ML20063E712

Provides Rept Per 10CFR50.46 as Notification of Changes or Errors in ECCS Evaluation Models

ML20063E712
Person / Time
Site:	Beaver Valley
Issue date:	02/02/1994
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9402100206
Download: ML20063E712 (34)

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Beaver Valley Power Station Shippingport, PA 16077-0004 JOHN O SIEBER (412) 393 5. 4 Senior Vice President and Fax (412) 643-806.

$"da"rN$r$N.*[on February 2, 1994 U.

S.

Nuclear Regulatory Commission Attn:

Document Control Desk Washington, DC 20555 i

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPP-73 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models This report is provided as notification of changes or errors inL the BVPS-1 and BVPS-2 Small Break LOCA Evaluation Models which are

)

reportable within 30 days.

Current information for the Large Break

. portions of the LOCA evaluation model has been included to satisfy.

annual reporting requirements.

The following attachments provide information as requested by 10 CFR 50.46:

Provides a

listing of each change or error 11n an acceptable evaluation model that affects the peak:

fuel cladding temperature (PCT) calculation' for particular transients.

It quantifies the effect of changes with respect to potential plant-specific impact on PCT for that transient and provides.an.

"index" into Attachment 2 (Descriptions).

Provides a

description (based

.on-information.

I provided by Westinghouse).for each model change-or error.

Provides a

list of references which occur in the i

various descriptions.

These documents have already been provided to the NRC by. Westinghouse.

The PCT effects, described in Attachment 1, have been~ applied'as penalties to the appropriate PCT calculations.

.This results in calculated PCTs for the large and small break LOCA-transients as follows:

BVPS-1 Large Break LOCA 2122*F BVPS-1 Small Break LOCA - 1834*F l

BVPS-2 Large Break LOCA - 2166*F

' - - v BVPS-2 Small Break LOCA - 2163*F

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THEYEAR 1

9402100206 940202 PDR ADOCK 05000334 O

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1 10 CER 50'.46 Report of Changes or Errors in EDCS Evaluation Models Page 2 1

1 Subsequent to our previous report (March 23, 1993), new analyses have been performed for both BVPS-1 and BVPS-2 Large and Small Break transients.

These analyses resulted in substantial reduction in calculated PCTs for Small Break LOCAs.

Since the BVPS-2 analysis information is not yet complete enough to support this 30-day report, this report reflects the status of changes and errors prior to taking credit for the new analysis.

Because the BASH code is now being used for both

units, future BVPS-2 reports will closely parallel BVPS-1 reports.

l With regard to the errors identified in Attachment 1 for the BVPS-1 Small Break LOCA

analysis, no re-analysis is planned.

The NOTRUMP drift flux flow regime errors have been shown (by actual code runs for a

conservative range of plants) to result only in PCT reductions if corrected.

In addition, Westi..ghouse indicates that there is adequate technical basis, substantiated by prototype test

data, to support their conclusion that a combination of the " Safety Injection In the Broken Loop" error with compensatory refinement of condensation modeling would likewise reduce PCT if implemented (Reference 13 of Attachment 3).

Since none of the changes or errors affecting the BVPS-1 Small Break LOCA analysis have the potential for making Small Break LOCA more limiting than Large Break LOCA or for exceeding the 2200 limit, no further action is currently planned.

Any questions pertaining to this subject may be directed to G.

L.

Beatty at (412) 393-5225.

j Sincerely,

'b b'i, Aulu.

0

/J.

D.

Sieber

//

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cc:

Mr.

L.

W.

Rossbach, Sr. Resident Inspector Mr.

T.

T.

Martin, NRC Region I Administrator Mr.

G.

E.

Edison, Project Manager M

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lh'OFR'56.46ReportofChangesorErrorsin ECCS Evaluation Models

'Page 3 1

bec:

Mr. M. L. Bowling (VEPCO) i

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ATTACHMENT 1

SUMMARY

OF PCT EFFECTS FOR BVPS LOCA TRANSIENTS Plant-Transient PCT Effect (*F)

Description (Pace)

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'BVPS-1 Large Break (Note 4)

II (Page 2)

(Note 4)

V (Page.10) 1 IX (Page 14)

O XI (Page 18)

-25 XVII (Page 23)

(Note 4)

XXII (Page 27)

BVPS-1 Small Break (Note 1)

IV (Page 8)

(Note 4)

V (Page 10)

(Note 4)

VI (Page 12) 0 XVI (Page 22)

(Note 4)

XX (Page 25)

-13 to -55 XXI (Page 26) i 150 XXIII (Page 28)

-150 XXIII (Page 28)

BVPS-2 Large Break 0

I (Page 1)~

s 25 V

(Page 10) 30 IX (Page 14) 0 XI (Page 18)

't (Note 2)

XIII (Page 19)

-25 XVII (Page 23) 0 XVIII (Page 24)

BVPS-2 Small Break (Note 1)

III.A (Page 4) 37 III.B (Page 4)

O III.C (Page 5) 0 III.D (Page 5)

O III.E (Page 5)

+

5 III.F (Page 6)

(Note 3)

III.G (Page 6) l (Note 1)

III.H (Page 6)

O IV (Page 8)

'i 37 V

(Page 10) 20

_ VI _

(Page 12)

O VII (Page 13)-

32 X

(PageL17) 20 XIII (Page_19)'

17 XIV (Page 20)-

j 0

XV '(Page 21).

-13 to -55 XXI (Page 26) 150 XXIII (Page 28)

-150 XXIII (Page 28)

Note 1:

Results in a small' PCT reduction if corrected.

l Note 2:

Results in an unquantified PCT reduction if corrected.

Note 3:

Under investigation by the NSSS vendor.

Note 4:

The specific impact on PCT for this item is no longer being estimated because a re-analysis has incorporated the change.

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A*ITACHMENT 2 TABI.E OF CONTENTS I

1.

MODIFICATIONS TO THE WREFLOOD COMPUTER CODE.......................................... 1 A

II. M ODIFICATIONS TO THE BASH ECCS EVALUATION MODEL...................................... 2.

111. MODIFICATIONS TO THE NOTRUMP SM ALL BREAK LOCA EVALUATION MODEL.'............

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IV. MODIFICATIONS TO THE SMALL BREAK LOCTA-IV COMPUTER CODE............................ 8 V.

FU EL R O D M OD EL REVISION S................................................................................ 10 VI. SMALL BREAK LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION........12 t

VII. NOTRUMP CODE SOLUTION CONVERGENCE...

.......................13 IX. STEAM GENERATOR FLOW AREA.......

........................14 X.

AUXILIARY FEEDWATER ENTHALPY SWITCHOVER FOR SBLOCA ANALYSIS..................17 i

XI. POWER SH APE SENSITIVITY MODEL (PSSM).......................................................... I 8 5

XIll. RCS TEM PERATURE DISTRIB UTION...................................................................... 19.

XIV. RCCA GUlDE THIMBLE AREA.........

...........................................................20 XV. AUXILIARY FEEDWATER FLOW TABLE ERROR....

..........................................21 XVI. STEAM GENERATOR SECONDARY SIDE MODELING ENHANCEMENTS........................... 22 3

XVll. STRUCTURAL M ETAL H E AT M ODELING............................................................... 23 XVHl. SPACER GRID HEAT TRANSFER ERROR IN B ART....................................................... 24 XX. BVPS-1 TURBINE DRIVEN AFW PUMP ACTUATION......................................

............ 25 L f

XXI. NOTRUMP DRIFT FLUX FLOW REGIME M AP ERRORS.,............................................. 26 t

XX11. BOL ROD INTERNAL PRESSURE UNCERTAINTY..................................................... 27 i

XX111. S AFETY INJECTION IN THE BROKEN LOOP............................................................. 281 i

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1. MODIFICATIONS TO THE WREFLOOD COMPUTER CODE

Background:

A modification was made to delay downcomer overfilling. The delay corresponds to backfilling of the intact cold i

legs. Data from tests simulating cold leg injection during the post-large break LOCA reflood phase which gave L

adequate safety injection flow to condense all of the available steam flow show a significant amount of subcooled liquid to be present in the cold leg pipe test section. This situation corresponds to the so-called maximum safety injection scenario of ECCS Evaluation Model analyses.

For maximum safety injection scenarios, the reflooding models in the Westinghouse 1981 ECCS Evaluation Model, i.

the Westinghouse 1981 ECCS Evaluation Model incorporating the BART analysis technology, and the Westinghouse 1981 ECCS Evaluation Model incorporating the BASH analysis technology use WREFLOOD, code it versions which predict the downcomer to overfill. Flow through the vessel side of the break is computed based l

upon the available head of water in the downcomer in WREFLOOD using an incompressible flow'in an open channel method. A modification to the WREFLOOD computer code was made to consider the cold leg inventory which would be present in conjunction with the enhanced downcomer level in the non-faulted loops.

Change

Description:

WREFLOOD code logic was altered to consider the filling of the cold legs together with downcomer overfilling.

Under this coding update, when the downcomer level exceeds its maximum value as input to WREFLOOD, liquid flow into the intact cold leg, as well as spillage out the break, is considered. This logic modification stabilizes the overfilling of the vessel downcomer as it approaches it equilibrium level. The appropriate 'WREFLOOD code versions associated with the 1981 Westinghouse ECCS Evaluation Model and the 1981 Westinghouse ECCS Evaluation Model incorporating the BART and basil technology have been modified to incorporate the downcomer overfill logic update.

Effect of Change:

This change represents a model enhancement in terms of the consistency of the approach in the WREFLOOD code -

and the actual response of the downcomer level in some cases this change could delay the overfilling process, which could result in a peak cladding temperature (PCT) penalty. The magnitude of the possible PCT penalty was :

assessed by reanalyzing the plant which is maximum safeguards limited (CD=0.6 DECLG case) and which is most sensitive to the changes in the WREFLOOD code. The PCT penalty of 16*F which resulted for this case represents the maximum PCT penalty which could be exhibited for any plant due to the WREFLOOD logic change.

Westinghouse has determined, for BVPS-2, that the effect of this change on Large Break LOCA PCT is O'F.

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11. MODIFICATIONS TO THE BASH ECCS EVALUATION Af 0 DEL

Background:

In the basil ECCS Evaluation Model (reference 3), the BART core model is coupled with equilibrium-NOTRUMP computer code to calculate the dynamic interaction between the core thermohydraulics and system behavior in the reactor coolant system during core reflood. The BASH code reflood model replaces the WREFLOOD calculation to produce a more dynamic flooding transient which reflects the close coupling between core thermohydraulic and-hmp behavior. Special treatment of the BASH computer code outputs ir used to provide the core flooding rate for use in the LOCBART computer code. The LOCBART computer code results from the direct coupling of the BART-computer code and the LOCTA computer code to directly calculate the peak cladding temperature.

Change

Description:

Modifications to the BASH ECCS Evaluation Model include the modifications made to the 1981 ECCS Evaluation Model, discussed previously, and the following previously unreported modifications; Several improvements were made to the BASH computer code to treat special analysis cases which are related to the tracking of fluid interfaces;

1) A modification, to prevent the code from aborting, was made to the heat transfer model for the special situation.

when the quench front region moves to the bottom the BASH core channel. The quench heat supplied to the fluid node below the bottom of the active fuel was set to zero.

2) A modification, to prevent the code from aborting, was made to allow negative initial movement of the liquid /two-phase and liquid-vapor interfaces. The coding these areas was generalized to prevent mass -

imbalance in the special case where the liquid /two-phase interface reaches the bottom of the BASH core channel.

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

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

5) Typographical 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 leg conditions, in a conservative manner, when the accumulators empty. The BASH model is initialized at the bottom of core recovery with the intact cold legs, 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.

~i in a typical basil calculation, the downcomer is nearly full when the accumulators emptied. The delay time, q

prior to the intact cold leg water reaching saturation, is sufficient to allow the downcomer to fill from the~

addition of safety injection fluid before the water in the cold legs reaches saturation. When the intact cold leg l

water reached saturation it merely flowed out of the break. The cold leg water therefore, did not affect the reflood transient.

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

Effect of Change:

For typical calculations, there is no effect on the PCT calculation for the majority of the changes discussed above.

A conservative estimate of the effect of the modifications on the calculations was determined to be less than 10*F, singly or in combination.

For BVPS-1, the effect of this change is no longer being estimated because a reanalysis has incorporated the -

change.

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111. MODIFICATIONS TO THE NOTRUMP SM ALL BREAK LOCA EVALUATION MODEL

Background:

The NOTRUMP small break LOCA ECCS Evaluation Model (Reference 4) was developed by Westinghouse in cooperation with the Westinghouse Owners Group to address technical issues expressed in NUREG-0611 "Small Break LOCA and Feedwater Transients in W PWRs," in compliance with the requirements of NUREG-0737,

" Implementation of the TMI Action Plan," Section ll.K.3.30. In the NOTRUMP small break LOCA ECCS Evaluation model, the NOTRUMP code is used to calculate the thermal-hydraulic response of the reactor coolant system during a small break LOCA and the SBLOCTA-IV computer program is used to calculated the performance of fuel rods in the hot assembly.

Several modifications have been made to the NOTRUMP computer (Reference 1) to correct erroneous coding or improve the coding logic to preclude erroneous calculations. The modifications indicated in A through I below have been incorporated into the production version of the code. Remaining corrections and modifications are not significant and will be incorporated during the next code update in accordance with the Westinghouse quality assurance procedures for computer code maintenance. The following modifications to the NOTRUMP small break LOCA ECCS Evaluation Model have been made:

A. Change

Description:

A modification was made to preclude changing the region designation (upper, lower) for a node in a stack.

which does not contain the mixture-vapor interface. The purpose of the modification was to enhance tracking of the mixture-vapor interface in a stacked series of fluid nodes and to preclude a node in a stack, which does not contain the mixture-vapor interface, from changing the region designation.

The update does not affect the fluid conditions in the node, only the designation of the region of the node. The region designation does not typically affect the calculations, except for the nodes representing the core fluid volume (core nodes). In core nodes which are designated as containing vapor regions, the use of the steam.

cooling heat transfer correlation is forced on the calculation in compliance with the requirements of Appendix -

K to 10CFR50, even if the node conditions would indicate otherwise. The use of the steam cooling heat transfer regime above the mixture level is documented on page 3-1 of reference 2.

Effect of Change:

In rare instances, an incorrect heat transfer correlation could be selected if the region designation was -

improperly reflected. An analysis calculation was performed for a three-loop plant which resulted in a decrease.

in the PCT of 6.5'F when the corrections were made for a calculation which would be affected by the change.

Westinghouse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is a small, unquantifed reduction.

B. Change

Description:

Typographical errors in the equations which calculate the heat transfer rate derivatives for subcooled, saturated, and superheated natural convection conditions for the upper region ofinterior fluid nodes were corrected. The heat transfer rate derivatives for subcooled, saturated, and superheated natural convection conditions for the upper region of interior fluid nodes are given by equations 6-55, 6-56, and 6-57 of reference 2.

A.

typographical error led to the use of the lower region heat transfer area instead of the upper region heat transfer area in the calculation of derivatives. The error affected only the upper region heat transfer derivatives which 4

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are used by the code to characterize the implicit coupling of the heat rates to changes in the independent nodal variables.

Effect of Change:

-1' In rare instances, the amount of heat that could be transferred to the fluid could be improperly calculated. The effect of the errors was expected to be small since the error would only affect the derivatives of the heat rates for vapor regions that are in natural convection. An analysis calculation was performed for a three-loop plant which resulted in a larger than expected increase in the PCT of 36.7'F when the correction was made on a

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calculation which would be affected by the change.

Westinghouse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is 37'F.

C. Change

Description:

Typographical errors in equations which calculate the derivatives of the natural convection mode of heat transfer in the subroutine HEAT were corrected. A conductivity term used in the equations which calculate the derivatives of the natural convection mode of heat transfer was incorrectly typed as CK (to be used for the

'Itom or McBeth correlations), instead of CKNC (to be used for the desired McAdams correlation).

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Effect of Change:

A review of the code logic was performed to assess the effect of the error. In all equitions thr.t contain the typographical error, the incorrect variable is multiplied by zero. Therefore the typographic:d errors have no

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1 D. Change

Description:

1 A typographical error was corrected in an equation which calculates the intemal energy for nodes associated

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with the reactor coolant pump model when the associated reactor coolant pump flow links are found to be in critical flow. An incorrect value for the mixture region internal energy in the fluid node downstream of a

.j pump flow link would be calculated if the pump flow link were in critical flow.

1 Effect of Change:

This section of coding is not expected to be executed for small break LOCA Evaluation Model calculations since critical flow in the reactor coolant pump flow links does not occur. Therefore this modification has no effect on the calculations. This was confirmed in an analysis calculation for a three-loop plant which -

demonstrated no change to the PCT.

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E. Change

Description:

A modification was made to properly call some doubly dimensioned variables in subroutines INIT and TRANSNT. Some variables are doubly dimensioned (X,Y) but were being used as if they were singly dimensioned.

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Effect of Change:

A detailed review of the code logic indicated that all of the doubly dimensioned variables had 1 as the second dimension in any of the erroneous calls. The computer inferred a 1 for the second dimension in the iniproper subroutine calls. Therefore, there is no effect of this modification on the PCT.

F. Change

Description:

A modification was made to prevent code aborts resulting from implementation of a new FORTRAN compiler.

Due to the different treatments of the precision of numbers between the FORTRAN compilers, the subtraction of two large, but close numbers resulted in exactly zero. The zero value was used in the denominator of a derivative equation, which resulted in the code aborts. This situation only occurred when the mass of a region in a node approached, but was not equal to zero.

1 Effect of Change:

An analysis calculation was performed for a four-loop plant which resulted in a larger than expected increase in the PCT of 4.8'F when the modification was implemented.

Westinghouse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is 5'F.

G. Change

Description:

An error in the implementation of equation 5-33 of reference 2 was corrected. Equations 5-33 describes the calculation of the flow link friction parameter ek for single phase flow in a non-critical flow link k. In the erroneous implementation, equation 5-33 was replaced by equation 5-34 which is used for all flow conditions.

For the case where the flow quality is zero, equation 5-34 is similar In form to equation 5-33 since the two-phase friction multipliers are exactly unity when the flow quality is zero and the donor cell and flow link fluids are saturated, equations 5-33 and.5-34 are equivalent. However, for subcooled flow the flow link specific volume vk n equation 5-33 is not equivalent to the saturated fluid donor cell specific volume (vk. donor (k)) in i

equation 5-34.

Effect of Change:

This modification was expected to have only a small beneficial effect on the analysis. However, an analysis calculation was performed for a three-loop plant to quantify the effect and a larger than expected decrease in the peak cladding temperature of 217*F resulted. Larger than expected peak cladding temperature sensitivities, in some instances, have been observed when analyses to support safety evaluations of the effect of plant design changes under 10CFR50.59 were performed using the NOTRUMP computer code. The unexpected sensitivity results are under investigation at Westinghouse and may be due to the artificial restrictions on loop seal steam venting placed on the model for conservatism. Evaluations of the effect of this change will be examined as part of the investigation of the larger than expected sensitivity results.

IL Change

Description:

A modification was made to correct an error in implementing equations L-28, L-52 aci L-29, L-53 of reference k

2. The two pairs of equations respectively describe the Partial derivatives of F with respect to pressure and k

specific enthalpy. F is an interpolation parameter that is defined by equations L-27, lAl of reference 2. In each pair the lower equation number is for the subcooled condition, and the higher equation number is for the 6

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superheated condition The denominator of each equation contams the differences between h and h -1 where hk k

k k

is defined by equations L-21, L-45 and h -1 is defined by equations L-22, L-46 of reference 2. Although the k

k expression defining h and h -1 were correctly calculated in NOTRUMP, they were not used in equations L-28,1<52 and L-29, L-53 as they should have been.

Effect of Change:

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An analysis calculation was performed for a four-loop plant which resulted in a decrease in the PCT of 12.8'F when the modification was made for a calculation which would be affected.

Westinghouse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is a small, unquantified reduction.

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IV. MODIFICATIONS TO THE SMALL BREAK LOCTA-IV COMPUTER CODE I

The following modifications to the LOCTA-IV computer code in the small break LOCA ECCS Evaluation Model -

have been made:

A. Change

Description:

A test was added in the tcf-to-steam radiation heat transfer coefficient calculation to prechade 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, b

Effect of Change:

There is ~i 'Tect of the modification on reported PCTs since the ernmeous use of the correlation forced the calculath.-

1.a aborted conditions.

B. Change

Description:

i An update was performed to allow the use of fuel rod performance data from the revised Westinghouse (PAD l

3.3) model.

Effect of Change:

1 An evaluation indicated that there is an insigni'icant effect of the modification on reported PCTs.

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C. Change Description-l Modifications supporting a general upgrade of the computer program were impleme, follows:

1) the removal of unuml or redundant coding,

2) better roding org mizatica to increase the efficiency of calculations, and

3) improvements in user friendliness a) through defaulting of same input variabl a, lj b) simplification of input, c) input diagno.stic checb, and d) clarification of the output.

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"frect of Change:

% fication analyses calculations demonstrated that there wu no effect on.the calculated output iesulting from 1

these changes.

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D. Change

Description:

j Two modifications improving the consistency betwe:en the Westinghouse fuel rod perfonnsnce data (PAD) and the small break LOCTA-IV fuel rod models were implemented:

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1) The form of the equation for the <lensity of uranium-dioxide in the specific heat correlation, which modeled three dimensional expansion was corrected to account for only two-dimensional thermal expansion due to

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the way the fuel rod is modele e calculated peak cladding temperature (PCT), calculations were performed which incorporated the changes, i:A. og the cladding strain model correction -

for the large break LOCA. For the large break LOCA Evaluation Model, additional calculations, incorporating l

only the cladding strain corrections were perfctmed and the results supported the conclusion that compensating I

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 large break LOCA Evaluation Model The effects werejudgxi applicable to older evaluation models. Several calculations were performed to assess the effect of the changes on the calculated results as follows:

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Blowdown Analysis -

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lt 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 q

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ua analysis was (itermined by performing a blowdown depressurization computer calculation for a typical i

' three-loop planc and a typical four-loop plant using the SATAN VI computer code.

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Hot Assembly Rod Heatup Analysis -

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f The hot rod heatup calculations would typically show the largest effect of the changes. Hot rod heatup j

computer analysis calculations were performed using the LOCBART computer code to assess the effect of -

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the changes on the hot assembly average rod, hot rod and adjacent rod.

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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.

i calculations confirmed that the effect of the ECCS Evaluation Model changes were insignificant as defined.

t by 10CFR5146(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 clad ('mg temperature for the BART Evaluation Model, while calculations performcJ for the '

Westinghotae 1981 Evaluation Model showed that the peak cladding temperature could. increase by

'l approximately 41'F.

Westinghouse has determined, for BVPS-2, that the effect of this change on Large Break LOCA PCT is 25"F. The effect of this change is no longer being estimated for BVPS-1 because a reanalysis has.

incorporated the change.

b) Small Dreak LOCA The effect of the changes on the small break LOCA analysis peak cladding temperature calculations was determined using the 1985 small break LOCA Evaluation Model by performing. a computer analysis 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(n)(3)(i). The calculations showed that 37'F would bound the effect on the calculated peak cladding temperstures for the four-loop plants and the three-loop plants. It wasjudged that an increase of 37*P would bound the effect of the changes for the 2-h>op plants.

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Westinghocse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is 37'F.

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The effect of this change is no longer being estimated for BVPS-1 because a reanalysis has incorporated the

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VL SM ALL BREAK'LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION I

Change

Description:

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~ The Westinghouse small break loss-of-coolant accident (LOCA) emergency core cooling system (ECCS) Evaluation

-l Model analyses assume that higher fuel rod initial fil pressure leads to a higher calculated peak cladding i

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 i

fuel rod initial fill pressure could then result in a higher calculated peak cladding temperature, j

The Westinghouse small break LOCA cladding strain model is based upon a correlation of Hardy's data, as f

described in Section 3.5.1 of Reference 5. Evaluation of the limiting fuel rod initial fill pressure assumption

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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. 'Ihe model was corrected to lit applicable data over the range of small break LOCA conditions. Correction of the chdding strain model affects

.l the small break LOCA Evaluation Model calculations through the fuel rod internal pressure initial condition auumption.

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Elfeet of Changes:

'j 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 inten.1 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 previously 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 c bl leg break in typical three-kiop and four-loop plants. The results indicated that the change to the calculated peak ladding temperature 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 ease the combined effect could be as large as 60'F.

' Westinghouse is currently evaluating a SBLOCA Burst / Blockage issue which is potentially more limiting than the issue discussed above. The Rod Internal Pressure item is associated with a transient configuration where rod burst does not occur. On the other hand, the Burst / Blockage issue applies if the rod bursts nt the limiting time in life. The j

rod burst causes a rathe sharp PCT spike as both sides of the clad react with water. Since a rod cannot both burst:

and not burst, the higher PCT penalty from either scenario is applied to PCT. The Burst / Blockage evaluation technique is PCT dependent. in an exponential fashion due to the dependence of the Zirc-water reaction.on clad temperature and therefore is derived after inclusion of other changes and error estimates affecting the model.

Westinghouse had determined, for BVPS 2, that the efTect of this change on Small Break LOCA PCT is 20 F. The -

clTect of this change is no longer being estimated for BVPS-1 because a reanalysis has incorporated the change.

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l 1-VII.. NOTRUMP CODE SOLUTION CONVERGENCE i

1 Change

Description:

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s In the development of the NOTRUMP small break LOCA ECCS Evaluation Model, a number of noding sensitivity.

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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 o

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studies, modifications were made to _ the NOTRUMP computer program to enhance code performance and implement necessary modifications (Reference 7). Subsequent to the modifications, solution convergence was not-re-confirmed.

To analyze changes in plant operating conditions, sensitivity studies were perfs ed 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 _

~i judgment. In addition, the direction of the PCT variation conflicted with engineering judgment expectations in some cases. The unexpected variability of the sensitivity study results indicated that the numerical solution may not

j be properly converged.

Sens rivity studies were performed for the time step size selection criteria which culminated in a revision to the -

recommended time 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 SUT-03 Semiscale experiment and it was confirmed that the code adequately predicts key small break phenomena, l

Elfeet of Changes:

Generally, the modifications result in small shifts in timing of core uncovery and recovery. Ilowever, these changes may result in a change in the calculated peak cladding temperature which exceeds 50*F for some plants. Based on l

representative calculanons, 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 solu' ion is plant specific, a generic-T V effect cannot be provided. Ilowever, it has been concluded that current licensing basis results remain valid l

since the results are conservative relative to the change.

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Westinghouse has determined, for BVPS-2, that the effect of this change on Small Break LOCA PCT is O'F.

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3 13 1

4

,I IX. STEAM GENERATOR FLOW AREA 4

Background:

Licensees are normally required to provide assurance that there exists only an extremely low probability of

'l aboermal lest. age or gross rupture of any part of the reactor coolant pressure boundary (General Design Criteria 14 ans 31). The NRC issued a regulatory guide (RG 1.121) which addressed this requirement specifically for steam generator tubes in pressurized v ater reactors. in that 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 j

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i the h> ads from a safe shutdown earthquake (SSE). These loads are combined for added conserve calculation of structural integrity. This analysis provides the basis for establishing criteria for renoving from service tubes which had experienced significant degmdation.

l Analyses performed by Westinghouse in support of the above requirement for various utilities, combined the most

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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

.l defonnation. This deformation reduces the flow area through the steam generator. 7&r 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.

1 Change

Description:

The elfeet 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

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1 (bliowing calculated phenomena-1

J 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 wedge supports at the periphery of the plate. The tube support plate deformation may cause tube deformation.

1 3.

During a postulated large LOCA, the primary side depressunzes to containment pressure. Applying the j

resulting press'ure differential to the defonned tubes causes some of these tubes to collapse, and reduces the y

elfective flow area through the steam generator.

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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).

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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 'his break was judged to result in the greatest loads on the steam generator, and thus the greatest flow area reduction. It was concluded that the steam generator would continue to meet 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.

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In April of 1990, in considering the effect of the combination of LOCA + SSE loadings on the steam generator component, it was detennined that the potential for flow area reduction due to the. contribution of SSE loadings; should be included in other LOCA analyses. 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 primary pipe Iveak affecting LOCA PCT, i.e., the reactor vessel inlet break (cold leg break LOCA).

had to be evaluated to confirm that 10CFR50.46 limits continue to be met and that the affected steam generators will con inue to perfonn their intended safety function.

14

4 C<msequently, the 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 k> ads was estimated. The l

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 E depending on the magnitude of the seismic loads. Since detailed non-linear seismic analyses are not available for Series 51 and earlier design steam Fenerators, 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 SSR. For these evaluations, the contribution of loadings at the tube support plates from the LOCA cold leg break was assumed -

i negligible, since the additional area reduction, if 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 perfonn their intended safety function. It is judged that this requirement applies to undegraded as well l'

as locally degraded steam generator tubes. Compliance with GDC 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. During a large break LOCA, the intended safety functions of the steam 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 perform i

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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

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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 Ill of the ASME Code recognizes that inelastic deformation can occur for faulted condition loadings. There are no requirements that eqaate steam generator tube 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

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Code and need not be 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 showing potential tube collapse for a limited number of tubes, the steam generators 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 j

tubes with partial wall cracks at the tube support plate elevations could progress to through-wall cracks during tube '

I deformation. This may result in the potential for significant 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 accondary to primary inteakage during resolution of the steam generator tube collapse item. This is a relatively new item. not previously addressed, since cracking et the tube support plate elevations had been.

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insignificant in the early 1980's when the tube collapse item was evaluated in depth. There is ample data available which demonstrates that undegraded tubes maintain their integrity under collapse loads. 'Diere is also some data.

which shows that cracked tubes do not behave significantly differently from uncracked tubes when collapse loads 1

l-are applied. However, cracked tube data is available only for round or slightly ovalized tubes.

It is important to recognize that the core melt frequency resulting from a combined LOCA + SSE event, subsequent tube collapse, and significant steam generator tube inleakage is very low, on the order of 10-8/RY or less. This i

estimate takes into account such factors as the possibility of a seismically induced LOCA, the expected occurrence l

of cracking in a tube as a function of height in the steam generator tube bundle, the kicalized 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 inleakr.ge, 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 fut the presence of cracking in the regions of the 15 i..

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tube support plate elevations near wedges that are most susceptible to collapse which may then lead to penetration

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of the primary pressure boundary and significant inteakage during a LOCA + SSE event.

1 Change

Description:

As noted above, detailed analyses which provide an estir ste of the degree of flow area reduction due to both seismic and LOCA forces are not available for all steam generators. The information that does exist indicates that j

the flow area reduction may range from 0 to 7.5 percent, depending on the magnitude of the postulated forces, and i

accounting for 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.

While a specific flow area reduction has not been determined for some earlier design steam generators, the risk.

l associated with flow area reduction and tube leakage from a combined seismic and LOCA event has been shown to i

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 aliocating the appropriate PCT margin, or by representing the area reduction by assuming additional tube plugging in the analysis.

Westinghouse has determined, for BVPS-1, that the effect of this change on Large Break LOCA PCT is l'F. For BVPS-2, a 30'F cffect is estimated.

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X. AUXILinRY FEEDWATER ENTilALPY SWITCilOVER FOR SBLOCA ANALYSIS 1

Change

Description:

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During a review of Westinghouse SBLOCA analysis methods, a question arose with respect to the computer code input used to represent the time required for the lower flow, lower enthalpy auxiliary feedwater to purge the higher enthalpy main feedwater from the feedwater piping after actuation of auxiliary feedwater. In the Westinghouse 1

SBLOCA ECCS Evahiation models using either the WFLASil or NOTRUMP analysis technologies, this time is J

used to switch the enthalpy of the fluid provided to the steam generators from the main feedwater enthalpy to the.

auxiliary feedwater enthalpy.

Effect of Change:

A review and investigation of the concern indicated that, in some instances, the time assumed for the auxiliary feedwater enthalpy purge delay time was shorter than times calculated from the actual plant configuration. The inconsistency between the Westinghouse SBLOCA ECCS Evaluation Model input value and a value corresponding i

to the plant configuration results from the specific guidance provided to the analyst for determining the auxiliary feedwater enthalpy delay time. In both the WFLASil and NOTRUMP methods, a standard purge delay time was l

recommended. In the NO't RUMP analysis methodology, a standard input value judged to be conservative based I

upon phenomena observed during experiment SUT-08 in the Semiscale test facility was used. However, further l

investigation showed that the standard input value could result in a non-conservative calculation of the peak cladding temperature. '

l Westinghouse has determined, for BVPS-2, that the cifeet of this change on Small Break LOCA PCT is 32*F.

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A XI POWER SilAPE SENSITIVITY MODEL(PSSM)

Background:

Large Break LOCA analyses have been traditionally performed using a symmetric, chopped cosine axial power shape. Recent calculations have shown that there was a potential for top-skewed power distrib'utions to result in Peak Cladding Temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution.

Westinghouse has developed a procus, which was applied to the reload for Beaver Valley Unit 2 Cycle 4, that ensures that the cosine remains the limiting power distribution, by defining appropriate power distribution surveillance data. This process, called the Power Shape Sensitivity Model (PSSM), is described in a topical report (WCAP 12909-P).

In May,1991, Westinghouse transmitted to the NRC the report titled, " Westinghouse ECCS Evaluation Model:

Revised Large Break LOCA Power Distribution Methodology," which describes the process that Westinghouse is now using to more accurately account for the effect of power distribution in the core reload design. In January, 1991, the implementation of this approach was discussed with the NRC. In a May,1991 meeting with the NRC, Westinghouse again told the NRC that they planned on implementing the PSSM process shortly after the topical report was submitted. Westinghouse indicated in the transmittal letter of the topical (NS-NRC-91-3578) that it was their intent to implement the PSSM process for future reload design applications. The NRC has informally stated that it is acceptable to use revised LOCA methodology that corrects a potential deficiency, until it has been' reviewed by the Staff.

Effect of Change:

Implementation of this methodology has no effect on calculated PCT.

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b XIII. RCS TEMPERATURE DISTRIBUTION Backgrouni

- While evaluating the effects of reduced thermal design flow (TDF) for BVPS-2, anomalies were discovered in the.

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interaction between the RCS temperature distribution methodology and the actual analysis inputs for both SBLOCA I

& LBLOCA analyser Action was taken to investigate and evaluate the anomalies.

A' portion of the apparent discrepancy is attribut to slight changes in the LOCA inputs that result from miscellaneous evolutionary changes to the plant characteristics such as the Steam Generator Fouling Factor, which was recently recalculated.' Other i

differences are attributed to deviations that occurred in the LOCA analyses themselves, such as an extraction error resulting in incorrect LBLOCA input.

- i

1 Effect of Change:

The LBLOCA analysis RCS T,vg inputs are greater than necessary for either current TDF or reduced TDF cases,;

' j and thus the analysis remains bounding, though the benefit is unquantified.

The SBLOCA analysis RCS T inputs are greater than necessary for either current TDF or reduced TDF cases, ayy and thus a PCT penalty of 20*F is assigned. This PCT penalty is considered to be reportable under 10CFR50.46 as

- an error in the application of the Evaluation Model.

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~l T'V. RCCA GUIDE TillMBLE AREA

Background:

The only VANTAGE 511 Zircaloy grid feature which significantly affects the SBLOCA analysis is the increase in

' design rod drop time. The Westinghouse Small Break model assumes the reactor core is brought to a suberitical condition by the negative reactivity of the control rods. The increase in the design rod drop time to a maximum value of 2.7 seconds exceeds the 2.4 second value in the existing SBLOCA analysis.

Effect of Change:

t An evaluation was performed which determined that a 3*F PCT penalty applied for the increase in rod drop time of 0.3 seconds. The decrease in core pressure drop associated with thimble plug removal has an inconsequential effect '

on the S'BLOCA analysis. liowever, the SBLOCA analysis did not model the effect of the guide thimble interior area and volume on the transient response. Studies have shown that the fluid in this volume _will interact with the remaining core fluid. An assessment of this interaction based on previous sensitivities indicated that a 17'F PCT penalty applies to BVPS-2.

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XV. AUX 1LIARY FEEDWATER FLOWTABLE ERROR i

- Change

Description:

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The steam generator auxiliary fcxdwater (AFW) flow rate is governed by the timing variable TIMESG(I). A minor logic error associated with this variable was discovered which led to a step change in the AFW flow rate once the transient time passed the value of TIMESG(7). Typically, this value is set equal to 11000 seconds and so this error.

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would only affeet very long transient calculations. In addition, the nature of the error is to allow the AFW flow rate to immediately revert to the full value of the Main Feedwater flow rate. This enormous step change has led to code aborts in the cases where it has occurred.

1 nis logic was corrected as a " Discretionary Change' as described in Section 4.1.1 of WCAP-13451. This determination is based on the fact that SBLOCA transients are generally terndnated before the logic error can have an e ffect coupled with the code's lack of capability to handle the step change if it does occur. Therefore, it was -

reasoned that the logic could not affect LOCA results.

Effect of Change:

This error correction has no effect on any current or prior applications of the Evaluation Model.

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XVI. STEAM GENERATOR SECONDARY SIDE MODELING ENHANCEMENTS t

Change

Description:

A set of related changes which make steam generator secondary side modeling more convenient for the user were

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implemented into NOTRUMP. This model improvement involved several facets of feedwater flow modeling.

l First, the common donor boundary node for the standard evaluation model nodalization has been separated into two -

j identical boundary nodes. These donor nodes are used to set the feedwater enthalpy. The common donor node configuration did not allow for loop specific enthalpy changeover times in cases where asymmetric AFW flow rates.

or purge volumes were being modeled for plant specific sensitivities.-

The second improvement is the additional capability to initiate main feedwater isolation on either loss of offsite power coincident with reactor trip (Iow pressurizer pressure) or alternatively on safety injection signal low-low -

pressurizer pressure). The previous model allowed this function only on loss of offsite power coincident with '

reactor trip. The auxiliary feedwater pumps are still assumed to start after a loss of offsite power with an appropriate delay time to model diesel generator start-up and buss loading times.

The final improvement is in the area of modeling the purging of high enthalpy main feedwater after auxiliary feedwater is calculated to start. This was previously modeled through an approximate time delay necessary to purgq

.i the lines of the high enthalpy main feedwater before credit could be taken for the much lower enthalpy auxiliary feedwater reaching the steam generator secondary. This time delay was a function of the plant specific purge volume and the auxiliary feedwater flow rate. The new modeling allows the user to input the purge volume r;

directly. This then is used together with the code calculated integrated fectwater flow to determine the appropriate' time at which the feedwater enthalpy can be assumed to change.

These improvements are considered to be a " Discretionary Change" as described in Section 4.1.1 of WCAP-13451.

Since they involve only enhancements to the capabilities and useability of the evahiation model, and not changes to results calculated consistently with the previous model, these changes were implemented without prior review as discussed in Section 4.1.1 of WCAP-13451.

Effect of Change:

l Because these enhancements only allow greater case in modeling plant specific steam generator secondary side '

behavior over the previous model, it is estimated that no effect will be seen in evaluation model calculations, i

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XVII. STRUCTURAL METAL IIEAT MODELING t

Change

Description:

A discrepancy was discovered dunng review of the finite element heat conduction model used in the WREFLOOD-INTERIM code to calculate heat transfer from structural metal in the vessel during the reflood phase. It was noted that the material properties available in the code corresponded to those of stainless steel. While this is correct for the internal structures, it is inappropriate for the vessel wall which consists of carbon steel with a thin stainless i

internal clad. This was defined as a "Non-discretionary Change" p r Section 4.1.2 of WCAP-13451, since there -

was thought to be potential for increased PCT with a more sophisticated composite model. The model was revise 4 j

by replacing it with a more flexible one that allows detailed specification of structures.

Effect of Change:

)

The estimated effect of this correction is a 25'F PCT benefit.

Westinghouse has determined, for both BVPS 1 and BVPS-2, that the effect of this change on Large Break LOCA i

PCT is -25'F.

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'A XVill. SPACER GRID IIEAT TRANSFER ERROR IN BART Change

Description:

i During investigations into anomolous wetting and dryout behavior demonstrated by the BART grid nmdel a programming logic error was discovered in the grid heat transfer model. The error caused the solution to be performed twice for each timestep. The error was traced back to the original coding used in all of the BART and i

LOCBART codes. This was defined as a "Non-discretionary Change" per Section 4.1.2 of WCAP-13451. The' error was corrected, and a complete reverification of the grid model was conducted and transmitted to the NRC -

l (WCAP-10484, Addendum 1).

Elfect of Change:

Calculations performed with the affected code have consistently demonstrated significantly better grid wetting and

'f lower clad temperatures. A conservative estimate of zero degrees PCT penalty has been anigned for this issue.

j Westinghouse has determined, for BVPS-2, that the effect of this chanFe on Large Break LOCA PCT is O'F.

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XX. BVPS-1 TURBINE DRIVEN AFW PUMP ACTUATION

Background:

Westinghouse has performed a Small Break LOCA (SBLOCA) technical evaluation for (1) the removal of credit for L

the Turbine Driven AFW actuation via Loss-of-Offsite Power Undervoltage Relays and (2) as a compensatory measure, installing a SI actuation circuit for the Turbine Driven (TD) AFW pump using a delay time of 60 seconds. '

-l The BVPS-1 SBLOCA analysis of record is a NOTRUMP Evaluation Model analysis which assumes that AFW l

delivery actuates on the combination of Reactor Trip / coincident Loss-of-Offsite Power (LOOP), consistent with the model presented in the NOTRUMP Topical Report. In the new design, the Motor Driven (MD) pumps and-

)

Turbme Driven pump will actuate, with a 60 second delay.

Although two MD and one TD AFW pumps are available, the existing analysis credits only one MD and one TD pump due to the limiting single failure, of one Diesel Generator, which precludes operation of the second MD pump (as well as I train of SI Pumps).

I In the SBLOCA analysis, the Low Pressurizer Pressure (LPP) S1 and LPP Reactor Trip times are both modeled, with the SI signal occurring approximately 5 seconds later due to its lower setpoint. In addition, the analysis assumed a 60 second delay time for AFW pumps actuation. Therefore, crediting AFW operation based on LPP SI signal instead of LPP Reactor Trip / coincident LOOP represents an additional 5 second delay for AFW delivery.

Elfect of Change:

To con plete the evaluation, AFW delivery data and BVPS-1 plant specific AFW reduction sensitivities were utilized. The evaluation concluded that a PCT penalty of 6*F was incurred.

For BVPS-1, the effect of this change is no longer being estimated because a reanalysis has incorporated the change.

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.XXI. NOTRUMP DRIFT FLUX FLOW REGIME MAP ERRORS Change

Description:

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Errors were discovered in both WCAP-10079-P-A and related coding in NOTRUMP subroutine DFCORRS where the improved TRAC-P1 vertical flow regime map is evaluated. This model is only used during counter-current flow j

conditions in vertical flow links. The affected equation in WCAP-10079-P-A is Equation G-65 which previously j

allowed for unbounded values of the parameter C contrary to the intent of the original source of this equation, j

e This allowed a discontinuity to exist in the flow regime map under some circumstances. This was corrected by.

placing an upper limit of 1.3926' on the parameter C as reasoned from the discussion in the original source. As stated, this correction retumed NOTRUMP te consistency with the original source for the affected equation.

Further investigation of the DFCORRS uncovered an additional closely related logic error which ' led to discontinuities under certain other circumstances. This error was also corrected and returned the coding to consistency with WCAP-10079-P-A.

l Effect of Change:

Westinghouse has determined, for both BVPS-1 and BVPS-2, that the effect of this change on Small Break LOCA ^

PCT is -13 to -55'F.

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XXIL BOL ROD INTERNAL PRESSURE UNCERTAINTY

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Change

Description:

Large Break LOCA (LBLOCA) analyses are performed at near Beginning<>f-Life (BOL) fuel rod conditions, which

' have been shown to be limiting in sensitivity studies. The fuel rod performance utilized in the analyses corresponds j

to Rod Internal Pressure (RIP) calculations using NRC approved PAD models which provide RIP that contains some conservatism in its calculations. liigher RIP typically results in earlier and greater rod burst and bk>ckage, and ' ultimately a PCT penalty. Questions have been raised concerning the calculation of BOL RIP uncertainties.

J which contribute to the upper bound BOL RIP utilized in the LBLOCA analy:is. Although this issue is not yet fully evaluated by Westinghouse, the model used for the most recent BVPS-1 analysis incorporates changes which ensure inclusion of its effects.

Effect of Change:

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Using a conservative combination of actual BOL uncertainties results in a current estimate of 65 psi bounding j

increase in the upper bound BOL RIP. Sensitivity studies for increases in BOL RIP have been perfomed which indica'e a transient-specific affeet on PCT.

For BVPS-1, the effect of this change is no longer being estimated becausa a' reanalysis has incorporated the-change.

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s XXIII. SAFETY INJECTION IN THE BROKEN LOOP Change

Description:

- Westinghouse recently completed an evaluation of a potential issue concerning the modeling of Safety injection (SI) 2 flow into the broken RCS loop for small break, loss of coolant accidents (SBLOCA).

Previously it had been assumed that Si to the broken RCS loop would result in a lower calculated PCT. Herefore, the ECCS broken loop branch line was modeled to spill the Si to the containment sump. The basis for this assumption included.

consideration for the effect of back pressure on the spilling ECCS line for cold leg breaks, which would see a higher back pressure for Si connected to the broken RCS toop when compared to spilling against containment back j

pressure. Spilling to the higher RCS pressure would increase SI to the intact loops, which is a benefit for PCT.

The effect on intact loop Si flow rates as well as the assumption that some of the Si to the broken loop would aid in '

RCS/ Core recovery resulted in the Westinghouse ECCS model assumption that SI to the broken loop was a benefit.

l llowever, when Si is modeled to enter into the broken loop, a significant PCT penalty is calculated by_ the

. j NOTRUMP small break evaluation model.

l When a newer conservative model based on prototypic test is used to model the configuration of the Si piping to the RCS cold leg in a Westinghouse designed PWR, a net PCT benefit is calculated. Improved condensation of the loop steam in the intact loops results in lower RCS pressure and larger Si flow rates. The increase in Si flow rates, due to lower RCS pressure, leads to the lower calculated PCT. nus, the negative effects of SI into the broken loop can be offset by an improved SI condensation model in the intact RCS loops.

q The improved condensation model is based on data obtained from the COSI test facility. The COSI test facility is a j

1/100 scale representation of the cold leg and Si injection ports in a Westinghouse daigned PWR. COSI tests demonstrated that the current NOTRUMP condensation model under-predicted condensation in the intact loops i

during Si and thus is a conservative model. Use of the improved condensation model has demonstrated that the I

current NOTRUMP small break LOCA analyses without the improved condensation model and no Si into the broken loop are more conservative (higher calculated PCT) than a case which includes Si into the broken loop and the improved condensation model.

)

Effect of Change:

i Westinghouse had determined, for imth BVPS-1 and BVPS-2, that the effect of the erroneous assumption on Small Break LOCA PCT is 150'F. However, this nonconservatism is more than offset by the overconservatism of condensation modeling in the intact loops which artifically increased calculated PCT by more than 150 F.

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4 A'ITACHMENT 3 REFERENCES 1.

NS-NRt 61464 " Correction of Errors and Modifications to the NOTRUMP Code in the Westinghouse Small Bu. LOCA ECCS Evaluation Model Which Are Potentially Significant,* Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC), Dated October 5,1989.

2.

WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 (Non-Proprietary), " Westinghouse -

ECCS Evaluation Model-1981 Version," 1981, Eicheldinger, C.

3.

WCAP-10266-P-A, Revision 2 (Proprietary), WCAP-10267-A, Revision 2 (Non Propt tary), Besspiata,J.J...

et.al., "1981 Version of the Westinghouse ECCS Evaluation Model Using the BALU (' de," March 1987.

4.

WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non Proprietary), " Westinghouse Small Break' ECCS Evaluation Model Using the NOTRUMP Code," Lee, N., et. al., August 1985.

5.

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

6.

"BART-Al A Computer Code for the Best Estimate Analysis of Reflood Transients", WCAP-9695-A (Non-Proprietary), March 1984.

7.

"10CFR50.46 Annual Notification for 1989 of Modifications in Westinghouse ECCS Evaluation Models,"

NS-NRC-89-3463, Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC), Dated October 5, 1989.

8.

WCAP-12909-P (Proprietary), " Westinghouse ECCS Evaluation Model Revised LBLOCA Power Distribution Methodology," Dated May 22,1991.

9.

NS-NRC-91-3578, " Westinghouse ECCS Evaluation Model Revised LBLOCA Power Distribution Methodology," Dated May 22,1991.

10. WCAP.13451, " Westinghouse Methcxiology For Implementation of 10 CFR 50.46 Reporting."

11. WCAP-10484, Addendum 1, " Spacer Grid Heat Transfer Effects During Reflood," Shimeck, December 1992.

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12. ET-NRC-91-3633, " Methodology Clarifications to WCAP-12909-P,* Letter from S.R. TRITCH (Westinghouse) to R.C. Iones (NRC), Dated November 21,1991.

13. ET-NRC-93-3971, " Notification of a Significant Change To Westinghouse Small Break LOCA ECCS Evaluation Model, Pursuant To 10 CFR 50.46(a)(3)(ii): Safety injection (SI) in the Broken Loop," Letter from N. J. Liparuto (Westinghouse) to R. C. Jones (NRC), Dated September 21,1993.

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