ML20076C427

Forwards Annual Rept of Corrections to Westinghouse ECCS Evaluation Models for Aug 1990 to June 1991

ML20076C427
Person / Time
Site:	Prairie Island
Issue date:	07/15/1991
From:	Parker T NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9107220045
Download: ML20076C427 (20)

Text

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Northern States Power Company l 414 Nicollet Mall Minneapohs. Minnesota 55401 1927 Teleph<,ne (612) 330-5500 July 15, 1991 U S Nuclear Regulatory Commission 10 CFR Part 50 Attn: Document Control Desk Section 50.46 Washington, DC 20555 PRAIRIE ISIAND NUCLEAR CENERATING PIANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Annual Report of Corrections to ECCS Evaluation Models Attached is the annual report of corrections to Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for the period from August 1990 to June 1991. This report is being submitted inaccordance with the provisions of 10 CFR 50, Section 50.46.

We have applied the applicable corrections noted in Attachment 1 to our current ECCS analyses of record, and found all analyses to be in compliance with the applicable acceptance criteria.

Please contact us if you require further infctmation related to this request.

/D?#'

Thomas M Parker Manager Nuclear Support Services c: Regional Administrator-III, NRC NRR Proj ect Manager, NRC Resident Inspector, NRC G Charnoff Attachments: Changes to the Westinghouse ECCS Evaluation Models 910722OO.r, 9107),e re me e.am m (i ,

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l ATTACHMENT 1-CHANGES TO THE WESTINGHollSE ECCS EVALUATION MODELS 1

AUGUST 1990 - MAY 1991 o

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Changes to the Westinghouse ECCS Evaluation Medels

1.0 INTRODUCTION

Provisions in 10CFR50.46 require the reporting of corrections to or changos in the ECCS Evaluation Model (EM) approved for use in performing safety ana yses for the loss of coolant accident (LOCA). This report describes correctior and revisions to the Westinghouse ECCS EM in the period from August 1990 t'iro;gh May 1991. The current Westinghouse ECCS EM are named as listed in Table 1, and consist of several computer codes'with specific functions.

Westinghouse has completed the evaluation of several items related to the Westinghouse ECCS Evaluation Models listed in Table 1. Each of these items is discussed in the following sections, which include a description of the item, the assessment which was performed, the resulting change to the Evaluation Model, and the effect of the change on the PCT.

Some of the subjects discussed represent changes to program coding or to inputs directly related to the physical modelt, or solution technique. These are described in Section 2.0.

-Some items represent changes to the assumptions made when the Evaluation Model is applied to a specific plant. These are discussed in Section 3. Al so included, for information, are items for which a technical assessment is continuing, and items for which it was concluded that no change was necessary.

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E:\ECC$ EM\EM 91 As 8

TABLE 1

SUMMARY

OF WESTINGHOUSE ECCS EVALUATION MODELS NAME: 1978 MODEL APPllCATION: Analysis of large Break LOCA CODES USED: PURPOSE:

REFERENCE:

SATAN-VI Blowdown h/draulic transieni 1.

WREFLOOD Reflood hydraulic transient 2.

LOCTA Fuel rod thermal transient 3.

C0C0 or LOTIC Containment pressure transient 4.,5.

NOTE: The NRC has determined that this EM is no longer acceptable for use in new analyses. However, it serves as the licensing basis for some plants.

NAME: 1981 NonEL APPLICATION: Analysis of large Break LOCA CODES USED: PURPOSE:

REFERENCE:

SATAN-VI Blowdown hydraulic transient 1.,6.

WREFLOOD Reficod hydraulic transient 2.

LOCTA Fuel rod thermal transient 3.

C0C0 or LOTIC Containment pressure transient 4.,5.

NOTE: This model superseded the 1978 EM and included changes to the flow blockage model, consistent with requirements in NUREG 0630.

NAME: 1981 MODll WITH BART APPLICATION: Analysis of Large Break LOCA CODES USED: PURPOSE

REFERENCE:

SATAN-VI Blowdown hydraulic transient 1.,6.

INTERIM-WREFLOOD Reflood nydraulic transient 2.,7.

BART Hot assembly thermohydraulics 7.

INTERIM-LOCTA Fuel rod thermal transient 3.

COCO or LOTIC Containment pressure transient 4.,5.

NOTE: This model was developed to provide a more realistic calculation of heat transfer during the reflood portion of the transient.

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TABLE I (CONTINUE 0)

NAME: 1981 MODEL WITH BASH APPLICATION: Analysis of Large Break LOCA CODES USED: PURPOSE

REFERENCE:

SATAN-VI Blowdown hydraulic transient I.,6.

BASH Reflood hydraulic transient 8.

LOCBART Hot assembly thermohydraulics 3.,7.,8.

and fuel rod thermal transient l WREFLOOD/COC0/LOTIC Containment pressure transient 2.,4.,5.,8.

l NOTE: this model wc3 developed to further improve the reflood portion of the Evaluation Model.

NAME: UPI WCOBRA/ TRAC APPLICATION: Analysis of Large Break LOCA for plants with upper plenum safety injection.

CODES USED: EVRPOSE REFERENCE COBRA / TRAC' Combined thermal and hydraulic transient 9.

NOTE: This model uses a best estimate computer code, but includes required l

features of Appendix K.

l NAME: 1975 SBLOCA MODEL APPLICATION: Analysis of Small Break LOCA l

CODES USED: PURPOSE REFERENCE WFLASH System hydraulic transient 10., 11.

SBLOCTA Fueirod Thermal transient 3.

l NOTE: This model is no longer used, but some plants are licensed under this methodology.

l NAME: 1985 SBLOCA MODEL l

APPLICATION: Analysis of Small Break LOCA CODES USED: PURPOSE REFERENCE NOTRUMP System Hydraulic transient 12., 13 SBLOCTA Fuel rod thermal transient 3.

NOTE: This model was developed to provide more realistic SBLOCA simulations, as required by NRC, following THI, E:\ECCS EM\EM 91 A5 10

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2.0 EVALUATION MODEL CODE CHANGES This section describes changes and revisions to the Westinghouse ECCS Evaluation Model computer codes. Except where noted, these corrections will be implemented in all future applications of the Evaluation Model.

2.1 FUEL ROD MODEL REVISIONS During the review of the original Westinghouse ECCS Evaluation Model following the promulgation of 10CFR50.46 in 1974, Westinghouse committed to maintain consistency between future loss-of-coolant accident (LOCA) fuel rod computer nodels and the fuel rod design computer models used to predict fuel rod normal operation performance. These fuel rod design codes are also used to establish initial conditions for the LOCA analysis.

Chance Descriotion:

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

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.

Mcdifications 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 3, was being calculated twice each time step instead of once. The coding was corrected to properly integrate the strain rate equation.

Affected Evaluation Models:

1981 Large Break LOCA Evaluation Model 1981 Large Break LOCA Evaluation Model, With BART 1981 Large Break LOCA Evaluation Model, With BASH 1975 Small Break LOCA Evaluation Model 1985 Small Break LOCA Evaluation Model l

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t 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 10CFRSO.46(a)(i).

To quantify the effect on the calculated peak cladding temperature (PCT),

calculations were performed which incorporated the changes, including the cl adding 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.

a) Large Break LOCA The effect of the changes on the large break LOCA peak cladding temperature was determined using the BASH 1arge break LOCA Evaluation Model. The effects were judged applicabi' to older Evaluation Models.

Several calculations were performed to a ess 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 performing a blowdown depressurization computer calculation for a typical three-loop plant and a typical four-loop plant using the SATAN-VI computer code.

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 Peak Cladding Temperature The effect of the changes on the calculated peak cladding temperature 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 10CFR50.46(a)(3)(i). The calculations showed that the peak cladding temperatures increased by less than by 10*F for the BASH Evaluation Model. It was judged that 25'F would bound the effect on the peak cladding temperature for the BART Evaluation Model, while calculations performed for the Westinghouse 1981 Evaluation Model showed that the peak cladding temperature could increase by approximately 41'F.

E:\ECCS EM\EM 91- A5 - 12

b) Small Break LOCA The effect of the changes en the small break LOCA analysis peak cladding temperature calculations was determined using the 1985 small break LOCA Evaluation Model by performing a computer analysit 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 were insignificant as defined by 10CFR50.46(a)(3)(i). The calculations showed that 37'F would bound the effect on the calculated peak cladding temperatures for the four-loop plants and the three-loop pl ants , it was judged that an increase of 37'F would bound the effect of the changes for the 2-loop plants.

Status:

Changes completed and implemented.

2.2 SMALL BREAK LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION Chance Descriotion:

The Westinghouse small break losr -of-coolant accident (LOCA) emergency core cooling system (ECCS) Evaluation Model analyses assume that higher fuel rod initial fill pressure leads to a higher calculated peak cladding temperature (PCT), as found in studies with the Westinghouse large break LOCA ECCS 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 based upon a correlation of Hardy's data, as described in Section 3.5.1 of Reference 3.

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. 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 internal pressure initial condition assumption.

Affected Evaluation Models:

1975 Small Break LOCA Evaluation Model 1985 Small Break LOCA Evaluation Model 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 E \ECCS EM\EM-91 A5 - 13

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

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internal -pressure and increases the gap after RCS pressure falls below the rod internal pressure. Since the cladding typically demonstrates very little creep toward U fuel pellet prior to core uncovery when the RCS pressure exceeds the; m internal pressure, implementation of the correlation for the appropriate- range has a negligible benefit on the peak cladding temperature calculation during this portion of the transient. However, after the RCS pressure' falls below the rod internal pressure, implementation of an accurate correlation for cladding creep in small break LOCA analyses would reduce the expansion of the cladding away from the fuel compared to what was previouslyL calculated and 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 peak cladding t *oerature resulting from the cladding strain model change would be less than 20*F. The effect on the calculated peak cladding temperature depended upon when the peak cladding temperature occurs and whether the rod- internal pressure was above or below the system pressure when the peak cladding temperature 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 typically bounded by 40*F. However, in an extreme case the combined effect could be as large as 60'F.

Status:

Modifications to the _small break LOCA cladding strain model for application to the appropriate range of conditions have been implemented and the effect of l- the rod internal- pressure initial condition assumption assessed. Since i changes to -_the strain model - may also affect assumptions concerning the limiting time- in'the core cycle due to the propensity for cladding-burst, the small break LOCA -limiting time in- the core cycle assumptions are being reviewed _ and a ~ conclusion regarding their continued- validity will be determined by the end of 1991, 2.3 UPI MODEL REVISIONS Chance'Descriotion:

Revisions were made to the MCOBRA/ TRAC large break LOCA Evaluatton Model used l

for plants equipped with upper plenum injection-(UPI). These changes, and

their effects, were previously reported to the NRC (Reference 14).

Affected-Evaluation Model UPI WCOBRA/ TRAC Status:

Complete.

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2.4 NOTRUMP CODE SOLUTION CONVERGENCE Chance Descriotion:

In the development of the NOTRUMP small break LOCA ECCS Evaluation Model, a number of noding sensitivity studies were performed to demonstrate acceptable solution convergence as required by Appendix K to 10CFR50. Temporal solution convergence sensitivity studies were performed by varying input parameters which govern the rate of change of key process variables, such as changes in the pressure, mass, and internal energy. Standard input values were specified for the input parameters which govern the time step size selection. However, since the initial studies, modifications were made to the NOTRUMP computer program to enhance code performance and implement necessary modifications (Reference 15). Subsequent to the modifications, solution convergence was not re-confirmed.

To analyze changes in pl ant operating conditions, sensitivity studies were performed with the NOTRUMP computer code for variations in initial RCS pressure, auxiliary feedwater flow rates, power distribution, etc., which resulted in peak cladding temperature (PCT) variations which were greater than anticipated based upon engineering judgement. In addition, the direction of the PCT variation conflicted with engineering judgement expectations in some cases. The unexpected variability of the sensitivity study results indicated that the numerical solution may not be properly converged.

Sensitivity studies were performed for the time step size selection criteria which culminated in a revision to the recommended tir.e step size selection criteria inputs. Fixed input values originally recommended for the steady state and all break transient calculations were modified to assure converged results. The NOTRUMP code was re-verified against the SVT-08 Semiscale experiment and it was confirmed that the code adequately predicts key small break phenomena.

Affected Models:

1985 Small Break LOCA Evaluation Model Effect of Chanaes:

Generally, the modifications result in small shifts in timing of core uncovery and recovery. However, these changes may result in a change in the calculated peak cladding temperature which exceeds 50'F for some pl ants . Based on representative calculations, however, this change will most likely result in a reduction in the calculated peak cladding temperature. Since the potential beneficial effect of a non-converged solution is plant specific, a generic PCT effect cannot be provided. However, it has been concluded that current licensing basis results remain valid since the results are conservative relative to the change.

Status:

This change has been implemented and will be used in all future analyses.

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3.0 EVALUATION MODEL APPLICATION CHANGES The following section describes changes in the way the LOCA evaluation model is applied, or provides additional information on the method of application.

3.1 LARGE BREAK LOCA POWER DISTRIBUTION ASSUMPTION Backaround:

Appendix K to 10CFR50 requires that the power distribution which results in the most severe calculated conseque'nces be used in the ECCS Evaluation Model calculations. The power distributions to be studied are those expected to occur during the core lifetime.

The current basis for all Westinghouse large LOCA Evaluation Model is the chopped cosine power distribution. This distribution is symmetrical and is defined by two quantities: the ratio of peak linear power relative to the average (FQT), and the ratio of hot rod integral power relative to the average (FAH). This power distribution was found to produce the highest peak cladding temperature (PCT) when compared to power distributions skewed to the top or bottom of the core in studies performed by Westinghouse and submitted to the NRC. Typically the power distributions were assumed to peak at discrete elevations in the core (4, 6, 8, and 10 feet). It was also assumed that the key parameters affecting PCT were the FQT, FAH, the peak power location, and integral of power to the peak power elevation.

Calculations performed with the advanced LOCA Evaluation Models, BART and BASH, which examined peak power locations and power distributions which were not considered in the original analyses, under some circumstances lead to PCTs greater than those calculated with the cosine distribution. This behavior was revealed when performing power distribution studies for core designs with relatively low FQT and relatively high FAH. Further studies revealed that, in addition to FQT, FAH, and the peak power location, the nature of the axial distribution of power affected the results. That is, two power distributions with the same FQT, FAH, and peak power location, but whose power was distributed differently along the rod could result in significantly different PCTs.

Westinghouse has completed an analysis effort to understand and properly account for the effect of skewed power distributions on the calculated large break LOCA PCT. This effort i.vluded the identification of the worst power distributions that could occur during core life with full consideration of the current generation of reload co"e designs.

Chance Descriotion:

As a result of these studies, revisions have been made to the current reload and safety analysis methodology which accounts for the variability in power distributions from cycle to cycle and plant to plant. This revision provides a means of determining that the current licensing basis (i.e., the chopped cosine) is expected to remain limiting, but also provides for identifying and analyzing the most severe expected power distribution, if different from the chopped cosine.

E:\tCCS EM\tM 91 A5 16

6 Affected Eyv.aluation Models:

1981 ECCS Evaluation Model 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH Status:

In order to verify that a plant was not affected by this item, a large break LOCA power distribution surveillance factor was applied to confirm that the power shapes identified as potentially being more limiting are not present.

The owners of the affected pl ants were advised v temporarily apply this surveillance factor to their normal' flux map measurements. 'n some cases, a temporary 100*F PCT margin allocation was applied, rather than the surveillance factor. This margin assured that, if limiting power shapes did occur,10CFR50.46 limits would still be met.

The process described in Reference 16 will be used to assess specific core designs. In this process, each power distribution calculated in the core design will be evaluated to determine whether it is more limiting than the cosine power distribution. Adjustments will be made to the core design operating bands to eliminate these limiting distributions and surveillance factors will be defined to assure that plant safety limits are met. This will assure that a change to the ECCS Evaluation Model is not required, since the chopped cosine power distribution will remain limiting.

3.2 LARGE BREAK LOCA BURST AND BLOCKAGE ASSUMPTION Backaround:

The cladding swelling and flow blockage models were reviewed in detail during the NRC's evaluation of the Westinghouse Evaluation Model. However, the use of the average rod in the hot assemt,1y may not have been documented in a manner detailed enough to allow the staff to adequately assess this aspect of the model.

Appendix K to 10CFR50 requires consideration of the effects of flow blockage resulting from the swelling and rupture of the fuel rods during a loss-of-coolant accident (LOCA). 10CFR50 Appendix K Paragraph I.B states:

....To be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not underestimated."

In Westinghouse ECCS Evaluation Model calculations, the average rod in the hot assembly is used as the basis for calculating the effects of flow blockage.

If a significant number of fuel rods in the hot assembly are operating at power , levels greater than that of the average rod, the time at which cladding

welling and rupture is calculated to occur may be predicted later in the LOCA transient, since the lower power rod will take longer to heat up to levels where swelling and rupture will occur.

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A review of the Westinghouse model used to predict assembly blockage waa performed. This model was developed from the Westinghouse Multi-Rod Burst

, Tests (MRBT) and was the model used to determine assembly wide blockage until replaced by the NUREG-0630 model starting in 1980. There models provide the means for determining assembly wide blockage once the mean burst strain has been established. Implementation of these burst models has relied upon the average rod to provide the mean burst strain. The average rod is a low power rod producing the power of the average of rods in the hot assembly and is primarily used to calculate the enthalpy rise in the hot assembly. Use of the average rod in the model assumes that the time at which blockage is calculated to occur is represented by the burst of the average rod. A review of current hot assembly power distributions indicates that in general the average rod in the hot assembly is also representative of the largest number of rods in the assembly, so that. burst of this rod adequately represents when most of the i rods will burst. With this representation, however, the true onset of blockage would likely begin earlier, as the highest power rods reach their burst temperature. This time is estimated to be a f?w seconds prior to the time when the average rod bursts.

Large break '0CA Evaluation Models which use BART or BASH simulate the hot assembly rod with the actual average power, while older Evaluation Models use an average rod power which is adjusted downward to account t'.c thimbles (this is described in detail in Addendum 3 to reference (7)). If burst occurs after the flooding rate has fallen below one inch per second, the time at which the blockage penalty is calculated will be delayed for these older Evaluation Models.

Chance Descriotion:

Ample experimental evidence currently exists which shows that flow blockage does not result in a heat transfer penalty during a LOCA. In addition, newer Evaluation Models have been developed and licensed which demonstrate that the older Evaluation Models contain a substantial amount of conservatism.

Westinghouse concluded that further artificial changes to the ECCS Evaluation Models to force the calculation of an earlier burst time were not necessary.

In rare instances where burst has not occured prior to the flooding rate falling below 1.0-inch /second, the results of the ECCS analysis calculation are supplemented by a permanent assessment of margin. Typically this will only occur in cases where the calculated PCT is low. Westinghouse concludes that no model change is required to calculate an earlier burst time.

Affected Evaluation X3dels:

1978 ECCS Evaluation Model 1981 ECCS Evaluation Model 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH 4

Status:

Complete.

Er\ECC$ EM\tM 91+A5 1B

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3.3 STEAM GENERATOR FLOW AREn Backoround:.

Licensees are normally required to provide assurance that there exists only an extremely low probability of abnormal leakage or gross rupture of any part of the reactor coolant pressure boundary (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 th:t guide, the staff required analytical and 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 provides the basis for establishing criteria for removing from service tubes which had experienced significant degradation, i 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 deformation. 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, wnich potentially could increase the calculated PCT.

The effect of tube deformation and flew 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 ralculated phenomena:

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

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

l 3. During a postulated large LOCA, the primary side depressurizes to containment pressure. Applying the resulting pressure differential to the deformed tubes causes some of these tubes to collapse, and reduces the

( effective 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 peak cladding temperature (PCT).

l 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 steam generator outlet break, because this break was judged to result in the greatest loads on the I steam generator, and thus the greatest flow area reduction. It was concluded that the steam generator would continue to rieet its safety function because the degree of flow area reduction was small, and the postulated break at the steam generator outlet resulted in a low PCT.

I Ei\ECCs EMiEM 91 A5 19

In April or 1990, in considering the effect 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 analyses. With SSE loadinjs, 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 primary pipe break affecting LOCA PCT, i.e., the react:r vessel inlet break (cold leg break LOCA), had to be evaluated to confirm that 10CFR 50.46 limits continue to be met and that the affected steam generators will continue to perform their intended safety function.

Consequently, the action was taken to address th6 safety significance of steam I generator tube collapse during a cold leg break LOCA. The effect of flow area l reduction from combined LOCA and SSE loads was estimated. The magnitude of the flow area reduction was considered equivalent to an increased level of steam generator tube plugging. Typically, the area reduction was estimated to range from 0 to 7.5%, depending on the magnitude of the seismic loads. Since detailed non-linear seismir: analyses are not available for Series 51 and earlier design steam generators, some area reductions had to be estimated based on available information. For most of these plants, a 5 percent flow area reduction was assumed to occur in each steam generator as a result of the SSE. For these evaluations, the contribution of loadings at the tube support plates from the LOCA cold leg break was assumed negligible, since the additional area reduction, d.f it occurred, would occur only in the broken loop steam generator.

Westinghouse recognizes that, for most plants, as required by GDC 2, " Design Basis for Protection against Natural Phenomena", that steam generators must be able to withstand the effects of combined LOCA + SSE loadings and continue to perform their intended safety function. It is judged that this requirement applies to undegraded as well as locally degraded steam generator tubes.

Compliance with GOC 2 is addressed below for both conditions.

For tubes which have not experienced cracking at the tube support plate elevations, it is Westinghouse's engineering judgment that the calculation of steam generator tube deformation or collapse as a result of the combination of LOCA loads with SSE loads does not conflict with the requirements of GDC 2.

l During a large break LOCA, the intended safety functions of the steam l generator tubes are to provide a flow path for the venting of steam generated in the core through the RCS pipe break and to provide a flow path such that the other plant systems can perfom their intended safety functions in mitigating the LOCA event.

Tube deformation has the same effect on the LOCA event as the plugging of

( steam generator tubes. The effect of tube deformation and/or collapse can be taken into account by assigning an appropriate PCT penalty, or accounting for the area reduction directly in the analysis. Evaluations completed to date show that tube deformation results in acceptable LOCA PCT. From a steam generator structural integrity perspective,Section III of the ASME Code recognizes that inelastic deformation can occur for faulted condition i

loadings. There are no requirements that equate steam generator tube l- deformation, per se, with loss of safety function. Cross-sectional bending stresses in the tubes at the tube support plate elevations are considered secondary stresses within the definitions of the ASME Code and need not be Et\ECCS EM\tM 91 As 20

considered in establishing the limits for allowable steam generator tube wall degradation. Therefore, for undegraded tubes, for the expected degree of flow area reduction, and despite the calculation shewing potential tube collapso for a limited number -t tubes, the steam ger.erators continue to perform their required safety functions after the combination of LOCA + SSE loads, meeting the requirements of GDC 2.

During a November 7, 1990 meeting with a utility and the NRC staff on this subject, a concern was raised that tubes with partial wall cracks at the tube suppo.t plate elevations could progress to through-wall cracks during tube deformation. This may result in the potential for significant i secondary to primary inleakage during a LOCA event; it was noted that inleakage is not addressed in the' existing ECCS analysis. Westinghouse did not consider the potential for secondary to primary inleakage during resolution of the steam generator tube collapse item. This is a relatively new item, not previously addressed, since cracking at the tube support plate elevations had been insignificant in the early 1980's when the tube collapse item was evaluated in depth. There is ample data  !

l available which demonstrates that undegraded tubes maintain their integrity under collapse loads. There is also some data which shows that cracked tubes do not behave significantly differently from uncracked tubes when collapse loads are applied. However, cracked tube data is available only for round or slightly ovalized tubes. )

l it is important to recognize that the core melt frequency resulting from a i combined LOCA+SSEevent,subsequenttubecollapse,andsiggificantsteam generator tube inleakage is very low, on the order of 10' /RY or less.

)

This estimate takes into account such factors as the possibility of a seismically induced LOCA, the expected occurrence of cracking in a tube as i a function of height in the steam generator tube bundle, the localized effect of the tube support plate deformation, and the possibility that a tube which is identified to deform during LOCA + SSE loadings would also contain a partial through-wall crack which would result in significant inleakage. To further reduce the likelihood that cracked tubes would be subjected to collapse loads, eddy current inspection requirements can be established. The inspection plan would reduce the potential for the presence of cracking in the regions of the tube support plate elevations naar wedges that are most susceptible to collapse which may then lead to oenetration of the primary pressure boundary and significant inleakage during a LOCA + SSE event.

Chanoe Description

/is noted above, detailed analyses which provide an estimate of the degree o f. flow area reduction due to both seismic and LOCA forces are not available for all steam generators. The information that does exist indicales that the flow area reduction may range from 0 to 7.5 percent, I depending on the magnitude of the postulated forces, and accounting for l uncertainties. It is difficult to estimate the flow area reduction for a particular steam generator design, based on the results of a different design, due to the differences in the design and materials used for the tube support plates.

E \ECcF EM\EN 91 A5 21

While a specific flow area reduction has not been determined for some earlier design steam generators, the risk associated with flow area redaction and tube leakage from a combined seismic and LOCA event has been shown to be exceedingly low. Based on this low risk, it is considered adequate to assume, for those plants which do not have a detailed analysis, that 5 percent of the tubes are susceptible to deformation.

The effect of potential steam generator area reduction on the cold leg break LOCA peak cladding temperature has been either analyzed or estimated for each Westinghouse plant. A value of 5 percent area reduction has been applied, unless a detailed non-linear analysis is available. The effect of tube deformation and/or collapse will be taken into account by allocating. the appropriate PCT margin, or by representing the area r reduction by assuming additional tube plugging in the analysis.

Affected Evaluation Models:

1978 Large Break ECCS Evaluation Model 1981 Large Break ECCS Evaluation Model 1981 Large Break ECCS Evaluation Model with BART 1981 Large Break ECCS Evaluation Model with BASH Status:  !

Complete.

3.4 B40 KEN LOOP SAFETY INJECTION FLOW IN SMALL BREAKS Backcround:

In the Westinghouse NOTRUMP small break Evalur. tion Model, it is assumed that the safety injection water which flows to the loop in which the break ,

is postulated to occur _is entirely discharged to the containment. The practice of not taking credit for safety injection into the broken loop preceded the development of calculational models used to satisfy the requirements of 10CFR50.46 or the older Interim Acceptance Criteria (IAC).

It was assumed that neglecting safety injection flow to the broken loop would reduce- the capability for core cooling because the flow would not contribute to the reactor coolant system inventory. It was also assumed that the interactive _ effects on the break flow and the condensation of steam would overall result in better core cooling. The basis for these assumptions was questioned. ,

The spatial representation of the reactor coolant system, the representation of safety irdection flow into the intact loops, and the model for the calculation of the amount of steam condensation as a result of interaction with the safety injection water were selected for conservatism _ in the Westinghouse 1985 small break LOCA Evaluation Model.

This modelL was reviewed in detail by the NRC and approved. This model, however, is not appropriate for evaluating the effects of safety injection E \ECCS EM\EM 91 A6 22

l l

flow into the broken loop due to the interactive effects of the safety injection fluid with the break flow and steam condensation. To evaluate the effect of safety injection flow into the broken loop, a change was made tn the ECCS Evaluation Model to provide a more appropriate representation of the interaction of the safety injection fluid with steam in the RCS. The revised model for condensation of steam due to interaction with the safety injection (SI) fluid was developed based upon test data obtained from the COSI test facility. (The COSI test facility is a

1/100 scale representation of the cold leg and 51 injection ports in a W PWR). The revised steam-SI condensation model was incorporated into a modified version of the Evaluation Model and analysis calculations were performed for a typical three-loop pi, ant.

Analysis calculations which included safety injection flow into the broken loop with the more appropriate revised steam-SI condensation model showed a 54'F benefit over the current model analysis calculation, in which S1 into the broken loop is not modeled. However, an increase in PCT was noted when 51 was modeled in the broken loop with the revised steam-SI condensation model, when compared to the revised steam-SI condensation model case without 51 injection (see summary of results below). Although incorporation of safety injection flow into the broken loop shows a penalty on the peak cladding temperature calculation, it is Westinghouse judgement that the penalty results from the required models of Appendix K to 10CFR50 regarding break flow for the existing spatial representation of the RCS. It is Westinghouse judgement that the actual system response to a small oreak LOCA event would demonstrate that inclusion of safety injection flow into the loop containing the break would mitigate the consequences of the event to a greater extent than if safety injection flow to the loop containing the break was not delivered to the reactor coolant system.

Westinghouse concluded that the practice of neglecting safety injection ficw into the broken loop in combination with a conservative condensation model as in the current version of the Westinghouse 1985 small break LOCA Evaluation Model is conservative and in compliance with the regulatory requirements. Therefore a model change is unnecessary. In order to reach this conclusion, however, the Evaluation Model was changed for application to th S analysis scenario.

SUMMARY

OF RESULTS PCT'F Current model without safety injection into the ' ' en loop 2037 Revised model with safety injection into the bi xen loop 1983 Revised model without safety injection into the broken loop 1806 While no change to- the Evaluation Model is contemplated as a result of this evaluation, it is possible to view the effect of safety injection flow into the broken loop as significant, since the revised steam-SI condensation model significantly reduces the calculated PCT overall. in accordance with 10CFR50 Appendix K, 11.3:

" Appropriate sensistivity studies shall be performed for each evaluation model to evaluate the effect on the calculated results of variations in noding phenomena assumed to predominate, including pump operation or locking, and values of parameters over their applicable ranges. For items to which the results are shown to be sensitive, the l

choices made shall be justified."

E:\ECCS*EM\[M 914 23

. I The existing model is justified as adequately conservative under the requirements of Appendix This was discussed K towith 10CFR50 and willoftherefore not be revised.

meeting on January 22. informally 1991. representatives the NRC staff at a Chanae Desgjotion:

Upon evaluation, it was de Evaluation Model was necessary.termined that no change tc the ECCS Affected Evaluation Models:

1985 Small Break LOCA Evaluation,Model Ltit2}i Complete l

l l

E \tcct EM\EM 91 M 24 i

,mm . .--

4.0 REFERENCES

1. " SATAN-VI Progra: Comprehensive Space Time Analysis of loss of-Coolant",

WCAP-8306 (Nan-Proprietary), June 1974.

2. " Calculational Model for Core Reflooding after a Loss of Coolant Accident (WREFLOOD Code)", WCAP-8171 (Non-Proprietary), J ne 1974.

3. "LOCTA-IV Program: Loss-of-Coolant Transient Analysis", WCAP-8305, (Non-Proprietary), June 1974.

4. " Containment Pressure Analysis Code (C0CO)", WCAP-6326 (Non-Preprietary),

June 1974.

5. "Long Term Ice Condenser Containment Code -

Lotic Code", WCAP-8355 (Non-Proprietary), July 1974.

6. " Westinghouse ECCS Evaluation Model: 1981 Version," WCAP-9221-A, Revision 1, (Non-Proprietary).

7. "BART-Al; A Computer Code for the Best Estimate Analysis of Reflood

' Transients", WCAP-9695-A (Non-Proprit'ary), March 1984.

8. "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", WCAP-Il524-A (Non-Proprietary), March 1987.

9. " Westinghouse large Break LOCA Best Estimate Methodology", WCAP-12130-A, (Non-Proprietary), Vols. 1, 2, December 1988.

10. "WFLASH -

A Fortran IV Computer Program for Simulation of Transients in a Multi-Loop PWR", WCAP-8261-A (Non-Proprietary).

1 1 11. " Westinghouse Emergency Core Cooling System Small Break October 1975 Model", WCAP-8971-A (Non-Proprietary).

12. "NOTRUMP: A Nodal Transient Small dreak and General Network Code",

WCAP-10080-A (Non-Proprietary).

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

WCAP-10081-A (Non-Proprietary).

14. " Westinghouse large Break LOCA Best Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions," WCAP-12130, Revision 2, Volume 1, Addendum 4, (Non-Proprietary), Nissley. M. E., et al, August 1990.

l-

15. "10CFR50.46 Annual Notification for 1989 of Modifications in Westinghouse ECCS Evaluation Models," NS-NRC-89-3633, Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC), Dated October 5,1988.

16. "Large Break LOCA Power Distribution Methodology", WCAP-12935 (Non-Proprietary), May 1991.

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