ML20082E047
| ML20082E047 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/25/1991 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9107310119 | |
| Download: ML20082E047 (22) | |
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A. WhN#y fih% P' blsltOn s,m , w.3., a tm aa Km D H B[k vu e n,- w w w. c. me July 25, 1991 U. S. Nuclear Regulatory Commission Attn: Doc 1mont Control Desk Washington, 20555
Subject:
11onver Valley Power Station, Unit No. 1 and No. 2 IIV-1 Docket No. 50-334, License No. DPR-66 IIV-2 Docket No. 50-412, License No. NPF-73 10 CPR 50.46 Report of Significant Changen or Errorn in ECCS Evaluation Models This report is provided as notification of changes or errors in ECCS ovaluation models which have been determined to be significant or potent 3-11y significant as defined in 10 CFR 50.46. The rationale for this de rmination is as follows:
13VPS Unit I has recently been provided a revised large break LOCA analysis which incorporates changes described in Attachments 1 and 2. The outcome of the current large break LOCA analysis, had it been performed without incorporating these changes (i.e.,
the "last acceptable model"), has not been separately quantified. Therefore, assc.ssment against the 10 CPR 50.46 reporting criteria is not possible.
IIVPS Unit 2 changes descr2 bed in Attachments 1 and 2 are significant because the sum of the absolute magnitudes of the respective temperature changes resulting from changes to the large break LOCA model differs by 55'r from the temperature calculat ed for the limiting transient (also large break LOCA) using the last acceptable model.
The following attachments provide informatio- as requested by 10 CTR 50.46:
Attachment 1 Provides a listing cf each chnge or error in an acceptable evaluation model that affects the peak fuel cladding temperature (PCT) calculation for particular transients. It quantifies the etfect of each change with respect to its potential plant-specific impact on PCT for that transient and provides an "index" into Attachment 2 (Generic Descriptions).
9107310119 910725 l0' PDR ADOCE 05000 h4 I t P PDP l l
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- Bot.ver Valley Power Station, Unit 110. I and 11o. 2
. BV-1 Docket tio. $0-334, License lio. DPR-66 ,
BV-2 Docket No. 50-412, License lio. HPF 73 10 CFR 50.46 Report of Jignificant Changes or Errors in i ECCS Evaluation ModelF Page 2 ,
i Attachment 2 Provides a generic 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. Thesa 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: l BVPS-1 Large Break - 2149'F DVPS-1 Small Break - 2010*F BVPS-2 Large Break - 2176*F BVPS-2 Small break - 2121*F ,
Since none of the calculated temperatures exceed 2200*F, no i further action is required. ,
Sincerely, ;
k 1f
<J. . df)eber Vice President Nuclear Group- .
cc: Mr. J. Beall, Sr. Resident Inspector !
l Mr. T. T. Martin, NRC Region.I Administrator i
Mr. A. W. DeAgazio, Project Manage?;
Mr. M. L. Bowling (VEPCO)
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Rummary of-PCT Effects for BVPS LOCA Transients Attachment 2 Plant Transient PCT Effect ('F) DescriptionfPage)
BVPS-1 Large Break 10 II (Page 2)-
10 V (Page 10) 0 VIII (Page 14)
(Note 1) IX (Page 15)
BVPS-1 Small-Break 0 III.A (Page 4) 37 III.B (Page 4) 0 III.C (Page 5) 0 III.D (Page 5) l (Note 2) IV (Page 8) 37 V (Page 10) 53 */I (Page 12) ,
(Note 3) VII (Page 13) l 81 X (Page 18)
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BVPS-2 Large Break 0 I (Page 1) 25 V (Page 10) j 0 VIII (Page 14)- '
L 30 IX (Page 15)
, BVPS-2 Emall Break (. lote 2) -III.A (Page 4) 37 III.B (Page 4) 0 III.C (Page 5)
O III.D (Page-5) 0 III .E - (Page 5) 5 III.F-(Page 6)
L (Note 4) III.G (Page 6) _
(Note 2) III.H (Page 6)
- 3 7. V (Page 10) 60 VI (Page 12)
(Note 3) VII'(Page 13) 32 X (Page.18)-
Note 1: Plant specific impact -on PCT itsing the last acceptable
- j. has not been. assessed.- The current large break.
mode) l LOCA- model includes' these -effects for calculetion~of i: PCT.
Note 2: Results in a small PCT reduction if corrected.
I ~ Note 3: Results in an unquantified PCT reduction is corrected.
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l . Note 4: Under investigation by the-NSSS vendor.
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Attachment 2
- 1. h10DIFICA110NS TO Tile WREI LOOD COh1PUTER CODE
( llackground:
A modification was made to delay downcomer overfilhng. The delay corresponds to backtilhng of the mtact cold legs, Data from tests simulating cold leg injection during the post large break LOCA iellood phase which gave adequate safety injection llow to condense all of the available steam flow show a sigmlicant amount of subcooled liquid to be preaent in the cold leg pipe test section. This situalmn conesponds to the so called masimum safel) in}ectmn scenario of I!CCS Evaluation N1odel analyses.
For maximum safety injestion scenanos, the tellooding models in the Westmphouse 1981 ECCS Evaluatmn hlcJel, the Westinghouse 1981 ECCS Evaluation hiodel incorporating the 13 ART analysis technology, and the L Westinghouse 1981 ECCS Evaluation 51odel incorporating the llASil analysis technology use WREFI OOD code versions whi,h predict the downeomer to userfill. Flow through the sessel side of the break is computed based upon the asailable head of water in the downeomer in WREFLOOD using an incompressible flow in an open thanno, method. A modification to the WRiil LOOD computer code was made to consider the cold leg inventory which would be present in conjunction with the enhanced downcomer lesel in the non taulted h> ops.
Change De.seription:
. WREFLOOD code logie was altend to consider the filling of the cold legs together with downcomer methlling.
Under this coing update, when the downcomer level exceeds its maximum value as input to WREFLOOD, Gquid 110w into the intact cold leg, as well as spillage out the break, is considered. This logie radification stabilizes the overfilling of the vessel downeomer as it approaches it equilibrium level. The appropriate WREFLOOD code f versions associate l with the 1981 Westinghouse ECCS Evaluation hiodel and the 1981 Westinghouse ECCS
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/ Evaluation hiodel incomorating the 11 ART and ilASil technology have been modified. to incorpoiate the E downeomer overfill logie update.
'M Ettect of Change:
C.
This change represents a model enhancement in terms of the consistency of the appmach in the WREFLOOD code 2 and the actual response of the downeomer level. In some cases this change could delay the overtillmg process, which could result in a peak cladding temperature (PCT) penalty. he magnituJe of the possible PCT penalty was assessed by reanalyr.ing the plant which is maximum safeguards limited (CD-0.6 DECLG case) and which is mo;t 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 logie change.
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Attachment 2
. !!. MODIFICATIONS TO Tile basil ECCS EVALUATION MODEL Dackground:
in the basil ECCS Evaluation Model (reference 3), the BART core model is coupled with equihbrium-N_OTRUMP computer code to calculate the dynamic interaction between the core thermohydraulics and system behavior in the reactor coolant system during core reflood. The Basil code renood model replaces the
- WREFLOOD calculation to produce a more dynamic flooding transient which reflects the close coupling between core thermohydraulic and loop behavior. Special treatment of the basil computer code outputs is 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 basil ECCS Evaiuation Model include the modifications made to the 1981 ECCS l Evaluation Model, discussed previously, and the following previously unreported modifications:
Several improvements were made to the basil 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 tnmsfer model for the special situation when the quench front region moves to the tuttom tne BASli 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 codirig these areas was generalized to prevent mass imbalance'in the special case where the liquid /two-phase interface reaches the bottom of the basil core channel.
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- 3) Modifications, to prevent the code fmm 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 of information to be provided to j' .LOCBART.
- 5) Typographical errors in the coding of some convective heat transfer imms were corrected, but the corrections j have no effect on the B ASil analysis results since the related terms are always set equal to zero.
- 6) A modification was made to the ilASl{ coding to reset the cold leg conditions, in a conservative manner, when the accumulators empty. - The BASl! model is initialized '.t the bottom of core recovery with the intact
,_ cold legs, lower plenum full of liquid. Flow into the downeomer then equals the accumulator flow. The l= modification removed most of the intact cold leg v/ater at the accumulator empty time by resetting the intact cold leg conditions to a high quality two phase mixture.
- In a typical basil calculatbn, the downcomer is nearly full when the accumulators emptied. The delay time, 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 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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s Attachment 2 llowever, 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 downeomer tilled, which artificially
, perturbed the transient response by incorrectly altering the downcomer fluid conditions causing the code to abort.
F.ffect of Change:
For typical calculations, there is no effect on the PCT calculation for the nujority 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, ,
l singly or in combination.
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E Attachment 2 Ill. h10DIFICATIONS TO Tile NOTRUh1P Shi ALL BREAK LOCA EVALUATION h10 DEL 11ackground:
The NOTRUhtP small break LOCA ECCS Evaluation hiodel (refe enee 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 Thil A'etion Plan,' Section ll.K.3.30. In the NOTRUh1P suull break LOCA ECCS
, Evaluation model, the NOTRUh1P code is u.ed to calculate the thernud-hydraulie sesponse of the reactor coolant system during a small break LOCA and the SilLOCTA-IV computer program is used to calculated the performance of fuel rods in the hot assembly, i
Several modifications have been nude to the NOTRUhlP 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 i assurance procedures for computer code maintenance. The following nuxlitientions to the NOTRUh1P small break LOCA ECCS Evaluation hiodel have been made; ;
A. Change
Description:
A modification was n.ade to preclude changing the region designation (upper, lower) for a mxle 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 mxles and to preclude a node in a stack, which does not contain the mixture-vapor interface, from changing the region designation. l The update does not affect the fluid conditions in the rmde, only the designation of the region of the node.
The region designation does not typically af fect the calculations, except for the nodes representing the core
' fluid volume (core nodes). -In core ncdes 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 mxte conditions would indicate otherwise. The use of the steam coohng heat transfer regime almve 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.
B. Change
Description:
Typographical errors in the equations which calculate the heat transfd 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 subenoted, saturated, and superheated natural convection conditions for the upper region of interior fluid rmdea are given by equaticas 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-are uw' by the code to characteric the implicit coupling of the heat rates to changes in the independent nodal variables.
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- Etfeet of Change:
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 atfect the derivatives of the heat rates for vapor regions that are in natural convection. An analysis calculation was performed for a three-hop plant which resulted in'a larger than expected increase in the PCT of 36.7*F when the coricetion was made
- on a calculation which would be affected by the change.
C. Change
Description:
Typographical errors in equations which calculate the derivatives of the natural convection mode of heat transfer in the subrentine llEAT 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 Thom or hicBeth correlations, instead of CKNC (to be used for the desired hicAdams correlation.
- Effect of Change:
A review of the code logic was performed to assess the effect of the error, in all equations that contain the typographical error, the incorrect variable is multiplied by 2cro. Therefore the typographical errors have no effect on the PCT results of the calculations.
D, Change
Description:
A typographical error was corrected in an equation which calculates the internal energy for rmdes associated 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 nodo downstream of a pump flow link would be calculaN! if the pump flow link were in critical flow.
Effect of Change:
This section of coding is not (xpected to be executed for snudi break LOCA Evaluation hiodel 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,
- E. Change
Description:
A modification was made to properly call some doubly dimensioned variables in subroutines INIT and TRANSNT. Some vari _abka are doubly dimensioned (X,Y) but were being used as if they were singly dimensioned.
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- l. - Effect of Change:
L A detailed review of the code logic indicated that all of the doubly dimensioned variables had I as the second ll __
dimension in any of the erroneous calls. The computer inferred a 1 for the second dimension in the improper I
subroutine calls. Therefore, there is no effect of this modification on the PCT.
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- i. Attachment 2 i b-g - F. Change
Description:
! A modification was 'made to _ prevent cale aborts resulting from implementation of a new FORTRAN jl compiler, Due to the different treatments of the precision of numbers between the FORTRAN compiltrs, the t 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 nmie approached, but was not equal to zero. >
i i 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. ,
O, . Change
Description:
l 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 ck for single phase flow in a non critieel now link L. In the erroneous implementation, equation 5 33 was replaced hv 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 imity when the flow quality 'is zero and the donor cell and flow link fluids are saturated, equations 5 33 and 5-34 are equivalent, llowever, for subcooled flow the flow link specific volume yk ni equation 5-33 is not equivalent to the saturated fluid donor cell specific volume
- (vk. donor (k)) in equation 5-34.
l : Effect of Change:
This modification was expected to have only a small beneficial effect on the analysis llowever, 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 hiop seal steam venting placed on the model for conservatism. Evahiations of the effect of this -
change wil_t be examined as part of the investigation of the larger than expected sensitivity resets.
IL Change Descriptioni A modification was made to correct an error in implementing equations L-28, L-52 and L-29, L-53 of reference 2 The two pairs of equaGons respectively describe the Partial derivatives of Fk with respect to -
pressure and specific enthalpy, Fk is an interpolation parameter that is defined by equations L 27, L-51 of l'
' reference 2, In each pair the lower equation number is for the subcooled condition, and the higher equation .
mimber is for the superheated condition The denominator of each equation contains the differences betweet, 7
' hkand hk-1 where hk is' defined by equations L-21, L-45 and kh -1 is defined by eqWons L-22, L-46 of E - reference 2. Although the expression defining hk and hk-1 were correctly calculated in NOTRUMP, they were not used in equations L-28, L-52 and L 29, L-53 as they should have been.
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Attachment 2
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-.Effect of Change: ,
An analysis calculation was performed for a four loop phmt which resulted in a decrea e in the PCT of 12,8'F when the modir. cation was made for a calculation which would be affected, t
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Attachment 2 ~
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IV. MODlHCATIONS TO Tile SM ALL llREAK LOCTA IV COMPUTlitt CODE The following modifications to the LCCTA IV computer code in the small break LOCA ECCS Evaluation Mohl have been made:
- A. Change
Description:
A test was added in the rod-to-steam radiation heat transfer coefficient calculation to preclude the use of the correlation when the wall to4 team temperature-dif ferential dropped below the useful range of the correlation.
This limit was derived based upon the physical limitations of the radiation phenomena.
Effect of Change:
There is no effect of the nodification on reported PCTs since the erroneous use of the correlation forced the -
calculations into alerted conditions.
IL Change Description.
An update was performed to allow the use of fuel md performance data from the revised Westinghouse (PAD . -l 3.3) model. '
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Effect of Change:
' An evaluation indicated that there is an insignificant effect of the mmlification im repated PCT..
C Change
Description:
Modifications supporting 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
- 3) improvement = in user friendliness
- a) through defaulting of some input variables,
' b) simplification of input, c) input diagnostic checks, and .
d) clarification of the output.
- Ef fect of Change
- Verification analyses calculations demonstrated that there was no effect on the calculated. output resulting from these changes.
D.' Change Descriptiom
-Two' modifications improving the' consistency between the Westinghouse fuel rod performance data (PAD) -
and the small break LOC 1'A-IV fuel rod models were implemented:
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Attachment 2
- 1) The form of the equation for the dernity-of Uranium-dioxide in the specific heat correlation, which modelled three dimensional expansion was correctal to accomt for only twov'imensional thermal.
- expansion due to the way the fuel rod is modeled. .
- 2) _An error in the equation for the pellet / clad contact pressure wn currectO The contact resistance is never used in licensing calculations.
- Effect of Change:
The Uranium-dioxide density corroetion is estimated to have a up)um 1*C'r berme of less than 2*F, while the contact resistance modification has no PCT effect since it is not uwo.
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Attachment 2 V.' FUEL ROD MODEL REVISIONS
Background:
During the review of the original Westinghouse ECCS Evaluation Model following the promulgation of 10CFR50.46 in 1974, Westinghouse committed to maintain consistency between future loss-of-coolant accident (LOCA) fuel rod computer models and the fuel rod design computer models used to predict fuel md normal '
operation performance. These fuel ral design codes are also used to establish initial conditions for the LOCA analysis.
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 consistent with the fuel rod design code relative to the flux depression factors at higher fuel enrichment.
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The~ LOCA codes were not consistent with. the fuel rmi design code relative to the fuel nd 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 nudels used in the LOCA Evaluation Models to maintain consisteney-with the latest approved version of the fuel rod design code, in addition, it was determined that integration of the cladding strain rate equation used in the large break LOCA ;
Evaluation Model, as described in Reference 5, was being calculated twice each time step instead of once. The coding was corrected tu properly integrate the strain rate equation.
Effect of Changes:
l~ The changes made to make the LOCA fuel rod models consistent with the fuel design codes were judged to be .
t 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 _
calctlations, incorporating only the cladding strain corrections were performed and the results suptwrted the
, conclusion that compensating effects were not presenti ne PCT effects reported below will bound the effects taken sei arately for the large break LOCA.
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' a)Large Break LOCA The effect of the changes on the large break LOCA peak cladding temperature was determined using the -
basil large break LOCA Evaluation Model. The effects were judged applicable to older Evaluation Models.
Several calculations were performed to assess the effect of the changes on the calculated results as follows:
y 1. Blowdown Analysisi It was determined that the changes will have a small effect on the core average rod and hot assembly average rod performance during the blowdown analysis. The effect of the changes on the blowdown 10
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Attachment 2 analysis was determined by performing a blowdown depressuritation computer calculation for a typical
- three-loop plant and a typical four loop plant using the SATAN-VI computer code.
- 2. Ilot Assembly Rod Heatup Analysis -
The hot rod heatup calculations would typically show the largest effect of the changes. Ilot rod heatup computer analysis calculations were performed using the LOCBART computer code to assess the effect of the changes on the hot assembly average rod, hot rod and adjacent rod,
- 3. Determination of the Effect on the Peak Cladding Temperature -
The effect of the changes on the calculated peak cladding temperature was determined by performing a calculation for typical three-loop and four loop plants using the BASH Evaluation hioel. The malysis calculations confirmed that the effect of the ECCS Evaluation hiodel changes were insignificant 'as defined by 10CFR50.46(a)(3)(i). The calculations showed dat the peak cladding temperatures incrrwed by less than by 10*F for the basil Evaluation hiodel. It was judged that 25'F would bound the effect on the peak cladding temperature for the BART Evaluation h'odel, while calculations performed for the Westinghouse 1981 Evaluation htodel showed that the peak cladding temperature could increase by approxinutely 41'F.
b) Small Break 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 hiodel by performing a computer analysis ciculations tot 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 hiodel were insignificant as defined by 10CFR50.46(a)(3)(i). The calculations showed that 37'F would bound the effect on the calculated peak cladding temperatures for the four-loop plants and the three-loop plants. It was judged that an increase of 37_*F would bound the effect of the changes for the 2-loop plants.
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Attachment 2 VI.' ShlALL 13REAK LOCA ROD INTi:RNAL PRilSSURE INITIAL CONDITION ASSUhiPTION <
I Change
Description:
The Westinghouse small break loss-of-coohmt accident (LOCA) emergency core cooling system (ECCS)
Evaluation Model analyses assume that I.igher fuel rod inhial fill pressure leads to a higher calculated peak cladding temperature (PCT), as found in studies with the Westinghouse large Neak LOCA UCCS Evahuition Model. llowever, lower fuel rod internal pressure could result in decreased cladding cierp (rod swelling) away from the fuel pellets when the fuel rod internal pressure was higher than the reactor soolant system (RCS) pressure. A lower fuel rod initial fill pressure could then result in a higher calculated peak cladding temperature.
The Westinghouse small break LOCA cladding strain model is based upon a correlation of liardy's data, ar described in Section 3.5.1 of Reference 5. Evaluation of the limiting fuel nxl initial fill pressure assumption revealed that this model was used outside of the applicable range in the small break LOCA Evaluation Model calculations, allowing the cladding to expand and contract more rapidly than it should. The model was corrected to fit applicable data over the range of small break LOCA conditions. Correction of the cladding strain model affects the small break LOCA Evaluation Model calculations through the fuel rod internal pressure initial condition assumption.
Effect of Changes:
Implementation of the corrected cladding ercep equation results in a snt .1 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 fuct pellet prior to core uncovery when the RCS pressure exceeds the rod internal pressure, implementation of the correlation for the appropriate range has a negligible benefit on the peak cladding temperature calculation during this portion of the
' ransient, llowever, after the RCS pressure talls below the rod internal pressure, implementation of an accurate correlation for cladding creep in small break LOCA enalyses would reduce the expansion of the cladding away from the fuel compared to what was previously cateulated 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 moditientions for the hmiting three inch equivalent diameter cold leg break in typical three-loop and four-loop plants. The results indicated that the-change to the calculated peak cladding 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 is,itial conditions, the combined effect of the fuel rod internal pressure and the cladding strain model revision is typicady bounded by 40'F. However, in an extreme case the combined effect could be as large as 60*F.
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Attachment 2 - )
l Vll, NOTRUh1P CODE SOI.UTION CONVERGENCE Change
Description:
~
In the development of the NOTRUh1P snudt break 1.OCA ECCS Evaluation hiodel, a number of miding sensitivity studies were performed to demonstrate acceptable solution convergence as required by Appendix K to 10CFR50. Temporal solution convergence sensitivity studies were performed by varying input parameters which govern the rate of change of key process variables, such as changes in the pressure, nuss, and internal energy, Standard input values were specified for the input parameters which govern the time step sim selection, llowever, since the initial studies, modification.< were made to th(NOTRUh1P computer program to enhance code perfornumee 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 NOTRUhlP computer code for variations in imtirl RCS prusure, auxiliary feedwater flow rates, power distribution etc.,
which resulted in peak cladding temperature (PCT) variations which were greater than anticipated bamt ulxm engineering judgement. ~ In addition, the direction of the PCT variation conflicted with engineering judgement expectations in some cases. The unexpected variability of the sensitivity study results indicated that the numerical solution may not be properly converged.
Sensitivity studies were performed for the time step sim selection criteria which culminated in a revision to the recommended time step size selection criteria inputs. Fixed input values originally reconunended for the steady state and all break transient calculations were modified to assure converged results. The NOTRUhiP code was re-verified against the SUT-08 Semiscale experiment and it was confirmed that the code adequately predicts key small break phenomena, Effect 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 representative calculations, however, this change will most lik.ely result in a reduction in the calculated peak cladding temperature. Since the potential beneficial effect of a non-converged solution is plant specific, a
- generic PCT ef fect cannot be provided, llowever, it has been concluded that current licensing basis results remain valid since the results are conservative relative to the change.-
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- Attachment 2 4,
Vill. LARGE BREAK LOCA 13URST AND ULOCKAGE ASSUhtPTION
Background:
The cladding swelling and flow bhickage moJels were reviewed in deta;l during the NRC's evaluation of the Westinghouse Evaluation Model. Ilowever, the use of the average rod in the hot assembly may not have been documented in a manner vetailed enough to allow the staff to adequately assess this aspect of the model. ,
-Appendix K to 10CFR50 requires consileration of the effects of flow blodage resulting from the swelling and rupture of the fuel rods dunng a loss of-coolant accident (l.OCA). 10CFR50 Appendix K Paragraph LB states:
"... To tu acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not underestimateo."
i in Westinghouse ECCS Evaluation Model calculations the average rod in the hot assembly is used as the basis for calculatmg the effects of flow blockage, if a significant number of fuel mds in the hot assembly are operating at pcwer levels grewer 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 6ince the lower power rod will take longer to heat up to levels where swelling and rupture will occur.
A review of the Westinghouse model used to predict assembly bkxkage was performed. This model was -
developed from the Westinghouse Multi-Rod Durst Tests (MRBT) and was the model used to determine assembly v!ae bhxkage until replaced by the NUREG-0630 model starting in 1980. These models provide the means for determining assembly whle bhickage once the mean nurst strain has been established. implementation of these burst models has relied upon the average rod to provide the mean burst strain. The average rod is a low power rod producing the power of the average of rods in the hot asse.nbly and is primarily used to calculate the enthalpy rise in the hot assembly. Use of the average rod in the model assumes that the time at which blockage is calculated to occur is represented by the burst of the average rod. A review of current hot assembly power distributions indicates that in general the average rod in the hot assembly is also representative of the largest number of rods in the assembly, so that burst of this rod adequately represents when most of the rods will burst.
With this representation, however, the true onset of blockage would likely begin earlier, as the highest power rods reach their burst temperature. This time is estimated to be a few seconds prior to the time when the average rod bursts.
Large break LOCA Evaluation Models which use BART or basil simulate the hot assembly rod with the actual average power, while older Evaluation Models use an average rod power which is adjusted downward to account 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 bhxkage penalty is calculated will be delayed for these older Evaluation Models.
Change Descriptiom Ample experimental evidence currently exists which shows that flow bhickage does not result in a heat transfer penalty during a LOCA. In addition, newer Evaluation Models have been developed and licensed which demonstrate that the older Evaluation Models contain a substantial amount of conservatism. Westinghouse j concluded that further artificial changes to the ECCS Evaluation Models to force the calculation of an earlier burst l- time were not necessary. In rare instances where burst has not occurred prior to the flooding rate falling below I 1.0-inch /second, the results of the ECCS analysis calculation are supplemented by a permanent assessment of '
l-margin. Typically thia will only occur in cases where the calculated PCT is low. Westinghouse concludes that no r
- model change is required to calculate an earlier burst time.
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Attachment 2 -
IX. STEAM GENLRATOR FLOW AREA
Background:
Licensees are normally required to provide assurance that there exists only an extremely low probability of abnormal leakage or gross rupture of any part of the reactor coolant pressure boundary (General design criteria 14 and 31). The NRC issued a regulatory guide (RO 1.121) which addressed this requirement specifically for steam generator tubes in pressurized water reactors. in that guide, the staff sequired analytical and experimental evidence that steam generator tube integrity will be maintained for the combinations of the loads resulting from a LOCA with the loads from a safe shutdown earthquake (SSE). These loads are combined for added conservatism in the calculation of structural integrity. This analysis provides the basis for establishing criteria for removing from service tubes which had experienced significant degradaticn.
Analyses performed by Westinghouse in support of the almve requirement for various utilities, combined the most severe LOCA loads with the pa..t *necific SSE, as delineated in the design criteria and the Regulatory Guide.
Generapy, these analyses showed that wnile tube integrity was maintained, the combined loads IcJ to some tube deformation. This deformation reduces the flow area through the steam generator. The reduced flow area increases the resistance through the steam generator to the flow of steam from the core during a LOCA, which i potentially could increase the calculated PCT, Change Descriptiom The effect of tube deformation and flow area reduction in the steam generator was analyzed and evaluated for rome plants by Westinghouse in the late 1970's and early 1980's. The combination of LOCA and SSE loads led to the following calculated phenomena:
- 1. LOCA and SSE loads cause the steam generator tube bundle to vibrate.
- 2. The tube support plates may be deformed as a result oflateral loads et the wedge supports at the periphery of the plate. The tube support plate deformation may cause tube deformation.
- 3. During a postulated large LOCA, the primary side depressurizes to containment pressure. Applying the resulting pressure differential to the deformed tubes causes some of these tubes to collapse, and reduces the effective flow area through the steam generator.
- 4. The reduced flow area increases the resistance to venting of steam generated in the core during the reflood phase of the LOCA, increasing the calculated peak cladding temperature (PCT).
The ability of the steam generator to continue to perform its safety lunction was established by evaluating the effect of the resulting flow area reduction on the LOCA PCT. The postulated break examined was the steam generator outlet break, because this break was judged to result in the greatest loads on the steam generator, und 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.
-In April of 1990, in considering the effect of the combination of LOCA + SSE loadings on the steam generator component, it v.ts determined that the potential for flow area reduction due to the contribution of SSE loadings should be inchied in other LOCA analyses. With SSE loadings, flow area reduction may occur in all steam generators (nct just the faulted loop). Therefore, it was concluded that the effects of flow area reduction during the most limiting primary pipe break affecting LOCA PCT, i.e., the reactor vessel inlet break (cold leg break LOCA), had to be evaluated to confirm that 10CFR50.46 limits contime to be met and that the affected steam generators will continue to perform their intended safety function.
15
Attachment 2 Coneguently, the action was taken to address the safety sigmficance of strum genciator tube collapse dusing a cold leg break LOCA. The effect of flow area reduction from combined 1.OCA and SSE loads was estimated.
The magnitude of the flow area reduction was considerni equivalent to an mereased level of steam yenerator tube plugging. Typically, the asca reduction was estimatal to range from 0 to 7.55, depending on the magnitude of l' the seismic loads. Since detailed non-linear seismic analyses are not available for Scrica $1 and earlier design
[-
steam generators, some area reductions had to be estimated based on available infornation. For most of these plants, a 5 percent flow area reduction was assumed to occur in each steam generator as a result of the SSE. For
, these evaluations, the contribution of loadings at the tube support plates fmm the LOCA cold leg break was assumed negligible, since the add.tional 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 generatoru must be able to withstand the effects of coinbined LOCA 4 SSE loadings and continue to perform their intended safety function. It is judged that this requirement upphes to undegraded as well as locally degradcJ steam generato tubes. Compliance with GDC 2 is addressed below for both conditions, For tubes whi:h have not experienced craeking at the tube- support plate elevations, it is Westinghouse %
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 salety lunctions of the steam genciator tubes are to provide a flow path for the venting of steam genesated in 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 esent.
Tube deformation has the same effect on the LOCA event as the plugging of steam generator tubes. The effect of tube deforma' ion and/or collapse can be taken into account by assigning an appropriate PCT penalty, or acccunting for the' area seduction directly in the analysis. Evaluations completed to date show that tube deformation results ir acceptable LOCA PCT. From a 1.tcam generator structural integrity perspective, Section 111 of tne ASME Code recogmia that inelastie deformation can occur for faulted condition loadingo There are no requirements that equate steam generetor tube deformation, per se, with loss of safety funetion. Cros.s-sectional
- lunding stresses in the tubes at the tube support plate _ elevations are considered secondary stresses within the definitions of the ASME Code and need not be considered in establishing the limits for allowable steam generator tube wall degradation. Therefore, for undegraded tubes, for the expected degree of _ flow area seduction, and despite the calculation showing potential tube collap.<c for a limited number of tubes, the steam generators continue to perform their tequired safety functions after the combination of LOCA + SSB loads, meeting the requirements of GDC 2.
During a November 7,1990 meeting with a utility and the NRC staff on this subject, a concern was raised that tubes with partial wall cracks at the tube support plate elevations could progress to through. wall cracks during tube deformation. This may result in the potential for significant secondary M primary inleakage during a LOCA event; it was noted that inleakage is not addressed in the existing liCCS ana ysis. Westinghouse did not consider
- the potential for secondary to primary inleakage during resolution of the ste4 m generator tube collaps- item. This is a relatively new item, not previously aoJressed, smce cracking at the tuce support plate elevations had been insignificant in the early 1980% when the. tune collapse item was evaluatsd in depth. There is amplo data available which demonstrates that undegraded tubes maintain their integrity mder collapse loads. There is also some data whi h shows that cracked tubes do not behave significantly differently fmm uncracked tubes when collapse loads are applied. Ilowever, cracked tube data is available only for round or slightly ovahzed tubes.
I It is important to recognize that the core melt frequency resulting imm a combined LOCA + SSE event, subsequent tube collapse, and significant steam generator tube micakage is very low, on the order of 104/RY or less. This estimate takes into account such tactors as the possibility of a seismically induced LOCA, the expected l-occurrence of cracking in a tube as a function of height in the steam generator tube bundle, the locahzed ef fect of i
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^
Attachment 2 '
the tube support plate defornution, and the possibility that a tube which is identified to deform during LOCA +
SSU loadings would also contain a partial through-wall crack which would result in significant inleakage. To further reduce the likelihood that cracked tubes would be subjected to collapse loads, eddy current inspection requirements can be established. The inspection plan would reduce the potential for the presence of cracking in the regions of the tube support plate elevations near wedges that are most susceptible to collapse which may then lead to penetration of the primary pressure boundary and significant inleakage during a LOCA + SSE event.
Change
Description:
As noted above, detailed analyses which provide an estimate of the degree of How area reduction due to both seismic and LOCA forces are not available for all steam generators. The information that does exist indicates that the Dow 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 e>timate the Dow 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 How area reduction has not been determined for some earlier design steam generators, the risk -
associated with flow area reduction and tube leakage from a combined seismic and LOCA event has been shown
' to be exceedingly low.13ased 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 defornution.
The effect of potential steam generator area reduction on the cold leg break LOCA peak cladding temperature has z been either analyud or estimated for each Westinyhouse 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 additional tube plugging in the xitalysis.
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Attachment 2 X. AUXILIARY FEI!DWATiiR ENTilALPY SWITCilOVliR FOR SBLOCA ANALYSIS
' Change
Description:
During a review of Westinghouse SilLOCA unalysis methods, a question arose with respect to the computer cale 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 SHLOCA ECCS Evaluation nnlels using either the WFLASil or NOTRUMP analysis technologies, this time is used to swi'ch the enthalpy of the fluid provided to the steam generators fmm the main
-- feedwater enthalpy to the auxiliary feedwater enthalpy. ,
Ef fect of Change:
A review and investigation of the concern indicated that, in some instances, the time assumed for.the auxiliary fe:dwater enthalpy purge delay time was shorter than times calculated from the actual plant configuration. The inconsisteney between the Westinghouse SULOCA ECCS Evaluation Model input value and a value -
correspmding to the plant configuention results from the specifie 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 recornmended. In the NOTRUMP analysin methodology, a standard input value judged to be conservative based upm phenomena observed during experiment SUT-08 in the Semiscale test facility was used.
. Ilowever,- further investigation showed that the standard input value could result in a non-conservative calculation i of the peak cladding temperature.
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is I 1
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Attachment 3 REFERENCES 1.
NS-NRC-89 3464
- Correction of Errors and hiodifications to the NOTRUh1P Code in the Westinghouse Small Break LOCA ECCS Evaluation hiodel Which Are Potentially Significant," Letter from W. J. Johnson (Westinghouse) to T. E. h1urley (NRC), Dated October 5,1989 2.
WCAP-9220-P A, Revision 1 (Proprietary), WCAP-9221 A, Revision 1 (Non-Proprietary),
- Westinghouse ECCS Evaluation htoJel - 1981 Version,' 1981, EichelJinger, C.
3.
WCAP-10266-P-A, Revision 2 (Proprietary), WCAP-10267-A, Revision 2 (Non-Proprietary), Besspiata,J.J.,
et.al., *1981 Version of the Westinghouse ECCS Evaluation h1odel Using the basil Code," h1 arch 1987.
4.
WCAP-10054 P-A (Proprietary), WCAP-10081+A (Non-Proprietary), ' Westinghouse Small Break ECCS Evaluation hiodel Using the NOTRUh1P 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-l Proprietary), h1 arch 1984.
7.
- 10CFR50.46 Annual Notification for 1989 of hiodifications in Westinghouse ECCS Evaluation Afodels,"
NS-N RC-89-3463, Letter from W. J. Johnson (Westinghouse) to T. E. h1urley (NRC), Dated October 5, 1989.
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Attachment 3 REFERENCES 1, NS NRC-89 3464
- Correction of Errors and hiodifications to the NOTRUh1P Code in the Westinghouse Small Break LOCA ECCS Evaluation hiodel Which Are Potentially Significant," letter from W. J Johnson (Westinghouse) to T. E. h1urley (NRC), Dated October 5,1989,
- 2. WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Recision 1 (Non-Psoprietary), " Westinghouse ECCS Evaluation hiodel- 1981 Ver, ion," 1981, Eicheldinger, C,
- 3. WCAP-10266-P-A, Revision 2 (Proprietary), WCAP 10267-A Revision 2 (Non-Proprietary), Besspiata,J.J.,
et.al., *1981 Version of the Westi.ighouse ECCS Evaluation hiodel Using the Basil Code," h! arch 1987.
- 4. WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), ' Westinghouse Small Break ECCS Evaluation hiodel Using the NOTRUh1P Code," Lee, N., et. al., August 1985.
- 6. "bART-Al A Computer Code for the Best Estimate Analysis of Refk>od Transients', WCAP-9695-A (Non-Proprietary), h1 arch 1984.
- 7. "10CFR50.46 Annual Notification for 1989 of blodifications in Westinghouce ECCS Evaluation hiodels,"
NS-NRC-89-3 463, Letter from W. J. Johnson (Westinghouse) to T. E. hiurley (NRC), Dated October 5, 1989.
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