated September 16, 1988.

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2. NS-NRC-89-3463, "10CFR50.46 Annual Notification for 1989 of Modifications in the Vestinghouse ECCS Evaluation Model," Letter from i V. J. Johnson (Vestinghouse) to T. E. Murley (NRC), dated October 5,

• 1989.

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3. NS-NRC-89-3464, " correction of Errors and Modifications to the NOTRUMP Code in the Vestinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant," Letter from V. J. Johnson (Vestinghouse)-

to T. E. Murley (NRC), dated October 5, 1989.

4. VCAP-10266-P-A, . Revision 2 (Proprietary), VCAP-10267-A, Revision 2 (Non-Proprietary), Besspiata, J. J., et al., "1981 Version of the Vestinghouse ECCS Evaluation Model Using the BASH Code," March 1987. ,

5. " Report on Small Break Accidents for Vestinghouse Nuclear Steam Supply

• System," VCAP-9601 (Non-Proprietary), June 1979, VCAP-9600 '

(Proprietary), June 1979. -

6. " Generic Evaluation of Feedvater Transients and Small Break  :

Loss-of-Coolant Accidents in Vestinghouse Designed Operating Plants,"

NUREG-0611, January 1980.

7. " Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.

8. "NOTRUMP - A Nodal Transient Small Break and General Network Code,"

VCAP-10079-P-A (Proprietary), VCAP-10080-A (Non-Proprietary), <

Heyer, P. E., et al., August 1985.  !

9. "Vestinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," VCAP-10054-P-A (Proprietary), VCAP-10081-A (Non-Proprietary), {

Lee, N., et al, August 1985.

10. " Clarification of THI Action Plan Item II.K.3.31," NRC Generic Letter  !

83-85 from D. G. Eisenhut, November 2, 1983.

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11. "Vestinghouse Small Break ECCS Evaluation Model Generic Study Vith the NOTRUMP Code," VCAP-11145, Rupprecht, S. D., et al., August 1985.

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ATTACIBfENT B  ;

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-EFFECT OF SAFETT EVALUATIONS AGAINST THE ECCS (LOCA) ANALYSIS RESULTS POUND IN CHAPTER 15, SECTIONS 3 AND 4 0F THE - l FARLET UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT I t 1ARGB EREAK 1hCA

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DESCRIPTION OF PLANT MODIFICATIONS ,

The large break Loss +of-Coolant Accident (LOCA) analysis results have been i supplemented by safety evaluations of plant design changes under 10CFR50.59 j that have assessed penalties to the fuel peak cladding temperature (PCT).

Specifically, a safety evaluation of the effect of loose parts in the RCS [

(unrecovered grid strap sections) was performed to determine the effect of '

this condition on the large break LOCA analysis PCT results. The evaluation *

!: determined that the large break LOCA analysis PCT results could be affected  :

by a 60'F increase.

RESULTANT LARGE BREAK LOCA PCT

'As discussed above,. plant modifications could affect the resultant PCT as follows: ,

A'. Resultant PCT from ECCS Evaluation Model Modifications Reported in Attachment A 2027'F B. 10CFR50.59 Safety Evaluation for Loose Parts (Grids) + 60'F Total Large Break Resultant PCT 2087'F l

SMALL BREAK LOCA -i l

DESCRIPTION OF PLANT MODIFICATIONS The small break LOCA analysis results have been supplemented by safety evaluations of plant design changes under 10CFR50.59 which have assessed  ;

penalties to the PCT as follows: ,

1) A safety evaluation of the effects of a plant design change for upflov 1 conversion (Unit 1 only) was performed. As documented in Section L 15.3.1.2.2 of the Farley Units 1 and 2 Final Safety Analysis Report i (FSAR), the evaluation of the effect of this plant design change on the l small break LOCA analysis PCT results was calculated. The study L determined that the LOCA analysis PCT results could be affected by a ll?'F increase.

4 2) A safety evaluation of the effect of loose parts in the RCS (unrecovered grid strap sections) was performed. An evaluation of the effect of this-condition on the small break LOCA analysis PCT results was performed.

The evaluation determined that the small break LOCA analysis PCT results could be affected by a 32'F increase.

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, l RESULTANT SMALL BREAK LOCA PCT As discussed above, plant modifications could affect the resultant PCT as '

follows: l 1

A. Resultant PCT from ECCS Evaluation Model Modifications Reported in Attachment A 1756'F. .

B. 10CFR50.59 Safety Evaluation for Upflow Conversion ->

(Unit 1 only) -+ 117'F  !

C. 10CFR$0.59 Safety Evaluation for Loose Parts (Grids) + 32*F Total Resultant PCT 1905'F i

CONCLUSIONS An evaluation of the effect of modifications to the Vestinghouse ECCS Evaluation Model as reported in References 2 and 3 and discussed in -

Attachment A vas performed for both the large break LOCA and small break  !

LOCA analysis results found in Chapter 15, Sections 3 and 4 of the Parley l Units 1 and 2 FSAR. It was determined that compliance with the ,

requirements of 10CFR50.46 vould be maintained when plant design changes, i performed under 10CFR50.59, which could affect the LOCA analysis results were combined with the effect of the ECCS Evaluation Model modifications applicable to Farley Units 1 and 2.

JAR:msd 15.13 i

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