ML20076N053
| ML20076N053 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/21/1991 |
| From: | Woodard J ALABAMA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9103270246 | |
| Download: ML20076N053 (14) | |
Text
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. A o, woooue . March 21, 1991 AlabamaPower Vee Prew3ent-Nuclear fo ney Protect Pe scutwo thette tywm:
10 CFR 50.46 Docket Nos. 50-348' 50-364 U. S. Nuclear Regulatory Comnission
- ATIN: Document Control Desk ,
Washington,,D.-C. 20555- [
t Gentlemen::
Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1990 he October 17,'.1988, revision to.10 CFR 50.46 requires applicants and holders cf operating licenses or. construction permits to annually notify the Nuclear Regulatory Comnission (NRC) of insignificant errors and changes in the ECCS Evaluation Models. Enclosed is Alabama Power Company's report for calendar ' year 1990 in compliance with this requirement for Joseph M.
Farley Nuclear Plant Units 1 and 2.
. Attachment A provides information regarding the effect of the ECCS Evaluation Model modifications on the peak cladding temperature (PCT)-
results.< - Attachment B provides a sumary of the plant change safety
- evaluations performed through December: 31, 1990, that impact PCT under the provisions of 10 CFR 50.59. Please note that the facility change safety evaluations included in Attachment B reflect only those which result in non-zero PCT penalty assessments. This information package constitutes Alabama.?ower Company's report for 1990 to the NRC as part of annual reporting required'by 10 CFR 50.46(a)(3)(ii).
It. bas been determined that compliance with the requirements of 10 CFR 50'46-continues to be maintained when the effects of plant design-
-changes performed under 10 CFR 50.59 are combined with the effects of the-ECCS Evaluation Model-modifications applicable to Farley. Units 1 and 2.
'This determination is based on the fact that the total large-break and small-break resultant PCTs reported in Attachment B (i.e., including ECCS Evaluation Model modifications and all non-zero PCT penalties associated with the plant change safety evaluations performed under 10 CFR 50.59) are
'below the PCT limit of 2200 F.
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9103270246 910321 ,
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e U'.S. Nuclear Regulatory Commission Page 2-Alabama Power Cmpany recently became aware of an application error in the calculation of PCT for Farley Nuclear Plant Unit 2. The error was associated with the downflow configuration in Unit 2 when using the 1981 Evaluation Model with IESH, which resulted in a 90'F increase in the large-break IOCA PCT for Unit 2. (The earlier sensitivity studies using the 1978 Evaluation Model had indicated the the Unit 1 upflow configuration was conservative for Unit 2.) ' Itis error was significant and thus reportable within 30 days under the provisions of the 10 CFR 50.46 (a)(3).
A separate report was recently submitted by Alabama Power Cortpany on February 18, 1991, which provided a detailed discussion of the error and the correction to the large-break IDCA PCT for Unit 2. It should be noted that the Unit 2 results reported in the enclosed annual report do not address this error since the annual report covers 1990 calendar year only.
If there are any questions, please advise.
Respectfully submitted, ALA!W % P N ER COMPANY nk
. Woodard J W/ REM:maf0811 Attachments cc: Mr. 3. D. Ebneter Mr. S..T. Hoffman Mr. G. F. Maxwell
.- ATrACIMENT A EFFECT OF WESTINGiKUSE ECCS INALIATION MODEL MODIFICATIONS ON ' HIE LOCA ANALYSIS RESULTS BACFGOUND
'Ihe October 17, 1988, revision to 10 CFR 50.46 requires applicants and holders of operating licenses or construction permits to annually notify the Nuclear Regulatory Com,nission (NRC) of insignificant errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models. Reference 1 defines a significant error or change as one which results in a calculated fuel peak cladding temperature (PCT) dif ferent by more than 50 F from the 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 absolute magnitudes of the respective temperature changes is greater than 50 F.
In References 2 and 3, information regarding modifications to the Westinghouse large-break and small-break Loss-of-Coolant Accident (IDCA)
ECCS Evaluation Models was submitted to the NRC. It should be noted that the information on large-break IDCA provided in Reference 2 is applicable to rarley Unit 1 only, since the Earley Unit 2 large-break IDCA was reanalyzed separately in support of a recently approved license amendment (Reference 4). The license amendment provided changes to the Unit 2 Technical Specifications to allow an average of 15% steam generator tube plugging (SGTP) with a peak of 20% in any one steam generator (from 10%
uniform which was in effect for Unit 1 as of December 31, 1990). The amendment also included an approximate 1.5% reduction in the reactor coolant system thermal desicrn flow to 261,600 gpm (from 265,500 gpm).
The following presents an basessment ot the effect of the modifications to the Westinghouse ECCS Evaluation Models on the IOCA analysis results. As stated above, the modifications to the Westinghouse ECCS Evaluation Models on the large-break LOCA analysis apply to Unit 1 only, because the current licensing basis analysis for Unit 2 uses a modified version of the methods used for Unit 1.
- ARGFeBREAK IDCA ECCS EVALUAT7CN MODEL The large-breah LOCA analyses for rarley Units 1 and 2 were examined to assess the effect of the applicable modifications to the Westinghouse large-break LOCA ECCS Evaluation Model on PCT results. The large-break LOCA analyses results for Unit 1 were calculated using the 1981 version of the Westinghouse large-break IDCA ECCS Evaluation Model incorporating the BASH analysis technology. For Unit 2, a modified version of the 1981 Evaluation l
kTTACHMIWrA Page 2-Model with BASH was used in support of the license amendment discussed above. The Unit 1 and Unit 2 analyses assumed the following information important to the large-break LOCA analyses:
Unit 1 Unit 2 Core Power = 1.02 x 2652 tWr Core Power = 1.02 x 2652 twr 17x17 Standard ruel Assembly 17x17 Standard ruel Assembly rg = 2.40 r, = 2.32 T-delta-H = 1.62 T-delta-H = 1.62 SGTP* = 10% Uniform SGTP* = 15% Average /20% Peak Upflow configuration Assumed Upflow configuration Assumed
- SGTP = Steam Generator Tube Plugging Limit For rarley Unit 1, the limiting break resulted from the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of C, = 0.4. The calculated peak cladding temperature was 2013*F.
For Farley Unit 2, the limiting break size previously established was reanalyzed to support the recently approved changes in the Technical Specifications. The analysis-of-record peak clad temperature was 2049* r as obtained by the modified version of the 1981 Evaluation Model with BASH.
The analysis-of-record peak clad temperature for Unit 2 also included the combined effects of previous evaluations in order to form a new design basis for Unit 2.
The modifications to the Westinghouse ECCS Evaluation Modols discussed in Reference 2 which could affect the large-break LOCA analysis results for Unit 1 are described below.
MODIFICATIONS % MjE BASH ECCS EVALUATION MODEL (Farley Unit 1 Only)
Several improvements were made to the BASH computer code t.o treat special analysis cases which are related to the cracking of fluid interfaces and which could affect the plant analysis results.
- 1) A modification to prevent the code f rom 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.
hITACI1 MENT A Page 3
- 2) A modification to prevent the code from aborting was made to allow negative initial movement of the liquid /two-phase and liquid-vapor interfaces. The coding in these areas was generalized to prevent mass imbalance in the special case where the liquid /two-phase interface reaches the bottom of the BASit core channel.
- 3) Modifications to orevent the code from aborting were made to increase the dime.a. ions of certain arrays for special applications.
- 4) A modification was made to write additional variables to the tape of information to be provided to LOCaART.
- 5) Typograp! . cal errors in the coding of some convective heat transfer terms were corrected, but the corrections have no effect on the BASH analysis results since the related terms are always set equal to zero.
- 6) A modification was made to the BASH coding to reset the cold lery conditions in a conservative manner when the accumulators empty.
The BAS!! model is initialized at the bottom of core recovery with the intact cold legs and lower plenum full of liquid. Flow into the downcomer then equals the accumulator flow. The modification removed most of the intact cold leg water at the accumulator empty time by resetting the intact cold leg conditions to a high quality two-phase mixture.
In a typical BAsti calculation, the downcomer is nearly full when the accumulators empty. The delay time, prior to the intact cold leg water reaching saturation, is sufficient to allow the downcomer to fill f rom the addition of safety injection fluid before the water in the cold legs reaches saturation. When tha intact cold leg water reaches saturation, it merely flows out vi the break.
The cold leg water, therefore, does not affect the reflood transient.
However, in a special case where a substantial time was required to fill the downcomer af ter the accunulators 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.
The Parley Unit 1 WCA 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 rarley Unit 1 discussed above, a conservative estimate of 10 F has been assessed and tracked for use in determining the available margin to the limits of 10 Crn 50.46.
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t kTrhCHMFRr A Page 4 MODIFICATIONS 'IO 'IEE WREFIDDD COMPlTTER CODE (Farley Unit 1 Only)
In Reference 2, nodifications are reported for the 1981 ECCS Evaluation Model which form the fundamental framework for application of the BASH methodology. The modifications made to th? WREFLOOD computer code described for the Westinghouse 1981 ECCS Evaluation Model were carried into the WREFLOOD computer code used for BASH analyses.
In the BASH methodology, the WREFLOOD ccde is only used to calculate the bottom of core recovery time. 'Iherefore, this modification has no ef fect on the BASH ECCS Ev;1uation Model calculations.
MODIFICATION 'IO THE LOCBART COMPUTER CODE No modifications 4e beer, made since those outlined in Reference 5.
CONTAINMENT PURGE LINES OPEN EVALUATION 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 IOCA analysis PCT results was projected. The evaluation determined that the large-break LOCA analysis PCT results could be affected by a 4'F increase. 'Ihis effect was included in the Unit 2 reanalysis.
RESULTANT LARGE-BREAK LOCA PCT As discussed above, nodifications to the Westinghouse large-break LOCA ECCS Evaluation Model could affect the large-break LOCA analysis results by altering the PCT as shown below:
Unit 1 Unit 2 A. Analysis Calculated Result (Analysis-of-Record) -2013 F 2049 F B. Modifications to Westinghouse ECCS Evaluation Model + TD* r + N/A
- Containment Purge Lines Open Evaluation + 4F ~-
- C.
D. ECCS Evaluation Model Modifications Resultant PCT 7D77 F 70T9' F Therefore, the sum of the absolute magnitude of the PCT assessments introduced as a result of the modifications and errors in the large-break LOCA ECCS Evaluation Model are 14 F for Unit 1 and 0 F for Unit 2.
- The Unit 2 limiting large-break LOCA reanalysis was performed 15i, by Westinghouse to support the recently approved SGTP limit of average /20% peak. The reanalysis used the latest version of the 1981 Evaluation Model with BASH; thus, the previous 1981 Evaluation Model's (also with BASH) penalty is not applicable. Also included in the analysis-of-record for Unit 2 was the combined effects of previous safety evaluations.
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, A'TDONENT A--
Page 5 -
.-i SMALIr-BREAK toCA _
ECCS EVALUATION MODEL
- The small-break I4CA analyses _for Farley Units:1 and 2 were also examined-to assess the. ef fect of the applicable modifications - to the Westinghouse ECCS Evaluation Models on PCT results reported in Chapter 15,.Section 3 of- :t the FSAR. .The small-break IOCA analyses results:were calculated using-the L 1974 small-break IOCA ECCS Evaluation . Model _ incorporating - the. WFLASH analysis : technology. For Farley Units 1 and: 2 ,' the. limiting . size
.small-break resulted from'a six-inch equivalent diameter break in-the: cold v leg.- he calculated PCT was '.1712* F. The ' analysis assumed - the following :
informationLimportant to theLsmall-break LOCA analysect {
o- Core Power = 1.02'X 2652 twr Lo 17x17 Standard Fuel Assembly o r, a 2.32'
.t o- F-delta-H --1.55 a o Auxiliary Feedwater Flow =>1050 gpm (Total):
The; modifications to1the Westinghouse ECCS Evaluation 'Models. discussed in. ,
-References 2 and 3 which could affect the'small-break- toCA analysis results
- found1 1n. - Chapte rel5, Section 3 in; the : Farley: Units l __ and 2 . FSAR ~ are -
described below.
WFLASH ECCS EVALUATION MODEL CODE Following .the accident; at Three Mile; Island Unit: 2, : additional attention
- was focused ~on the small-break LOCA, and Westinghouse submitted a -report,i WCAP-9600--(
Reference:
6), to the Nuclear iRegulatory. Commission (NRC): J
- detailing the performance of theLWestinghouse small-break LOCA Evaluation
- Model?which utilized - the WFLASH computer code. -In NUREG-0611~(Reference .l 17).the: NRC' staff questioned the validity of ,certain models .in the WFLASH ~
-. computer code =and requiredclicensees to justify continued acceptance of the--
- model. LSection II.K.3.30 -of NUREG-0737- (Reference; 8) clarified: the :NRC
- post-TMI 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 fiss' lance 1of NUREG-0737~,- Westinghouse - and - the Westinghouse <
Owners Group decided to-develop the NOTRUMP -(Reference 9) computer ' code .for.
use in a- new small-breaki LOCA ECCS Eval uati on .Model (Reference 10). The NRC approved. the use of NOTRUMP for L small-break LOCA ECCS_ analyses in May 1985.
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. 'A'PTACHMWT A Page 6-- .
Since approval of the NCTTRUMP small-break LOCA ECCS Evaluation Model -in
- 1985, the_' WrIASH - computer code has not been maintained as part of the Westinghouse FCCS Evaltation Model computer code.
In . section II .K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small-break IDCA 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 11) relaxed the ,
requirements of = item II.K.3.31 by -allowing a more generic response and 4 providing--a basis for retention of the existing small-break toCA 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 - (NC7 TRUMP), plant-specific analyses using the new small-break 1/X'A Evaluation Model -would not be required. In WCAP-lll45 (Reference 12), Westinghouse and the Westinghouse Owners Group demonstrated that the-resultc obtained from calculations with .
WrtASH were conservative relative to those obtained with NOTRUMP..
- Compliance with item II.K.3.31 of NUREG-0737 for rarley has been conpleted by referencing WCAP-lll45 and determined to be acceptable by the NRC.
. Westinghouse, therefore, has not been modifying, investigating, or evaluating proposed changes _to the WFLASH small-break LOCA ECCS Evaluation Model.- Thus,:there are no modifications to report.
- SBIDCTA--IV COMPUTER CODE
'ihe following modifications to. the LOCTA-IV computer code in the
-small-break IOCA 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 wall-to-steam temperature differential dropped below the useful range .of the correlation. This limit was derived based upon the physical limitations of the radiation phenomena.
There is no effect of the modification on reported PCTs since the .
erroneous use= of the correlation -forced the calculations into aborted conditions.
2)- An update was performed to allow the use. of fuel rod performance data from the revised Westinghouse (PAD 3.3) model.
An evaluation indicated that there is an insignificant effect of the modification on reported PCTs.
)
i'PrACIMENT A Page 7
- 3) Modifications supporting a general upgrade of the computer program were implemented as follows:
a) removal of unused or redundant 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.
Verification analyses calculations demonstrated that there was no effect on the calculated output resulting from these changes.
- 4) Three modifications improving the consistency between the Westinghouse fuel rod performance data (PAD) and the small-break LOCTA-IV fuel rod models were implemented:
a) The form of the equation for the density of Uranium-Dioxide was corrected to calculate thermal expansion only in two dimensions, wnich is consistent with the way in which the fuel rod is modeled in the LOCTA codas, b) The correlation for the specific heat of water vapor at temperatures over 1590* F was improved, c) An error in the equation for the pellet / clad contact pressute was corrected.
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.
SAFE 1Y INJECTION ! SACK PRESSURE FIX EVALUATION A safety evaluation of the effect of spilling broken loop safety injection water to containment back pressure instead of to reactor coolant system back pressure was performed for both units. 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'F increase has been previously reported in an Alabama Power Company to NRC letter dated January 14, 1988.
.- 'ATPMJWENT A' Page 8.- m i
-RESULTANT SMALL-BREAK LOCA PCT As discussed above', modifications to the Westinghouse small-break LOCA ECCS Evaluation Model could affect the small-break IOCA analysis results by altering the PCT as shown below, Unit 1 Unit 2 A. Analysis Calculated Result (Analysis-of-Record) 1712*F 1712*F i
-B. Modifications to Westinghouse ECCS Evaluation Model - 7 'r - 7 'r C. Safety Injection Back Pressure Fix Evaluation + 7 *F +7* F-D. - ECCS Evaluation Model Modifications Resultant PCT T755' r T756* r -
Therefore, the sum of the absolute magnitude of the PCT assessments
-introduced es the result of the modifications and errors in the small-break.
LOCA ECCS Evaluation Model is 48 F for Units 1 and 2.
CONCLUSIONS
- An evaluation- of the effect of modifications to the Westinghouse ECCS Evaluation Model, as reported in References 2'and 3, was performed for both
.the -large-break - LOCA and small-break LOCA . analysis - results. _W hen the effects of the ECCS model changes were combined with the current plant analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 would be mainiained.
REFERENCFE
" Emergency Core Cooling Systems: Revisions to Acceptance - Criteria,"
~
, 1.
Federal Register, Vol. 53, No.180, pp. 35996-36005, dated September
-16, 1988.
V 2. NS -NRC-89-346 3,- "10 CFR. 50,46 Annual -Notification for 1989 'of Modifications in the Westinghouse ECCS Evaluation Model," Letter f rom _--
W. J. Johnson (Westinghouse) to T. E. Murley .(NRC), dated October 5,
-1989.
- 3. NS-NRC-89-3464, " Correction of Errors and Modifications to- the N0frRUMP Code in the Westinghouse' Small- Break LOCA ECCS Evaluation Model Which Are Potentially Significant,"' Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC), dated October 5, 1989.
- 4. Docket No. 50-364, " Issuance of Amendmant No. 79 to Facility Operating License No. NPF-8 Regarding' Steam Generator Tube Plugging - Joseph M.
Farley Nuclear Plant, Unit 2-(TAC No. 77236)," NRC letter from S. T. Hoffman to W. G. Hairston, III, December 6, 1990.
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. 'ATTACilMENT A Page 9
- 5. WCAP-10266-P-A, Revision 2 (Ptoprietary), WCAP-10267-A, Revision 2 (Non-Proprietary), Bessplata, J. J., et al., "1981 version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.
- 6. " Report on Small Break Accidents for Westhghouse Nuclear Steam Supply System," WCAP-9601 (Non-Proprieta'y' c June 1979, WCAP-9600 (Proprietary), June 1979.
- 7. " Generic Evaluation of Feedvater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants,"
NUREG-0611, January 1980.
- 8. " Clarification of 'IMI Action Plan Requirements," NUREG-0737, November 1980.
- 9. "NOTRUMP - A Nodal Transient Small Break and General Network Code,"
WCAP-10079-P-A (Proprietary), WCAP-10080-A (Non-Proprietary),
Meyer, P. E., et al., August 1985.
- 10. " Westinghouse Small Break ECCS Evaluation Model Using the terRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary),
Lee, N., et al, August 1985.
- 11. " Clarification of 'IMI Action Plan Item II.K.3.31," NRC Generic Letter 83-35 from D. G. Eisenhut, November 2, 1983.
- 12. " Westinghouse Small Break ECCS Evaluation Model Generic Study With the NOTRUMP Code," WCAP-11145, Rupprecht, S. D., et al., August 1985.
ATIACIDtENT B EFFECT OF SAFETY EVALUATIO4S AGAINST 'IllE LOCA ANALYSIS RESUL':'S IARGE-BREAK LOCA DESCRIPTION OF PLANT MODIFICATIONS As discussed below, the large-break Loss-of-Coolant Accident (LOCA) analysis results have been supplemented by safety evaluations of plant design changes under 10 CFR 50.59 that have assessed penalties to the fuel peak cladding temperature (PCT).
- 1) A safety evaluation of the erfect of loose parts in the Unit 1 RCS <
(unrecovered grid strap see'. ions) was performed to determine the effect of this condition on the. large-break LCCA analysis PCT results.
Completed in 1988, the evaluation determined that the Unit I large-break LOCA analysis _ PCT results could be af fected by a 60* F increase.
- 2) A safety evaluation of the effect of the temperature uncertainties on the large-break LOCA analysis was performed as part of a proposed Technical Specifications change to remove the R'ID bypass loops for Unit 1. 'Ihe temperature uncertainties are associated with the accuracy of the instrumentations, the accuracy of the calibration equipment, and the procedures used for calibrating and reading the instrumentations.
Completed in 1990, the evaluation determined that the Unit i large-break LOCA analysis PCT results could be affected by a 3F increase.
RESULTANT LARGE-BREAK LOCA PCT As discussed above, plant modifications could affect the resultant PCT as follows:
Unit 1 Unit 2
- 1. 10 CFR 50.59 Safety Evaluation for Loose Parts (Grids) +' 60*F + N/A
- 2. 10 CFR 50.59 Safety Evaluation for RCS Temperature Uncertainties + 3*F + * ,
Total Large-Break Resultant PCT 2090* F 2049'F l
- In the case of Unit 2, the large-break LOCA analysis-of-record which was performed in support of a recently approved SGTP limit of 15%
average /20% peak accounted for the offect of RCS temperature uncertainties.
l 1
l
. - .- - - - = .- -
[ ',..
- 'ATmOpeNP B = l Page.2t Therefore, the total PCT assessments ' introduced as a result of plant
- modifications are 63* r for tnit 1 and _0'r for Unit 2.
SMAL1r-BREAK LOCA I
DESCRIPTION OF PLANT MODIFICATIONS As described below, -the small-break U0CA analysis results have been supplemented by safety evaluations of plant design changes under 10 CPR 50.59 which have assessed penalties to the PCT.
' 1) .A' safety evaluation of the effects of a plant design change for upflow
-conversion (Unit 1 only) was performed. As documented in section 15.3.1.2.2 of the Parley Units 1 and 2 Final Safety Analysis Report (FSAR), the evaluation of the ef fect of this plant (k. sign change on the small-break LOCA analysis PCT results was calculated. The study determined that the Unit 1 small-break LOCA analysis PCT results'could be - af fected by a: 117 r increase. This evaluation was completed . in 1982.
- 2)- A safety evaluation of the effeet of loose parts in the Unit - 1 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 Unit 1 small-break LOCA analysis PCT results could be af fected by a 32* r increase. This evaluation was completed in 1988.
- 3) A safety evaluction of the effect of the temperature uncertainties on the small-break LOCA was performed as part of a . proposed - Technical Specificatione change to remove the RTD bypass loops for Unit 1. 'Ihe temperature' uncertainties are associated with the accuracy of the
. instrumentations, the accuracy of the calibration equipment, and the procedures used for calibrating and reading the instrumentations. The evaluation determined that the small-break LOCA analysis PCT results could ba affected by a 2 r increase. -
Since the instrumentations and ,
procedures are common between the' two units, the same penalty applies to Unit 2 also. This evaluation was completed in 1990.
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.Page--3 RESULTANT SMALL-BREAK LOCA PCT As discussed above, plant modifications could affect the resultant-PCT-as follows:
Unit 1 Unit 2 >
- 0. Resultant PCT from ECCS Evaluation Model Changes / Errors Reported in Attachment A -1756*r 1756*r
- 1. 10 CPR 50.59. Safety Evaluation for Upflow conversion + II7* r N/A
- 2. 10 CPR 50.59 Safety Evaluation for Loose Parts (Grids) +7' r N/A -
3, 10 CFr 30.59 Safety Evaluation for RTD Temperature
-Uncertainty + 2'r + 2* r Total Resultant PCT 1907*r 1758'r Therefore, the . resultant PCT assessments introduced as a results of plant.nodifications is-151'r for Unit 1 and 2*r for Unit 2.
CONCLUSIONS .
An evaluation of'the effect'of modifications to the Westinghousr E.c Evaluation Model as reported in Attachment- A's References 2, 3 and 4 was.
performed for both the large-break LOCA and small-break' LOCA licenW~
asis analysis results. "It was determined that compliance with the requirements of 10 CFR 50.46 would be maintained when plant design changes,
_p erformed under 10.CFR 50.59, which could affect the LOCA analysis results were-combined with the effeet of the ECCS Evaluation Mode 1 modlfications applicable to Farley Units 1 and 2.
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