ML20141M109

Forwards 10CFR50.46 Annual ECCS Evaluation Model Changes Rept for 1991,including Effect of Westinghouse ECCS Evaluation Model Mods on LOCA Analysis Results & Effect of Safety Evaluations Against LOCA Analysis Results

ML20141M109
Person / Time
Site:	Farley
Issue date:	03/25/1992
From:	Woodard J SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9204010136
Download: ML20141M109 (21)

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,,. o. % ,,,,, Southern Nudear Operating Company we snes,xm i a'k'y I'v.1 tne wunm e:ecn wtem 10 CFR 50.46 Docket Hos. 50-348 50-364 U. 5. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Joseph li. Farley Nuclear Plant 19 CFR 50s46 Annual ECCS Evaluation Model Chanaes Raport for 1991 Gentlemen:

The October 17, 1988, revision tc 10 CFR 50.46 required applicants and holders of operating licenses or construction permits to ennually notify the Nuclear Regulatory Commission (NRC) of insignificant errors and changes 3 in the Emergency Core Cooling System ECCS) Evaluation Models. Enclosed is Southern Nuclear Operating Company's r(eport for calendar year 1991 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 (PCI) results. Attachment B 3rovides a summary of the plant change safety evaluations performed tirough December 31, 1991, that impact PCT under the provisions of 10 CFR 50.59. Please note that the facility change safety evaluations included ii. Attachment B reflect only those which result in non-zero PCT penalty assessments. This information package constitutes Southern Nuclear Operating Company's report for 1991 to the NRC as part of annual reporting required by 10 CFR 50.46(a)(3)(ii).

It has 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 Mcdel 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 22000F. '

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l U. S. Nuclear Regulatory Commission Page 2 By letter dated March 11, 1992, the NRC approved the farley VANTAGE,5 fuel reanalysis. This reanalysis results in a reduced PCT for both inits and will be reported in the 10 CFR 50.46 annual report for 1992.

If there are any questions, please advise.

Respectfully submitted, 9 SOUTHERN NUCLEAR OPERATING COMPANY NS .) '

k 4.- . Woodard JDW/ REM / RAH: map 2022 Attachment cc: Mr S. D. Ebneter Mr. S. T. Hoffman Mr. G. F. Maxwell k

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4 ATTACHMENT A EFFECT OF WESTINGHOUSE ECCS EVALUATION MODEL

, MODIFICATIONS ON THE LOCA ANALYSIS RESULTS i BACKGROUND ,

The October 17, 1988, revision to 10 CFR 50.46 required applicants and holders of operating licenses or construction permits to annua'aly notify the Nuclear Regulatory Commission-(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) different by more than 500F 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 500F.

In Reference 2, information-was submitted to the NRC regarding o difications to 4

the Westinghouse large-break and small-break Loss-of-Coolant Accident (LOCA)

ECCS Evaluation Models as applicable to the Farley Nuclear Plant (FNP) analyses for the calendar year 1990.- It should be noted that the large-break LOCA '

analysis-of-record PCT values for 1991 are different from those reported in the ,

1990 annual report (Reference 2). This is because Unit I has been re-analyzed for a steam generator tube plugging level of 15% average /20% peak and because .

Unit 2 has been re-analyzed to explicitly account for a, downflow barrel baffle '

configuration.

L The following presents an assessment of the effect of modifications to _the Westinghouse ECCS Evaluation Models on the.Farley Nuclear Plant- LOCA analysis results for the calendar year 1991.

LARGE-BREAK LOCA I. ECCS EVALVATION MODEL The large-break LOCA analyses for farley 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.

LOCH analyses results for Units 1 and 2 were calculated using the 1981 version of the - Westinghouse large-break. LOCA .ECCS Evaluation -Model incorporating the BASH analysis-technology. . The Unit I analysis was performed -in support =of a license amendment to increase the Steam Generator Tube Plugging (SGTP) level to 15% average /20% peak l(Reference

3) .- The Unit .2 analysis was performed to explicitly account for the -

downflow configuration of the unit,- also at a SGTP level offl5%

average /20% peak, as reported in Reference -4. The Unit 1 and Unit-2

-analyses _-assumed the following- information important to the large-break LOCA analyses.

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ATTACHMENT A Page 2 Unit 1 Unit 2 Core Power - 1.02 X 2652 MWT Core Power 1.02 X 2652 MWT 17x17 Standard fuel Asscmbly 17x17 Standard fuel Assembly FQ - 2.32 FQ - 2.32 F-delta-H - 1.55 F-delta-H 1.55**

SGTP* - 15% Average /20% Peak SGTP* - 15% Average /20% Peak Upflow Configuration Assumed Downflow Configuration Assumed G GTP - Steam Generator Tube Plugging Limit

• • Note that the Unit 2 F-delta-H was reported as 1.62 in Reference 2. The correct value was 1.55.

For farley Unit 1, the limiting break size resulting from the double-ended guillotine rupture of the cold leg piping with i dis:harge

, coefficient of CD - 0.4, as previously estabitshed for 10% SGTP (Reference 5), was reanalved to support the 15% average /20% peak license amendment.

The calculated peak cladding temperature was 20490F. The value of 20690F previously reported in Reference 3 (which, when added to 40F to account for containment mini-purge operation, becomes 20730F) conservatively contained an additional 240F to account for a potential issue which was subsequently resolved.

For farley Unit 2, the limiting break size previously established was reanalyzed to reflect the actual downflow configuration of Unit 2. The analysis-of-record peak clad temperature was 21380F. It should be noted that the value of 21630F reported for the Unit 2 -large-break LOCA in Reference 4 conservatively contained an additional 250F due to a potential issue which has been subsequently resolved.

The analysis-of-record peak clad temperature for both units also included the combined effects of previous evaluations in order to form a new design basis.

The modifications to the Westinghouse ECCS Evaluation Models which could affect the _large-break LOCA analysis results for Units 1 and 2 are described below.

II. ELEAM GENERATOR Ft0W AREA REDUCTION Licensees are normally required to provide assurance that there exists only an extremely low probability of abnorm:1 leakage or gross rupture of any part of the reactor coolant pressure bouadary (General Design Criteria 14 and 31). The NRC issued a Regulatory Guide (RG 1.121) which addressed this requirement specifically for steam generator tubes in pressurized water reactors, in that guide, the staff required analytical and

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ATTACl!M NT A Page 3 experimental evidence that steam generator tube integrity will be maintained for the combinations of the loads resulting from a LOCA with the loads from a safe shutdown earthquake (SSE). These loads are combined for added conservatism in the calculation of structural integrity. This analysis would provide the basis for establishing criteria for removing tubes from service which had experienced significant degradation.

Analyses performed by Westinghouse in support of the above requirement for various utilities combined the most severe LOCA loads with the plant-specific SSE, as delineated in the design criteria and the Regulatory Guide. Generally, these analyses showed that, while tube integrity was maintained, the combined loads led to some tube defcrmation.

This deformation reduces the flow area through the steam generator. The reduced flow area increases the resistance through the steam generator to the flow of steam from the core during a LOCA which potentially could increase the calculated PCT.

The effect of tube deformation and flow area reduction in the steam generator was analyzed and evaluated for some plants by Westinghouse in the late 1970's and early 1980's. The combination of LOCA and SSE loads led to the following calculated phenomena:

1. LOCA and SSE loads cause the steam generator tube bundle to vibrate.

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2. The tube support plates may be deformed as a result of lateral loads at the wedge supports at the periphery of the plate. The ,

tube support plate deformation may cause tube deformation.

3. During a postulated large LOCA, the primary side depressurizes to containment pressure. Applying the resulting pressure di f ferential to the deformed tubes causes some of tliese tubes to collapsc and-reduces the effortive flow area through the steam generator.

4. The reduced flow area increases the resistance to venting of steam generated in the core during the reflood phase of the LOCA, increasing the calculated PCT.

The ability of the steam generator to continue to perform its safety function was established by evaluating the effect of the_ resulting flow area reduction on the LOCA PCT. The postulated break examined was the j steam . generator outlet break since this break m s judged to result in'the greatest loads on the steam generator and thus the' greatest flow area

, redaction. It was concluded that the steam generator would continne to l

me(t its safety function since the degree of flow area reduction-was- small and the postulated break at the steam generator outlet resulted in an acceptable PCT.

ATTACHMENT A Page 4 In April of 1090, in considering the offect of the combination of LOCA +

SSE loadings on the steam generator component, it was determined that the potential for flow area reduction due to the contribution of SSE loadings should be included in other LOCA anal /ses. With SSE loadings, flow area reduction may occur in all steam generators (not just the faulted loop).

Therefore, it was concluded that the effects of flow area reduction during the most limiting )rimary pipe break affecting LOJA PCT (i.e., the reactor vessel inlet brea< (cold leg break LOCA)) had to be evaluated to confirm that 10 CFR 50.46 limits continue to be met, and that the affected steam generators will continue to perform their intended safety function.

Consequently, action was taken to address the safety significance of steam generator tube collapse during a cold leg break LOCA. The effect of flow area reduction from combined LOCA and SSE loads was estimated. The magnitude of the flow area reduction was :onsidered equivalent to an increased level of steam generator tube plugging. Typical'., the area reduction was estimated to range from 0 to 7.5%, depen;.nr in the magnitude of the seismic loads. Since detailed non-lino lismic analyses are not available for Series 51 and earlier dam steam generators, some area reductions had to be estimated based on awilabic in format ion.

A bounding flow area reduction of five percent was established for the farley Model 51 steam generators. This flow area reduction, if included in the LOCA analysis, has been conservatively estimated to result in a 500f increase in peak cladding temperature. This 500f increase in PCT has been assessed and tracked for use in determining the available margin to the limits of 10 CFR 50.46.

l in addition, information has been submitted to the NRC which demonstrates j an allowable reduction of this penalty from 500f to 60f (Reference 6) l when proper credit is taken for farley-specific SSE and LOCA loads.

Furthermore, - a plant specific analysis based on recent Westinghouse crush tests for Model 51 steam generator support plates determined that there would be no tubes that experienced deformation greater than 0.025 inch as a result of combined LOPA + . SSE loading. As a result, significant tube leakage is not expected. Therefore, based on the latest analysis, the penalty could be reduced from 60f to zero. The latest analysis is documented in-Section 11 of Reference 16 which is currently undergoing review by the NRC Staff.

I 111. FVfL R0D INITIAL CONDITION INCONSISTENCY REVISIONS During the review of the original Westinghouse ECCS Evaluation Model following the promulgation of 10 CFR 50.46 in 1974, Westinghouse committed to maintain consistency between future LOCA fuel rod computer models and the fuel rod design computer nodels used to predict fuel rod normal operational performance. Thess fuel rod design codes are also used to I

establish initial conditions for the LOCA analysis.

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ATTACHMENT A Page 5 Chanae Descriotion it was fcund that the large-break LOCA codo versions were not consistent with fuel design codes in the following areas:

). The LOCA codes were not consistent with the fuel rod design code relhtive to the flux depression factors at higher fuel enrichment.

2. The LOCA codes were not consistent with the fuel rod design code relative to the fuel rod gap gas conductivities and pellet surface roughness models.

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3. The coding of the pellet / clad contact resistance model required revision.

Modifications were made to the fuel rod models used in the LOCA Evaluation Models to maintain consistency with the latest approved version of the fuel rod design code.

In addition, it was determined that integration of the cladding strain rate equation used in the large-break LOCA Evaluation Model, as described in Reference 15, was being calculated twice each time step instead of once. The coding was corrected to properly integrate the strain rate equation.

Effect of Chanaes The changes made to make the LOCA fuel rod models consistent with the fuel design codes were judged to be insignificant, as defined by 10 CFR 50.46 (a)(3)(i), To quantify the effect on the calculated PCT, calculations were performed which incorporated the changes, including the cladding strain model correction for the large-break LOCA. For the large-break LOCA Evaluation Model, additional calculations incorporating only the cladding strain corrections were performed, and the results supported the conclusion that compensating effects were not present. The PCT effects reported below will bound the effects taken separately for the large-break LOCA.

The effect of the changes on the large-break LOCA PCT was determined using the BASH large-break LOCA Evaluation Model. The ' effects were judged applicable to older Evaluation Models._ Several calculations were performed to assess the effect of the changes on the calculated results as follows:

1. Blowdown Analysis  ;

It was determined that the changes will have a small effect on the  !

core average rod and hot assembly average rod performance during the blowdown. analysis. The effect of the changes on the blowdown analysis was determined by perforuing a blowdown depressurization computer '

l calculation for a ' typical three-loop plant and a typical four-loop

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plant using the SATAN-VI computer coc'e.  !

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ATTACHMENT A Page 6

2. Hot Assembly Rod Heatup Analysis The hot rod heatup calculations would typically show the largest effect of the changes. Hot rod heatup computer analysis calculations were performed using the LOCBART computer code to assess the effect of the changes on the hot assembly average rod, hot rod, and adjacent rod.

3. Determination of the Effect on the PCT The effect of the changes on the calculated PCT was determined by performing a calculation for typical three-loop and four-loop plants using the BASH Evaluation Model. The analysis calculations confirmed that the effect of the ECCS Evaluation Model changes were insignificant as defined by 10 CFR 50.46(a)(3)(1). The calculations showed that the PCTs increased by less than 100F for the BASH Evaluation Model.

IV. BESULTANT LARGE-BREAK LOCA PCT As discussed above, modifications to the Westinghouse large-break LOCA ECCS Evaluation Model affect the large-break LOCA analysis results by altering the PCT as shown below:

Unit 1 Unit 2 l A. Analysis Calculated Result (Analysis-of-Hecord) ZQ120F ?J 380F B. Steam Generator Flow Area Reduction + 500F + 500F C. Fuel Rod initial Conditions Inconsistency Revisions + 100F + 100F D. ECCS Evaluation Model Modifications Resultant PCT 21090F 21980F l

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 is 500F for Units 1 and 2. It is-recognized that this represents a significant error, as defined in 10CFF.50.46 (a)(3)(ii). However, the-NRC has been previously notified of the 500F penalty associated with the -steam generator flow area reduction in l

References 3 and 6. Reference 6 includes justification to reduce this penalty to 60F. Reference 6 is currently under NRC review.

I i Note that -the PCT penalty associated with the effect of containment purge lines being open coincident with the'large-break LOCA. event was - fully incorporated into the analysis-of-record results repci tad above.

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l ATTACHMENT A Page 7 SMALL-BREAK LOCA

1. ECCS EVALUATION MODEL The small-break LOCA analyses for farley Units 1 and 2 were also-examined to assess the effect of the applicable modifications to the Westinghouse ECCS Evaluation Models on PCT results reported in Chapter 15, Section 3 of the FSAR. The small-break LOCA analyses results were calculated using the 1974 small-break LOCA ECCS Evaluation Model incorporating the WFLASH ,

analysis trchnology. For Farley - Units 1 -and 2, the- limiting size small-break resulted from a six-inch equivalent diameter break in the cold leg. The calculated PCT was 17120F. The analysis assumed the following ,

information important to the small-break LOCA analyses: >

o Core Power = 1 02 X 2652 MWT o 17x17 Standard fuel Assembly o FQ = 2.32 o F-delta-H = 1,55 o Auxiliary Feedwater Flow - 1050 gpm (Total)

The modifications to the Westinghouse ECC3 Evaluation Models which could affect the small-break LOCA analysis results found in Chatter 15, Section 3 in the Farley Units 1. and 2 FSAR are described below,

11. WFLASH ECCS EVALUF ION MODEL CODE-Following .the accident at Three Mile Island Unit 2, additional attention was focused on the small-break LOCA. . Westinghouse submitted a report, WCAP-9600 -(Reference 7), to the DNRC detailing ~ the performance of-the-Westinghouse small-break LOCA Evaluation Model which- utilized the WFLASH

-computer code. -In'NUREG-0611 (Reference 8)', the NRC staff questioned the validity .of certain models in--the WFLASH computcr code and required licensees.to justify continued acceptance'of:.the model.--Section II.K.3.30-of ~NUREG-0737 (Reference 9) clarified ~ the NRC: -post-TMI requirements regarding small-break LOCA modeling'and required. licensees to revise their

- small-break LOCA ECCS models along the gEidelines'specified in NUREG-0611.

L l: Following the issuance of - NUREG-0737, Westinghouse /and . the Westinghouse -

E Owners Group decided to develop the NOTRUMP (Reference 10) computer code  ;

L for use in' a _new small-break LOCA- ECCS- Evaluation Model (Reference 11).

i- The NRC approved theLuse of NOTRUMP for small-break LOCA ECCS analyses in May 1985.

Since approval of the NOTRUMP small-break LOCA ECCS Evaluation -Model' in 1985, the WFLASH. computer code has not been maintained as -part of the-Westinghouse ECCS Evaluation Model computer code.

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ATTACHMENT A Page 8 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 12) relaxed the requirements of item !!.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 would not be required, in WCAP-lll45 (Reference 13), Westinghouse and the Westinghouse Owners Group demonstrated that the results obtained from calculations with WFLASH were conservative relative to those obtained with NOTRUMP. Compliance with item II.K.3.31 of NUREG-0737 for Farley has been completed by referencing WCAP-11145 and determined to be acceptable by the NRC.

Westinghouse, therefore, has not been modifying the basic WFLASH small-break LOCA ECCS Evaluation Model. Thus, there are no modifications to the WFLASH small-break LOCA ECCS Evaluation Model to report.

Ill. SSLOCTA-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 l 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.

3) Modifications supporting a general upgrade of the computer program were implerrented as follows:

a) removal of unused or redundant coding,

, b) better coding organization to increase the efficiency of calculations, and i

j c) improvements in user friendliness

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ATTACHMENT A Page 9 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 tho equation for the density of Uraniunt-Dioxide was corrected to calculate thermal expansion only in two dimensions, which is consistent with the way in which the fuel rod is modeled in the LOCTA codes.

b) The correlation for the specific heat of water vapor at temperatures over 15900F was improved.

c) An error in the equation for the pellet / clad contact pressure was corrected.

The Uranium-Dioxide density correction is estimated to have a maximum PCT benefit of less than 20F, while tha contact resistance modification has no PCT effect since it is not used.

IV. SAFETY INJECTION BACK PRESSURE 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 460F increase. .This 460F increase has been previously reported in a letter to the NRC from Alabama Power Company dated January 14, 1988, and in Reference 2.

V. fillL30D INITIAL CONDITION INCONSISTENCY REVISIONS During the review of the original Westinghouse ECCS Evaluation Model following the promulgation of 10 CFR 50.46 in 1974, Westinghouse committed to maintain consistency between future LOCA fuel rod computer models and the fuel rod design computer models used to predict futl rod normal operation performance. These fuel rod design codes are also used to establish initial conditions for the LOCA analysis.

Chance Description It was found that the small-break LOCA code versions were not consistent with fuel design codes in the following areas:

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ATTACHMENT A Page 10

1. The LOCA codes were not consistent with the fuel rod design code relative to the flux depression factors at higher fuel enrichment.

2. The LOCA codes were not consistent with the fuel rod design code relative to the fuel rod gap gas conductivities and pellet surface roughness models.

3. The coding of the pellet / clad contact resistance model required revision.

Modifications were made to the fuel rod models used in the LOCA Evaluation Models to maintain consistency with the latest approved version of the fuel rod design code.

Effect of Chanaes The changes made to make the LOCA fuel rod models consistent with the fuel design codes were judged to be insignificant, as defined by 10 CFR 50.46(a)(3)(i). To quantify the effect on the calculated PCT, calculations were performed which incorporated the changes.

Small-Break LOCA The effect of the changes on the small-break LOCA analysis PCT calculations was determined using the 1985 small-break calculations for a typical three-loop plant and a typical four-loop plant. The analysis calculations confirmed that the effect of the changes on the small-break LOCA ECCS Evaluation Model was insignificant as defined by 10 CFR 50.46(a)(3)(1). The calculations showed that 370F would bound the effect l

on the calculated PCTs for the three-loop plants. This 370F increase has been previously reported to the NRC in Reference 14.

VI. R00 INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION Chanae Descriotipn l The Westinghouse small-break LOCA ECCS Evaluation Model analyses assume l that higher fuel rod initial fill pressure leads to a higher calculated

! PCT, as found in studies with the Westinghouse large-break LOCA ECCS l Evaluation Model. However, lower fuel rod internal pressure could result in decreased cladding creep (rod swelling) away from-the fuel pellets when the fuel rod internal pressure was higher than the reactor coolant system (RCS) pressure. A lower fuel rod initial fill pressure could then result in a higher calculated peak cladding temperature.

The Westinghouse small-break LOCA cladding strain model is bastd upon a correlation of Hardy's data. Evaluation of the limiting fuel rod initial fill pressure assumption revealed-that this model was used outside of the applicable range in the small-break LOCA Evaluation Model calculations, allowing the cladding to expand and contract more rapidly than it should.

ATTACHMENT A Page 11-The model was corrected to fit applicable data over the range of small-break LOCA conditions. Correction of the cladding strain model affects the small-break LOCA Evaluation Model calculations through the fuel rod inteinal pressure initial condition assumption.

Effect of Chances Implementation of the corrected cladding creep equation results in a small reduction in the pellet to cladding gap when the RCS pressure exceeds the rod internal pressure, and increases the gap after RCS pressure -falls below the rod internal pressure. Since _the cladding typically demonstrates very_ little creep toward the fuel

- pellet prior to core uncovery when the RCS pressure exceeds the rod internal pressure, implementation of the.-correlation for the appropriate range has a negligible- benefit on the-PCT calculation during this sortion of the transient. However, after the RCS pressure falls below tie rod internal pressure, implementation of an accurate correlation for cladding creep in the small-break LOCA analyses would reduce the expansion of the cladding away- from the fuel, compared to what was previously calculated, and

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results in a PCT penalty because the cladding is closer to the fuel.

Calculations were performed to assess the effect of the cladding strain modifications for the limiting three-inch equivalent diameter cold leg break in typical three-loop and four-loop plants. The results indicated that the change to the calculated PCT resulting_from the cladding strain ,

model change would be less then 200F, The effect'on the calculated PCT depended upon when the. peak cladding temperature-occurs and whether the rod internal pressure was above or below the. system pressure when the PCT occurs. For the range of fuel rod- internal pressure initial conditions, the combined effect of the fuel rod internal pressure -and the cladding strain model revision is bounded by-400F. This 400F increase has been previously reported to the NRC in Reference 14.

VII. AVXILIARY FEEDWATER ENTHALPY DELAY TIME INPUT Chanae Description In the small-break LOCA an'alysis, values are inserted for the Auxiliary Feedwater (AFW) delay time which would presume a certain' length of time from signal for pumps to startup and initiate delivery (60 seconds).

-Another delay is ' input for the -.enthalpy change. .This second delay accounts for the standing water in the pipe at feedwater inlet temperature-to ue purged, and for _the pumped AFW water at a reduced temperature :to finally ' arrive at the steam generator (SG)' injection point before considering any cooling -effect' from the AFW.

In WFLASH, the enthalpy:

delay is given in user documentation to- be a standard value of 132 seconds.

This issue was investigated for NOTRUMP following the discovery:of a user documentation -inconsistency, which, in effect, took credit for.the cooler water instantaneously, ignoring the time required to clear out the .line of -

the standing feedwater.

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ATTACHMENT A Page 12 When confirmatory calculations for the subject plant were performed to determine a realistic delay time for the AFW enthalpy change, it was found that the old WFLASH standard of 132 seconds was in doubt.. That is, the main feedline piping actually installed in the plant had volume great enough that, even assuming this volume of standing water to be swept out and precede the AFW lower enthalpy . flow delivery, the timo delay may in fact be greater than that assumed in - the analysis. -This determination prompted the evaluation for the WFLASH analyses as well. For the Farley WFLASH analysis, the enthalpy delay was input as 129.4 seconds.. This value was determined to be inconsistent with values used in other safety analyses. The evaluation for WFLASH analyses was to ascertain the actual enthalpy delay assumed and then to determine its adequacy-in view of plant piping cnnfigurations.

There are two basic mechanisms for heat removal during a small-break LOCA.

These are 1) heat removal via discharge of high enthalpy fluid through the break, and 2) heat transfer from the primary to secondary across the SG tubes. Increasing the delay time for the delivery of low enthalpy AFW into the SG secondary leaves a higher total energy on the secondary side.

This would tend to inhibit heat transfer from the primary to secondary and place greater reliance on break flow to remove energy from the RCS.

Smaller breaks rely more heavily on_ heat transfer across the SG tubes due to the limited break area resulting in limited break flow, and therefore would tend to be more sensitive to additional heat removal requirements.

On the other hand, the smaller breaks exhibit a slower blowdown transient resulting generally in a more moderate clad heatup rate. This trade-off suggests that the effects of delayed AFW enthalpy vary with break size, with no absolute certainty of which break-size will-be most affected by the delay. Therefore, the evaluation considered not only the limiting break size, but also other break sizes which could..potentially have a larger PCT effect which could result in a change in the most limiting break sire.

Effect of Chanaes In -work being done for another plant, an inconsistency was discovered between the main feedwater purge time assumed in the small-break . LOCA analyses utilizing the NOTRUMP Evaluation Model and the time indicated by pl ant geometry and operating conditions. _The purge _ time refers to the time required for- the _ low flow, low enthalpy AFW -to remove and replace stagnant high enthalpy main feedwater from the feedwater piping after AFW actuation.

The root cause of 'this error was determined not to be model-related in-that the computer coding was correct. It was an inconsistency- in the l' Evaluation- Model documentation provided to the user. By. adhering to user instructions, these incorrectly based input values would be input'into the-Evaluation Model. _The user documentation was subsequently corrected.

However, inasmuch as the NOTRUMP treatment of this AFW enthalpy delay is similar to the approach taken with the older WFLASH code, a general. survey of the delay time assumption iin plant-specific analyses was performed. _. As a result.of this survey,-it was found that the analysis-of-record for the

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ATTACHMENT A Page 13 Farley units was also affected negatively by assumption of inadequate AFW enthalpy delay time. An evaluation was performed to assess the effect it might have on the reported small-break analysis-of-record. That evaluation is described below.

Given that the small-break LOCA analysis-of-record has the AFW enthalpy delay input as 129.4, an estimate of the actual purge volume and delay time was calculated. Data extracted from the analysis showed an input enthalpy for the main feedwater as 416.2 Btu /lbm, and the auxiliary feedwater enthalpy input as 90.4 Btu /lbm. The auxiliary flow rate was 1050 gpm, or 350 gpm/ loop. Based on a sweep volume consistent with that used as a basis for feedline break analyses, the sweep time, once AFW is initiated, was calculated to be 139.7 seconds. Then, when considering the total delay time, the enthalpy change should not be credited until 199.7 seconds into the transient as opposed to 129.4 seconds assumed in the analysis-of-record. This calculation confirmed that the Farley WFLASH small-break LOCA analysis would be affected by this issue.

The PCT effect of this increased delay was calculated by considering the difference in enthalpy between main feedwater and auxiliary feedwater and projecting the effect of removing this amount of cooling capacity from the secondary (effectively assuming this unwarranted cooling would remain in the RCS primary side). As noted above, break flow is the other prominent heat removal mee.hanism in the transient. The PCT effect could then be estimated by assuming this increase of energy could be removed from the primary side solely by break flow. The result was expressed in terms of an additional amount of transient time to remove this energy before the PCT conditions would otherwise be achieved. Assuming continual rod heatup during this additional time of core uncovery, the projected PCT increase could then be extrapolated. The projected peak clad temperature penalty was calculated as 38.320F, which was rounded up to 390F for reporting purposes. This 390F increase has been previously reported to the NRC in Reference 14.

VIII. RESULTANT SMAll-BREAK LOCA PCT l

As discussed above, modifications to the Westinghouse small-break LOCA ECCS Evaluation Model could affect the small-break LOCA analysis results by altering the PCT as shown below.

Unit 1 Unit 2 A. Analysis Calculated Result (Analysis-of-Record) lll20F lll20F B. Modifications to Westinghouse ECCS Evaluation Model -

20F -

20F C. Safety Injection Ba:k Pressure Evaluation + 460F + 460F D. Fuel Rod Initial Condition Inconsistency Revisions + 370F + _370F E. Rod Internal Pressure Initial Coadition Assumption + 400F + 400F

!. Auxiliary Feedwater Enthalpy Deley Time Input +__120F + 390F G. ECCS Evaluation Model Modifications Resultant PCT 18720F 18720F

y _ _ _ - . ...._ _. . _. _ _ _ _ _ _ . _ _ _ __ __

ATTACHMENT A Page 14 Therefore, the sum of the absolute magnitude of the PCT assessments introduced as the result of the modifications and errors in the small-break LOCA ECCS Evaluation Model is 1640F for Units 1 and 2. 'Since this sum is greater than 500F, indicating a significant change, a 30-day report was sent to the NRC detailing all of the above assessments (Reference 14). It was noted at that time that the small-break LOCA had been reanalyzed as part of the VANTAGE-5 effort, taking these model changes into account. The NRC is currently reviewing this analysis.

CONCLUSIONS An evaluation of the effect of modifications to the Westinghouse ECCS Evaluation -

Model was performed for both the large-break LOCA and small-break LOCA analysis ,

results. When the effects of the ECCS Evaluation 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 maintained.

REFERENCES

1. " Emergency Core Cooling Systems; Revisions to Acceptance Criteria,_" Federal Register, Vol . 53, No.180, pp. 35996-36005,_ dated September- 16, 1988.

2. Letter from J. D. Woodard to USNRC, " Joseph M.- Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model--Changes Report- for-1990," March 21, 1991.

3. Letter from W. G. Hairston, III- to USNRC, " Joseph H. Farley Nuclear -Plant -

l Unit 1 RTO Bypass Elimination and Steam Generator Tube Plugging _ Amendment,"

October 26,-1990.

~

4. Letter from W. G. Hairston, III to USNRC, " Joseph it. -Farley Nuclear Plant -

Unit 2 Peak Clad Temperature (PCT)- Calculation," February 18,11991. c

5. Letter _ from R. P.: Mcdonald to USNRC, " Joseph M. FarleyLNuclear; Plant-- Units I

1 and 2 Proposed' Steam Generator Tube -Plugging Limit and Heat l Flux' Hot Channel Factor Technical Specifications Changes," June 2,1987.

6.- Letter- from- J. D. Woodard to- USNRC, - Joseph M.- ' Farley; Nuclear: Plant VANTAGE-5 Fuel Design Amendment,"LSeptember 10, 1991~.

7. " Report on Small Break Accidents- for Westinghouse Nuclear Steam LSupply

-System," WCAP-9601 (Non-Proprietary), June 1979. WCAP-9600__(Proprietary),

June 1979.

i

8. " Generic: Evaluation of Feedwater -Transients and-Small Break ~ Loss-of-Coolant l Accidents in-Westinghouse Designed Operating Plants," NUREG-0611, LJanuary-1980.

9. " Clarification of THI- Action Plan _ Requirements," NUREG-0737,< November 1980.

_ . ~ . _ _. . . _ . _ = _ ,, _ . . _ .;...- _ _ ., _ _._ .

u ATTACMMENT A Page 15

10. "N01 RUMP - 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.

11. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"

WCAP-10054-P-A (Proprietary),WCAP-10081-A (Non-Proprietary), Lee, N., et al, August 1985.

12. " Clarification of TMI Action Plan item II.K.3.31," NRC Generic letter 83-35 from D. G. Eisenhut, November 2, 1983.

13. " Westinghouse Small Break ECCS Evaluation Model Generic Study With the NOTRUMP Code," WCAP-lll45, Rupprecht, S. D., et al., August 1985.

14. Letter from J. D. Woodard to USNRC, " Joseph M. Farley Nuclear Plant Peak Clad Temperature (PCT) Calculation," July 26, 1991.

15. "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, (Non-proprietary), July 1974.

16. "J. M. Farley Units 1 and 2, SG Tube Plugging Criteria for ODSCC at Tube Support Plates," WCAP-12871, Revision 2, February 1992.

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4 ATTACHMENT B EFFECT OF SAFETY. EVALUATIONS AGAINST THE LOCA ANALYSIS RESULTS LARGE-BREAK LOCA DESCRIPTION OF PLANT MODIFICATIONS 1 As discussed below, the large-break Loss-of-Coolant Accident (LOCA) analysis results have been supplemented by a safety evaluation of a plant design change under 10 CFR 50.59 that has assessed a penalty to the fuel peak cladding temperature (PCT)..

A s.fety evaluation of the effect of loose parts in the Unit 1 RCS (unrecovered grio strap sections) was performed to determine the effect of this condition-on the large-break LOCA analysis PCT results. Completed in 1988.- the evaluation determined that the Unit I large-break LOCA analysis PCT re>ults could be affected by a 600F increase, l RESULTANT LARGE-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 Modifications Reported in Attachment A 219.9.0F 21280F

1. 10 CFR 50.59 Safety Evaluation for Loose Parts (Grids) +__kDOF + N/A Total Large-Break Resultant PCT 21690F 21980F

-Therefore, the total PCT assessments introduced as a result of plant modifications are 600F for Unit 1 and'00F for Unit 2.

SMALL-BREAK LOCA OfSCRIPTION OF PLANT MODIFICATIONS As-described below, the small-break LOCA analysis results have been supplemented by safety evaluations- of _ plant design changes under 10 CFR 50.59 which have:

assessed penalties to the PCT; i

l

u-

., , a ATTACHMENT B_ ,

Page 2

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 Farley Units 1 and_2 Final _ Safety Analysis Report (FSAR),the evaluation of the effect of this plant design 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 affected by a 1170F. increase.

This evaluation was completed in 1982,

2) A safety evaluation of the effect of loose parts in the Unit 1 RCS (unrecovered grid strap sections) was- performed. An e_ valuation of : the ,

effect of this condition on the small-break LOCA analysis - PCT results was performed. The evaluation determined that the Unit I small-break LOCA analysis PCT results could be affected by a 320F increase. This evaluation was completed in 1988.

3) A safety _ evaluation of the effect of the temperature uncertainties on the small-break LOCA was perfcrmed a s part of a proposed - Technical Specifications change to. remove the RTD bypass loops for ' Unit 1. The-temperature uncertainties -are ' associated with _ the: accuracy of the instrumentations, the accuracy _ of- the calibration- - equipment,- l and Jthe procedures used for- calibrating and reading the instrumentations. The evaluation determined that the small_-break LOCA analysis! PCT results could-be affected-by a 20F 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.-

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 _ . 18120F 18_J_20F

1. 10 CFR 50.59 Safety Evaluation:for Upflow Cenversion. + 1170F N/A-

2. 10 CFR 50.59 Safety Evaluation -for_ Loose Parts-(Grids) + 320F N/A__

3. 10 CFR 50.59 Safety Evaluation for RTD Temperature Uncertainty + 20F +_ _20F-Total Resultant PCT 20230F -18740F-Therefore, _ the resultant: PCT assessments introduced as a result .of plant.

modifications.are 1510F for. Unit 1 and 20F for Unit 2.

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

y ATTACHMENT B Page 3 CONCLUSIONS An evaluation of the effect of modifications to the Westinghouse ECCS Evaluation Model was perfor. 3d for both the large-break l.0CA and small-break LOCA licensing basis analysis results. It was determined that compliance with the requirements of 10 CFR 50.46 would be maintained when plant design changes, performed under 10 CFR 50.59, which could sffect the LOCA analysis results were combined with the effect of the ECCS Evaluation Model modifications applicable to Farley Units 1 and 2.

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l 002022

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a bc: Mr. R. P. Mcdonald Mr. J. D. Woodard Mr. K. W. McCracken Mr. D. N. Morey Mr. J. W. McGowan Dr. W. M. Andrews Mr. O. Batum Mr. S. Fulmer Mr. R. H. Marlow Mr. J. A. Knochel Commitment Tracking System (2)

Document Control (2)

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