ML20102A350
| ML20102A350 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/20/1992 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9207240250 | |
| Download: ML20102A350 (26) | |
Text
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e i, NM f'jbr Valley Power Station Sh pngpo PA 15071-0004 JOHN (> Sit BLR Wte Prescent - Nuclear Group July 20, 1992 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Beaver Valley Power Station, Unit No. 1 and No. 2 DV-1 Docket No. 50-334, License No. DPR-GG DV-2 Docket No. 50-412, License No. NPF-73 10 CPE 50.46 Report of Changen or Errors in ECCS Evaluation Models This report is provided as notification of changes or errors in ECCS evaluation models which are reportable annually. The following attachments provide information as requested by 10 CFR 50.46:
Attachment 1 Provides a listing of each change or error in 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 "inaex" into Attachment 2 (Generic Descriptions).
Abtachment 2 Provides a aeneric description (based on information provided by Westinghouse) for each model change or error.
Attachment 3 Provides a list of references which occur in the 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:
SvPS-1 Large Break LOCA - 2149'P BVPS-1 Small break LOCA - 2010*F DVDS-2 Largo P'as:e LOCA - 2191*P 0
BVPS-2 Small becgr LOCA - 2176*F
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9207240250 DR 920720 p ADOCK 05000334 PDR
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Boavor Valloy Powsr Station, Unit No. 1 and No. 2
.BV-1 Dockot No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. HPF-73 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models Pago 2 Since nono of the calculated temperatures exceed 2200*F, no further action is required.
Sincoroly, 9C jY=n*s J. D. Sieber cc: Mr. L. Rossbach, Sr. Resident Inspector Mr. T. T. Martin, NRC Region I Administrator Mr. A. W. De Agazio, Project Manager Mr. M. L. Bowling (VEPCO) a
I. 3 Attcchtent 1 Summary of PCT Ef fects for BVPS IOCA Transients Attachment 2 P1 ant Tranil19Bt PCT E[foct (*F) Eqp_griotion f Pacto)_
BVPS-1 Large Break 10 II (Page 2) 10 V (Page 10) 0 VIII (Page 14)
(Note 3) IX (Page 15)
DVPS-1 Small Break 0 III.A (Page 4) 37 III.B (Page 4) 0 III.C (Pago 5) 0 III.D (Page 5)
(Note 2) IV (Page 8) 37 V (Page 10) 53 VI (Page 12)
(Note 3) VII (Page 13) 81 X (Page 18)
BVPS-2 Large Break 0 I (Page 1) 25 V (Page 10) 0 VIII (Pago 14) 30 IX (Page 15)-
0 XI (Page 19)
(Noto 3) XIII (Page 21)
BVPS-2 Small Break (Note 2) III.A (Page 4) 37 III.B (Page 4) 0 III.C (Page 5) 0 III.D (Page 5) 0 III.E ;Page 5) 5 III.F (Page 6)
(Noto 4) III.G (Page 6)
(Note 2) III.H (Page 6) 37 V (Page 10) 20 VI (Pago 12) 0 VII (Page 13)
< 32 X (Page 18)
-145 XII (Page 20) 20 XIII (Page 21) 17 XIV (Page 22)
Note 1: Plant specific impact on PCT using the last acceptable model has not been assessed. The current large break LOCA model includes these effects for calculation of PCT.
Note 2: Results in a small PCT reduction if corrected.
Note 3: Results in an unquantified PCT reduction if corrected.
Note 4: Under investigation by the NSSS vendor.
Attachment 2
- 1. h10DIFICATIONS TO Tile WREFLOOD Coh1PUTER CODE
Background:
A modification was made to delay downeomer overfilling. The de. lay corresponda to backfilling of the intact cold legs. Data from tests simulating cold leg injection during the post large break LOCA reflood phase which gave ridequate safety irjection flow to condense all of the available steam flow show a significant amount of subcooled liquid to be present in the cold bg pipe test section. This situation corresponds to the so-called maximum safety injection scenario of ECCS Evaluation hiodel analyses.
For maximum safety injs,: tion scenarios, the reflooding models in the Westinghouse 1981 ECCS Evaluation hiodel, the Westinghouse 1981 ECCS Evaluation blodd incorporating the BART analpis technology, and the Westinghouse 1981 ECCS Evaluation hiodel incorporating the BAsl{ analysis technology use WREFLOOD code versions which predict the downeomer to overfill. Flow through the vessel side of the break is computed based 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 whkh 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 downeomer 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 its equilibrium level. The appropriate WREFLOOD code-venions associated with the 1981 Westinghouse ECCS Evaluation hiodel and the 1981 Westinghouse ECCS Evaluation blodel incorporating the BART and Basil technology have been modified to incorporate the downcomer overtill logic update.
Effect of Change:
This change represents a mmlet enhancement in terms of the consistency of the appmach in the WREFLOOD code and the actual response of the downcomer level. la some cases this change could delay the overfilling process, which could result in a peak cladding temperature (PCT) penaLy. The nugnitude of the possible PCT penalty was assessed by reanalyzing the plant whi,ii is nuximum safeguards limited (CD=0.6 DECLO case) and which is most sensitive to the changes in the WREFLOOD cale. The PCT penalty of 16'F which resulted for this case represents the maximum PCT penalty which could be exhibited fer any plant due to the WREFLOOD logic change.
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Attachment 2
!!. h10DIFICATnNS TO THE BASH ECCS EVALUATION h10 DEL i %ckground:
In the BASH CCS h luation a hiodel (reference 3), .ne BART core model is coupled with equilibrium-NOTRUh!P .,n., .act code to calculate the dynamic interaction between the core thermohydrauhes and system
- bs i n the reactor coolant system during core reflood. The BASH code teflood model replace he
,g h OD calculation to produer a more dynamic flooding transient which reflects the close coupling between W core, ther.nhydraulic and loop tvhavior. Special treatraent of the B ASH computer code outputs is uwt to provide the core flooJing rate for se in the LOCBART computer code. The LOCBAD.T computer code iW.ts from the direct coupling of the BART computer code and the LOCTA computer code to directly calculate the peak cladding
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h: :.. iim he BA SH ECCS LaRaion hiocel include the modif5 tiers made to t'.e 1981 ECCS
( Evalu . , dLeu., sed previously, and the following previously unrepcrted modifications;
, Set a mnts were made la the basil computer code to treat special analysis cases which are r9ated to the trd , .d interfsecs; {j
- 1) A modificahon, to prv.ent the code from aborting, was made to the heat transfer model for the special
, m...,n when the quench front region r,. oves to the bottom the BASH core channel. The queneb heat 1 ...ppli d to the fluid node below the battom of the active fuel was set to zero. .
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- 2) A mo6fication, 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 ia the special case where the liquiditwo-phase ineface reaches the bottom of the BASH core channel. 0
- 3) hiodifications, to prev' Fe cU' from aborting, were made to increase the dimensions of certain arrays fu, special annlications.
- 4) A moo-:, ation was made to write adoitional variables to the tape of information to be provided to 5 LOCBART.
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- 5) Typegraphicai errors in the coding of some coavective heat transfer terms were corrected, but he correctk _s h x no effect ou the B ASH analysis results since the related e_rms are always set equal to zero.
- 6) A modification i as mads e the BASH coding to reset the cold leg cor 'itions, in a conservative manner, when the accumulators empty. The BASH model is initialized at the botbm of core recovery with the intact i
cold legs, lower plenum full of liquid. Flow into the downcomer then equals the pecumulator flow. The modification removed most of the intact cold leg water at the accumulatnr empty time by resetting the intact i cold leg conditions to a high quality two phase ~tixture.
In a typical BASH calculation, the downcomer is nearly full when the accumulators emptied. The delay time.
pnor to the intact cold leg water reding saturation, in sufficient to albw the downecmer to fill from tne addition of safeti njectioni fluid before the wate. in the cold legs reaches saturation. When the intact cold leg water reached saturation it merely flowed out of the break The cold leg water therefore, did not afiect the re'lood transient.
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-4 However, in a special case, a substantial time was required 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.
I Effect of Chr.nge:
For typical calculations, there is no effect on the PCT calculation for the majodty 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.
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Attachment 2 111. MODIFICATIONS TO THE NOTRUh1P Sh1ALL Bl<EAK LOCA EVALUATION MODE'.
Background:
The NOTRUh1P 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 PNRs," in compliance with the requirements of NUREG-0737,
' implementation of the TMl 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 remonse of the reactor coolant system during a small break LOCA and the SBLOCTA-lV 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 ') to correct erroneous codirg or improve the coding logic to pro:lude erioneous calculations. The modifientions indicated in A through I tielow have been incorporateJ into the production version of the cot. Remaining correa > and modifications are not significant and will be incorporated during tne next code uj>da: i.. reordance w, he Westinghouse quaih 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 moditication was made to preclude changing the region designation (upp<r, lower) for a node in a stack which does not contain the mhture-vapor interface. The purpose of the modification was to enhance tracking of 'he mixture vapor interface in a stael ed 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 regic1 designation does not typically affect the calculations, except for the nodes representing the core fluid volume (core nodes). In core nodes whi;h are designated as containing vapor regions, the use of the steam cooling heat intnsfer correlation is forced on the calculation in compliance with the requirements of Appendix K to 10CFR50, even .f the node conditions would i9icate otherwise. The use of the steam coolins heat transfer regime ateve the mixture level is documented on i age 3-1 of refere..: e 2.
Effect of Change:
In rare instancer, an i'1 correct at transfer 'orrelation could be selected if the region designation was impro},erly retlecteJ. An analysis calculation was performed for a three-loop plant which resulted in a decrease in the PCT of 6.5'F when the correctious were made for a calculation which would be affected by the change.
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 of interior fluid nodes were corrected. The heat transfer rate derivatives for subcooled, saturated, and superheated natural convection conditions for the upper region of intenor fluid nodes are given by equationa 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 Y
region heat transfer area in the calculation of derivatives. The error affected only the apper region heat transfer derivatives which are used by the code to characterize the implicit coupling of the heat rates to ch.tnges in the independent nodal variables.
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Effect of Change:
In rare instances, the amount of heat that could be transferred to the fluid could be improperly calculated.
The efft zt of the errors was expected to be small since the error would only affect the derivatives of the beat 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 calculation which would be affected by the change.
C. Change
Description:
)
Typographical errors in ettuations which calculate the derivatives of the natural convection mode of heat ,
transfer in the subroutive llEAT were corrected. A conductivity term used in the equations which calculate . _I the derivatives of the natural convection mot of heat transfer was incorrce'; typed as CK (to be used for the Thom or McBeth correlations, instead , f CKNC (to be used for the Jesired hicAdams correlation.-
Effect of Change:
A review of the code logic was performed to assess the effect of Ge_ error. In all equations that contain the typographical error, the incorrect variable is multiplied by 7ero. Therefore the typographical errors have no effect on the PCT results of the calculations.
D. Change
Description:
A typographical e: Tor was corrected in an equation which calculates the internal energy for nodes associated with the reactor coolant pump model when the associated reactor coolant pump flow links are found to be in critical 81ow. - An incorrect value for the mixture region internal energy in the fluid node downstream of a i pump flow link would be calculated if the pump flow link were in critical flow, c Effect of Change:
This section of ~> ding is not expected to be executed for small break LOCA Evaluation Model calculations since criGeal fiow m the reactor coolan' pump flow links does not occur. Therefdre 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.
E. Change
Description:
A modification was made to properly call some doubly dimer.sioned variables in subroutines.INIT and-TRANSNT. Some variables are doubly:dimensioned (X,Y) but were being used as if they were singly -
dimensioned.
Effect of Change:
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- A detailed review of the code logie indicated that all of the doubly'dimensioned variables had I as the second dimension in any of the erroneous calls. The computer inferrsd a 1 for the second dimensions in the improper subroutine calls. Therefore, there is no effect of this modification on the PCT.
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s F. Change
Description:
A modification was rnade to prevent code aborts resulting from-implementation of a new FORTRAN compiler. Due to the differen. treatments of the precision of numbers between the FORTRAN comfilers, the subtraction of two large, but close numbers resulted in ext.tly_ 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.
Effect of Change:
[
i An analysis calculation was performed for a four Hop plant which resulted in a larger than expected increase in the PCT of 4.8'F when the modification was implemented.
i 4 G. Change
Description:
I An error in the implementation of equation 5 33 of reference 2 was corrected. Equations 5 33 describes the 4
calculation of the flow link friction parame:erke 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 in used for all flow conditions.
For the case where the flow quality is zero, equation 5-34 is similar la 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 spec.iic volume vk in equation 5-33 is not equivalent to the' saturated fluid donor cell specific volume (vg. donor (k)) in equation 5-34.
Effect of Change:
This modification was expected to have only a small beneficial effect on the analysis.1 However, an analysis -
calculation was perfonned for a three-loop plant to quantify the effect and a larger than expected decrease in--
the peak cladding 2mperature 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 NOThUMP 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.
H. Change
Description:
A modification was made to correct an error in implementing equations L-28, L-52 and L-29, L-53 of f' reference 2. - The two pairs of e nations respectively describe the Partial derivatives of Fk with respect to pressure and specific enthalpy. F ' is an interpolation' parameter that is defined by equations L-27, L-51 of -
reference 2. In each pair the lower equation mtmber is for the subcooled condition, and the higher equation - j number is for the superheated condition %e denominator of each equation contains the differences between hkand h k-1 where hk is defined by equations L-21, L 45 and hk1-is defimed by equations L-22, L-46 of '
refere~e 2. Although the expression defining h k and hk-l were correctly calculated m NOTRUMP, they '
were i.,x used in equations L-28, L-52 and L-29, L-53 as they should have been.
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. Atixhment 2 1 .
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Effect of Change:
a.
I An analysis calcuV :. was performed for a four-loop plant which resulted in a decrease in f.he PCT of ,
i 12.S*F when the u.d.itication was made for a calculation which would be affected.
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,4 - IV. h!ODIFICATIONS TO THE SM ALL BREAK LOCTA IV COMPUTER CODE f The following modifications to the LOCTA IV computer code in the small break LOCA ECCS Evaluation Model !
! have been made:
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f A. Change
Deaription:
- A test was added in the rod-to-steam radiation heat transfer coefficient calculation to preclude the use of the
- i. correlation when the wall-to-steam temperature 4ifferential dropped below the useful range of the correlation.
! This limit was derived based upon the physical limitati.ans of the radiation phenomena.
I Effect of Change:
! There is no effect of the modification on reported PCTs since the erroneous use of the correlation forced the l calculations into aborted conditions.
4 l
B. Change
Description:
i An update wa performed to allow the use of fuel rod performance data from the revised Westinghouse (PAD 3.3) model.
l j.
j Effe t of Change:
l w evAnation me.cated that there is an insignificant effect of the modification on reported PCTs.
C. Change Descripilon:
i '
- - Modifications supportmg a general upgrade of the computer program were Implemented as follows
{ 1) the removal of unused or redundant coding,
- 2) better coding organization to increase the efficiency Of Calculations, and a 3) improvements in user friendliness
- a) through defaulting of some input variables, l b) simplification ofinput, l c) inpM diagnostic checks, and
- d) clarification of the output.
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- j. Effect of Change:
- Verification analyses calculations demonstrated that there was no effect on the calchlated output resulting -
from these changes.
D. Change
Description:
Two modifications improving the consistency between the Westinghouse fuel rod performance d:'. (PAD) and the small break LOCTA-IV fuel rod models were implemented:
8
Attachment 2
!) The form of the equation for the density of Uranium *lioxide in the specific heat correlation; which modeled three dimensional expansion was corrected to account for only two-dimensional thermal expansion due to the way the fuel red is nnleted. ff
- 2) An error in the equation for the pe!!et/ clad contact pressure was corrected. The contact resistance is never used in licensing calcu'ations.
Effect of Change:
The tirariumaioxide density correction is estimated to have a maximum PCT benefit of less than 2'F. while the contact resistance modification has no PCT efrect since it is not used.
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Attachment 2
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V. FUEL ROD h10 DEL REVISIONS 4
I Dackground:
During the review of the -original Westinghouse ECCS Evaluation Model following the promulgation of u 10CFR30.46 in 1974, Westinghouse committed to maintain consistency between future loss-of-coslant accident (LOCA) fuel rod computer mode!s 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 ini:ial conditions for the LOCA analys:s.
Change
Description:
It was found that the large break and small break LOCA code versions were not consistent with fuel design codes in the following areas:
- 1. The LOCA codes were not censister:t with the fuel rod design code relative to the flux depression factors at higher fuel enrichment.
- 2. The LOCA codes were not consistent with the fuel rod design code relative to the fuel rod gap gas - ,
conductivities and pellet surface roughness models.
- 3. The coding of the pellet / clad contact resistance model required revision.
Modifications were made to the fuel rod models used in the LOCA Evaluation Models to maintain consistency with the latest approved version of the fuel rod design code, c in addition, it was determined that irtegration of the cladding strain rate equation used in the large break LOCA Evaluation Model, as described in Reference 5, was being calculated twice each time step instead of once. The -
coding was corrected to properly integrate the strain rate equation.
Effect of Changes:
The changes made to make the LOCA tuel rod mMe!s consistent with' the fuel design codes were judged to be insignificant, as defined by 10CFR50.46(a)(3)(i). To quantify the effect on the calculated peak cladding -
temperature (PCT), calculations'were performed which incorporated the changes, including the cladding strain -
model correction for the large break LOCA, .For the large ' break:LOCA Evaluation Model, additional calculations, incorporating only the cladding straic 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.
4 a)Large Break LOCA The effect of the changes on' the large break LOCA peak'claddirig temperature was determined using the
- BASH large break LOCA Evaluation Model. The effects werejudged applicable to older Evaluation Models.
- Several calculations were performed to assess the effect of the changes.on the calculated results as follows:
- 1. Blowdown Analysis . - !
It was determined that the changes will have a small effect on the core average rod and hot aseembly aven;e rod parformance during the blowdown analysis. The effect of the changes on the blowdown i
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I r.ialysis 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.
l 3. Determination of the Effect on th Peak Cladding Temperature -
i
- The effect of the changes on tlw calculated peak clad < ling temperature wao Jetermined 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 Evahiation 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 -
l on the peak cladding temperature for the BART Evaluation Model, while calculatious performed for the j Westinghouse 1981 Evaluatica Model showed that the peak c: adding temperature could increase by _
j approximately 41'F.
I b) Small Break LOCA
^
j The effect of the changes on the small break LOCA analysis peak cladding temperature calculations was determined usin; the 1985 small break LOCA Evaluation Model by performing a computer analysis i calectations for a typical three-loop plant and a typical four-loop plant. The analysis calculations confirmed -
l that the effect of the ch .nges on the small break LOCA ECCS Evaluation Model were insignificant as defined l by 10CFR50.46(a)(3)(i). The calculations showed that 27'F would bound the effect on the calculated peak 2
cladding ten peratures for the four-loop plants and the three-loop plants. -It was judged that an' increase af l 37'F would bound the effect of the changes for the 2-loop plants.
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Attachment 2 i.
VI. SMALL BREAK LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION 3
p Change
Description:
- The Westinghouse small break. loss-of.coulant accident (LOCA) emergency core cooling system (ECCS) J Evaluation Model analyses assume that higher fuel rod initial fill pressure leads to a higher calculated peak
, cladding temperature gPCT), as found in studies with the Westinghouse large break LOCA ECCS Evaluation Model. However, lower fuel rod intemal pressure could result in <lecrease I claddinp 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 teruit in a higher calculated peak cladding temperature.
The Westinghouse small break LOCA cladding strain model is based upon a ccrrelation of Hardy's data, as described in Section 3.5.1 of Reference 5. Evaluation of tia limiting fuel rod initial fill pressure assumption i revealed that this model was used outside of the applicable range in the smaji break ~LOCA Evaluation Model calculations, allowing the cladding to expand and contract more rapidly than it si m!J. The model was corrected i to fit applicable data over the rtnge of snull break LOCA conditions. Correction of the cladding strain model-i affects the small break LOCA Evaluation Model calculations through the fuel rod internal ytessure initial condition assumption.
Effect of Changes:
Implementation of the corrected cladding creep equation results in a small reduction in the pellet to cladding gap when the RCS pressure exceeds the nxi intemal 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 nx! intemal 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 red internal pressure, implementation of an accurate correlation for cladding creep in small break LOCA analysu would reduce the expaasion 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 colJ leg break in typical three-loop and four-loop plants. The results ind;cated that the
- change to the calculated peak cladding 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 shether the rod intemal pressure was abo've or below the system pressure when the peak -
cladding temperature occurs. For the range of fuel rod intemal pressure initial conditions. the combined effect of tb 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 i a li rge as 60*F.
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Attachment 2 Vll. NOTRUMP CODE SOLU 90N CONVERGENCE Change
Description:
In the development of the NOTRUMP small break LOCA ECCS Evaluation Model, a number of noding sensitivity studies were performed to demonstrate acceptable solution cc.nvergence as required by Appendix K to 10CFR50. Temporal solution convergence sensi:ivity studies were performed by varying input parameters whkh govern the rate of change of key process variables, such as changes in the pressure, mass, and intemal 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 enh:.nce 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 performed with the NOTRUMP computer code for variations in initial RCS pressure, auxiliary feedwater flow rates, nnwer distribution, etc.. . o which re:a.:'c-1 m peak cladding temperature (PCT) variations which were greater tha: mticipated based upon engineering judgment. In addition, the directioi of the PCT variation confheted with engineering judgment-expectations in some cases. The unexpected variability of the sen sitivity study results indicated that the riumerical solution may not be properly converged.
Sensitivity stuJies ure performed for the time step size selection criteria which culminated in a ievision to the-recommended time step size selection criteria inputs. Fixed input values originally recommended for the steady.
state and all break trans'ent cakulations were modified to assure converged results. The NOTRUMP code was re-verified against the SUT-08 Semiscale experiment and it was confirmed that the code adequately predicts key small break phenomena.
x Effect of Changes:
Generally, the modifications result in s.wl! shifts in timing of core uncovey and recovery. However, these changes m.9 result in a change in the calculated peak cladding temperature which exceeds 50'F f at some plants.
Based on representative caktlations, however, this change will most likely.-,esult in a reduction in the calculated peak cladding temperatu:e. Since the potential beneficial effect of a non-converged solution is plant specific, a generic PCT effect cannot be provided. Howevm it lus been concluded that current licensing basis.resrlts remaia valid since the results are conservative relativa to the change.
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Attachment 2 Vill, LARGE BREAK LOCA BURST AND BLOCKAGE ASSUMPTION
Background:
The claJJing swelling and flow blockage models were reviewed in detail during the NRC's evaluation of the Westinghouse Evaluatio- Model. However, the use of the averago rod in the hot asserrbly may t.ot have been documented in a manner detailed enough to allow the staff to adequately assess this aspect of the model AppecJix K to 10CFR50 requires consideration of the effects cf flow blockage resulting from the swellmg and rupture of the fuel rods auring a loss-of-coolant accident (LOCA). 10"FR50 Appendix K Paragraph I.B states:
. To be a.:ceptable the swelling and ru; ture calculations shall be based on applicable data in such a way that the degree of sv elling and incidence of rupture are not underestimated."
'n Westinghouse ECCS Evaluation Model catu.ations, the average rod in the hot assembl; is used as the basis for calculating the effects of flow blockage. if a significant number cf fuei rods in the hot assembly are operat:ng at power li els greater than that of the average rod, the time at which cladding swelling and rupture is calculated to occur may be predicted later in the LOCA transient, si 'e the lower power rod will take longer to heat up to levels where swelling and rupture will occur A review of the Westinghouse m c. ased to predict assembly blockage was performed. This model was developed from the Westmghouse 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. These u.adels provide the m:ans for determining assembly wide blockage once the mean burst strain hes been estabbshed. Implementation of these burst models has relied upon the average rod to provide the mean burst strain. The average rod is a low power g rod producing the power of the average of rods in the Lot assembly and is primarily used to calculate the enthalpy rise in the hot assembly. Use of the average rod in the raodel assumes that the time at wiich blockage is caleutated 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 rods will burst {'
With this representation, however, the true onset of blockage would likely begin earlier, as the highest power rods reach their bt rst temperature. This time is estimated to be a few seconds prior to the time whea the average rod bursts.
1.arge break LOCA Evaluation Models wnich use BART or BASH simulate tue hot assembly rod with the a:tuul average power, while older Evaluation Models use an avunge rod power which is adjusted downward to accoun for thimbles (this is described in detail in Addendum 3 to reference (6)). 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 o' der Evaluativn Models.
Change
Description:
Ample experimental evidence currently exists which shows that flow blockage does not result in a heat transfer f peaalty during a LOCA. In addition, newer Evalu tion Models have been developed and licensed which demonstrate that the cluer rivaluation Models contain a substantial amo*2nt of conservatism. Westinghouse concluded that further artificial changes to the ECCS Evaluation Models to force the calcu ation of an earlier burst
- 3 time were not necessary. In rare instances where burst has not occurred prior to the flooding rate falling tielow 1.0-inch /second, the results of the ECCS analysis calculation ar_ supplemented by a permanent assessment of margin. Typically this will only occur in cases ehere the calcula.ed PCT is low. Westbghouse concludes that no model change is required to calculate an earlier burn tace.
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- i. IX. STEAM GENERATOR FLOW AREA r i-i; Backgrouni L
h 1 Licensees are normally required to provide assurance that there exists only an extremely low probability of l abnormal leakage or gcoss rupture of any past of the reactor coolant pressure boundary (General design criteria' l.4 and 31). The NRC issued a regulatory guide (RG 1.121) which addressed this requirement specifically for steam _ ,
3 generator tubes in pressurized watet reactors. in that guide, 3he staff required analytical and experimental i eviden e that steam generator tube integrity will be maintair.ad for the combinations'of the loads resulting from a LOCA wi.h the loads com r a safe shutdown earthquake (SSE). These loads are combined for added conservatism l
[ in the calculation of stmetural integrity. This analysis provides the basis for establishing criteria for re noving -
- from service tubes which had experienced significant degradation.
l Analyses performed by Westinghouse it support of the above requirement for various utilities, combined the most
- j. severe LOCA loads with the plant specilic SSE, as delineated in the lesign 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 sei.uees the ticw area through the steam generator. The reduced How area increases the resistance through the steam generator to the flow of steam from the core during a LOCA, which potentially ccvid increase the calculated PCT,
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[ Change
Description:
1 The effect of tube deformation and flow area reduction in the steam generator was analyzed and evaluated for some plants by Westi nghouse in the late 1970's and early 1980's. . The combination of LOCA and SSE loads led to the following calculated phenomena-
- 1. _OCA and SSE loads cause the steam generator tube bundle to vibrate, t
i 2. The tube support plates may be deformed as a result of lateral loads at the wedge supports at the periphery of >
j the plate. The tube support plate deformation may cause tube deformation.
j 3. During a postulated large LOCA, ti s primary side depressurizes to containment pressure. Appiying the t
resulting pressure differer.tial to the Jeformed 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 thA 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 6e resulting flow area reduction a the LOCA PCT. The postulated bre-dc examined was the steam
- - generator outlet break, because this break wasjudged to result in t he gruatest 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.
l function because the degree of ibw area reduction.was small,'and the postulated break at the steam generator ,
t outlet resulted in a low PCT.'
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[ In April oi 1990, in considering the e fect of the combination of LOCA.+ SSE loadings on the steam generstor, p . component, it was determined that the potential for flow area reduction due to the contribution of SSE loadings
, should.be in luded in othar LOCA analyses. With SSE loadings, flow area reduction may occur in all steam . .
generators (notjust the f.ulted loop). Therefore, it was concluded that the effucts of flow area reduction during the most limiting prirnary pipe break affecting LOCA PCT,~i.eJ 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 continuc to perform their intended safety function.
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! Consequently, the action was taken to address the safety significance of steam generator tube collapse during a -1
! cold leg break LOCA. The effect of flow area reduction from combined LOCA and SSE loads was estimated.
l The magnitude of the flow area reduction was considered equivalent to an inctmed level of steam generator tube . ,
, plugging. Typically, the area reduction was estimated to range from 0 to 7.5 %, depending on the magnitude of -
l the seismic loads. Since detailed non-linear seismic analyses are not available for Series 51 and earlier design j steam generators, some area reductions had to be estimated based on'available' information, r st most of these l plants, a 5 percent flow area reduction was assumed to occur in each steam generator as a result of the SSE. For i l
1 these evaluations, the contribution of loadings at the tube support plates from the LOCA cold leg break was j assumed negligible, since the additional area reduction, if it occurred, *vould occur only in the broken loop steam generator. ,
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! 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 i loadings and c,tinue to perfmm their intended safety function. it is judged that this requirement applies to j undegraded as well as locody degraded steam generator tubes. Compliance with GDC 2 is addressed below for
, beth conditions, i
For tubes which have not experienced cracking at the tube support plate elevations, it is Westinghouse's i engineering judgment that the calculation of steam generator tube deformation or collapse as a' result of the l combination of LOCA loads with SSE loads does not conflict with the requirements of GDC 2. -During a large -
l break LOCA, the intended safety functions ~of the steam generator tubes are to provide a flow pa'.h for the venting i cf steam generated ic the core through the RCS pipe break and to provide a flow path such that the other plant systems can perform their intended safety functions in mitigating the LOCA event.
!- . Tube deformation has the same effxt on the L OCA event as the plugging of steam generator tubes. The ef fect of tube deformation and/or collapse can .se taken 'into account by assigning an appropriate PCT penalty,'or accounting for the area reduction directly in the analysis. Evaluations completed 'io date show th.. tube deformation results in acceptable LOCA PCT.- From a steam generator structuzd integrity perspective, 4 sion III
~
} of the ASME Code recognizes that inelastic deformation can occur for faulted condition loadings. s are no
'- requirements that equate steam generator tube deformation, per se, with loss of safety function. ' Joss sectiona!
l- bending stresses in the tubes at the tube support plate elevations are considered secondary stresses vithin the
[ definitions of the ASME Code and need not be considered in establishing the limits for allowabb steam generator tube wall degradation. Therefore, for undegraded tubes, for the expected degree of flow area redi etion,' and?
despite the calculation showing potential tube collapse for a limited number.of tubes, the steam generators j continue to perform their required safety functions after the combination of LOCA + SSE loads, mting the .
[ requirements of GDC 2.
j During a November 7,1990 nteting with a utility and the NRC staff on this subject, a concern was raised that
{ tubes with partial' wall cracks at the tube support plate elevations could progress to through-wall cracks during j tube 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. Westinghcuse did n'ot consider i' the potential for secondary to primary inleakage during resobthn of the steam generator tube collapse item. ' This -
E is a relatively new item,'not previously addressed, since cracking at the tube support plate elevationsc had been . l I ' insignificant in the early 1980's when the tube collapse item was evaluated in depth. There is ample datt j h available which demonstrates that undegraded tubes maintain their integrity under collapse loads. There is also. .I
[ some data which shows that cracked tubes do not behave significantly differently from unen.cked tabes when' j - collapse loads are applied. However, cracked tube data is available only for round or slightly ovalized tubes.
b 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 estinute takes hto account such factors as the possibility of a seismically induced LOCA, the expected . .
j
. occurrence of cracking in a tube as a function ~of height in the steam generator tube bundle, the localized 'effect of- ,
i-the tube support plate deformation, and the possibility that a tube which is ident:ried to deform during LOCA + 1 e
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l SSE loadings would also contain a partial through wall crack which would result in significant inleakage. To '
turther reduce the likelihood that craeked 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 terions of the tube support plate elevations near wedges that are most susceptible to collapse which may then lead to penctration of the primary pressure boundary and signitkant inleakage during a LOCA + SSE event. ;
Change
Description:
4 As noted above, detailed analyses whis provide an estimate of the degree of flow area reduction due to both seismic and LOCA forces an not available for all steam generators. The information that does exist indicates that the flow area reduction may range from 0 to 7,5 percent, depending on the magnitude of the postulated forces, and accounting for uncertainties, it is difficult to estimate the flow area reduction for a particular steam genertstor design, based on the results of a differtrt 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 stesm generators, the risk associred with flow area reduction 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 a:bes 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 esthnated 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 reduction by. assuming
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additional tube plugging in the analysis..
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Attadhment 2 X. AUXILIARY FEEDWATER ENTilALPY SWITCl!OVER FOR SBLOCA ANALYSIS i
Change
Description:
t 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 t1.0 higher enthalpy main feedwater from the feedwater piping after actuation of auxiliary feedwater, In the Westinghouse SBLOCA ECCS Evaluation mod;1s using either the WFLASH or NOTRUMP analysis technologies, this time is useo 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 reWew 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 th Westing':cuse SBLOCA ECCS Evaluation Model input value and a .uue corresponding to the plant ,:,afiguration' dts from tne specific guidance provided to the analyst for determining the auxiliary feedwater enthalpy delay time. ' In both the WFLASH and NOTRUMP methodsi a standard perge l del..y time was recom*nended, in the NOTRUMP analysis methodology, a standard input v6.e judged to bc conservative based apon phenomena observert during experiment SUT-08 in the Semiscale test Lacility'was used, However, further investigation showed that the standard input value could result in a non-conservative calculation ;
of the peak cladding temperature.
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Attachment 2 XL POWER SHAPE SENSITIVITY MODEL (PSSM)
Background:
Large Break LOCA analyses have been traditional'y performed using a symmetric,' chopped cosine axial power shape. Recent calculations have shown that there was a potential for top-skewed power distributions to result in.
Peak Cladd.ng Temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution.
Westinghous.' has developed a process, 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
- suruillance data. This process, called the Power Shape Sensitivity Model (PSSM), is described in a topical report (WCAP 12909-P),
. !r in May,1991, Westinghouse transmitted to the NRC the report titled, "Westinghe ise ECCS Evaluation Model:
Revised Large Break LOCA Power Distribution Methodology," which describes the process that Westinghouse is now using to more accurately iccount 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 suhtaitted. 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. When applied in the .
Reload Safety Analysis for Beuer Valley Unit 2 Cycle 4, the NRC staff had not yet issued notification of the acceptability of the PSSM methodology. But, 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 Caange:
Implementation of this methodology has no effect on calculated PCT.
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j X11. PLANT SPECIFIC INPUTS AND MISCELLANEOUS INPUT ERRORS J
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Background:
\
To maintain conformance with 10CFR50.46 PCT limits, compensatory benefits had to be documented for the
, BVPS-2 SBLOCA analysis. The compensatory benefit was derived by performing plant specific sensitivity i analyses. The benefit results from the following:
Plant specific peaking tactors consistent with the Cycle 3 & Cycle' 4 design / Technical Specification limits were used. The previous UFSAR analysis was based in part on bounding generic salues.
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, Plant specilic Axial Offset consistent with the Cycle 3 & 4 design /Technient Specification limits, using ,
- tL current p'.mt specific methodology v,as used. The previous UFSAR analysis was based on generic
- power shapes.
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- A miscellaneous input error (s) in the UFSAR analysis was corrected.'
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. Effect of Change:
The combination of the above changes and error corrections results in a 145' F reduction in calculated PCT.
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- g r XI!!. RCS TEMPERATURE DISTRIDUTION - <
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Background:
e While evaluating the effects of reduced thermal design flow iTDF), anomalies were discovered in the interaction between the RCS temperature distribution methodology and the actual analysis inputs for bo'h SBLOCA &
LBLOCA analyses. Action was taken to investiga'e and-cvaluate the anomatics.= fA portion of the apparent .
- discrepancy is attributed to slight changes in the LOCA inputs that result from miscellaneous evolutionary changes to the plant characteristics such as the Steam Generatur Foeling Factor, wPrh was recently recalculated Other
- differences are attributed to deviations that ecenrred in the LOCA analyses themseJves, such as an extraction error resulting in incorrect LBLOCA input.
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I Effect of Change: ,
The LBLOCA analysis RCS Tavg inputs are greater than necessary for either current TDF or reduced TDF cases, and thus the analpis remains bounding, though the benefit is unquantified.
The SBLOCA analysis RCS T,vg inputs are greater than necessary for either current TDF or reduced TDF cases _
and thus a PCT penalty of 20*F is assigned.' This PCf penalty is considered to be reportable under 10CFR50.46 as an error in the application of the Evaluation Model, L
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) XIV. RCCA GUIDE TillMBLE AREA i
Background:
{
' The only VANTAGE SH Zirealoy grid fer.ture which significently affects the SBLOCA analysis is the increase in 1 -- deeign rod drop time. The Westinghouse Srnall Break model assumes the' reactor core is brought to a suberitical j condifian by the negative reactivity of the control rods. The increase in the design rod drop time to 'a maximum ,
j value of 2,7 seconds exceeds the 2.4 second value in the existing SBLOCA analysis.
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! Effect of Change:
- 1. .
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j /a evaluation was performed which determined that a 3*F PCT penalty applied for the merease m rod drop time
! of 0.3 seconds. The decrease in core pressure drop associated with thimble plug removal has an inconsequential j effect on the SBLOCA analysis. However, the.SBLOCA unalysis did not model the effect of the guide thimble - +
i interior area and volume on the transient response. Studies l ave shown that the fluid in this volume will interact l with the remaining core fluid. An assessment of this interaction based on previous censitivities indicated that a 4= 17'F PCT penalty applies.
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REFERENCES i
- 1. NS-NRC-89 3464 " Correction of Errors and hiodifications to:the NOTRUh1P Code in the Westinghcuse Small Break LOCA ECCS Evalustion Model Which Are Potentially Significant," Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC). Dated October 5,1989.
l- 2. WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9'221 A, Revision 1 (Non-Proprietary), ' Westinghouse.
{ ECCS Evaluation Model - 1981 Version," 1981, Eicheidinger, C.
- 3. WCAP 10266-P-A, Revision 2 (Proprietary), WCAP-10267-A, Resision 2 (Non-Proprietany), Besspiata,J.J.,
j et.al., "1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,' March 1987.
- 4. WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietr.ry), 'We. 'inghouse Small Break ECCS j Evaluation Model Using the NOTRUMP Code," Lee, N., et. al.,cAugust 1985.
i .
i L 6. "BART Al A Computer Code for the Best Estimate Analysis of Reflood Transients *, WCAP-9695 A (Non-.
- Proprietary), March 1984. a y
j- 7. '10CFR50.46 Annual Notification for 1989 of Mixlifications in Westinghouse ECCS Evaluation Models,"
! - NS-NRC-89 3463, Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC). Dated October 5. ,
! 1989. .
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