ML20086H946
| ML20086H946 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 12/06/1991 |
| From: | Skolds J SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9112100201 | |
| Download: ML20086H946 (24) | |
Text
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.. South Car:Hna E4ctric & oas CImpany 10CMM,b P.O Box 08 Vice President
.. '[i . JinMnsviHL SC 29065 Nuclear Operations (803) 345-4040
..SCUSG amaww DEC o S399; Document Concrc'.. Desk U.-S. Nuclear Regulatory Commission
. Washington, 0, C. 20555 Gentlemen:
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Subject:
. VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO.-50/395-OPERATING LICENSE NO. NPF-12
-ECCS EVALUATION MODEL REVISIONS REPORT Attacned-is:the annual Emergency Core Cooling System (ECCS) Evaluation Model
- Revisions Report-for the Virgil C. Summer Nuclear Station (VCSNS). This report is pursu;nt to 10CTJ50.46 which requires licensees to notify the NRC on an'annuel basis of errors or changes in the ECCS Evaluation Models. In a, addition, this k tter fulfills the 10CFR50.46 (a)(3)(ii) requirement to
- repcrt significing (,r.anges or errors within 30 days. Significant or potentially- s!gnificant issues are described in the attachment.
Tables 2= ", of the attachment summarize the changes in peak clad. temperature for the prev 1ous analysis of record (Vantage'5 core) and the current analysis
, of1 record (Vantage + IIRLO core). None of the individual model changes is considered significant, however, the sum total of the changes for Table 4 does :aeet the definition of significant. The VCSNS ECCS analyses are up to date and were approved by NRC as part of the ZIRLO Technical Specification change, Amendment 105.-
I declare-that the statements and matte s set forth-herein are true and correct to the best of my knowledge, information, and belief.
'If:you have any questions, please call.
^
Very truly y urs, l
. 20fu~l John L. Skolds
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' c: 0 W. Dixoa, Jr. NRC Resident Inspector 4 ~R. R.' Mahan- J. B. Knotts, Jr.
.R. J. White L. R. Cartin S.'D. Ebneter NSRC G.-F. Wunder RTS (ANN 2200)
General Managers File (813.12-4, 818.02-17) ,
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NUCLEAR EXCELLENCE - A SUl#lER TRADITION! I 913 fDR h$Ijf '
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4 C11ANGES TO Tite WESTIN.110VSE ECCS EVALUATION MODELS AUGUST 1990 - MAY 1991 i
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CHANGES TO Tile WESTING 110US*. ECCS EVALUATION MODELS
1.0 INTRODUCTION
Provisions in 10CFR50.46 require the reporting of corrections to or changes in the ECCS Evaluation Model (EM) approved for use in performing safety analyses for the loss of coolant accident (LOCA). This report describes corrections and revisions to the Westinghouse ECCS EM in the period from August 1990 through May 1991 which are applicable to V. C. Summer Nuclear Station (VCSNS). The current Westinghouse ECCS ems are listed in Table 1, and consist of several computer codes with specific functions.
Westinghouse has completed the evaluation of several items related to the Westinghouse ECCS Evaluation Models listed in Table 1. Each of these items is discussed in the following sections, which include a description of the item, the assessment which was performed, the resulting change to the Evaluation Model, and the effect of the change on the Peak Clad Temperature (PCT).
Some of the subjects discussed represent chang 2s to program coding or to inputs directly related to the physical models or solution technique. These are described in Section 2.0.
Some items represent changes to the assumptions made when the Evaluation Model is applied to a specific plant. Applicable changes for VCSNS are discussed in Section 3.0. Also included, for information, are items fo-which a technical assessment is continuing, and items fur which it was concluded that no change was necessary.
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TABLE 1
SUMMARY
OF WESTINGHOUSE ECCS EVALUATION MODELS FOR VCSNS NAME: 1981 MODEL WITH BASH APPLICATION: Analysis of Large Break LOCA CODES USED: PURPOSE:
REFERENCE:
SATAN-VI Blowdown hydraulic transient 1, 6 BASH Reflood hydraulic transient 8 LOCBART Hot assembly thermohydraulics 3,7,8 and fuel rod thermal transient WREFLOOD/C0CO/LOTIC Containment pressure transient 2,4,5,8 N_ME: 1985 SBLOCA HODEL APPLICATION: Analysis of Small Break LOCA CODES USED: PURPOSE:
REFERENCE:
NOTRUMP System Hydraulic transient 12, 13 SBLOCTA fuel rod thermal transier.t 3
.b ,
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2.0 EVALUATION H0 DEL CODE CHANGES This section describes changes and revisions to the Westinghouse ECCS Evaluation Model computer codes. Except where noted, these corrections will be implemented in all future applications of the Evaluation Model. ,
2.1 FUEL R0D MODEL REVISIONS During the review of the original Westinghouse ECCS Evaluation Model following the promulgation of 10CFR50.46 in 1974 Westinghouse committed to maintain consistency between future loss-of-coolant accident (LOCA) fuel rod computer models and the fuel rod design computer models used to predict fuel rod normal operation performance. These fuel rod design codes are also used to establish initial conditions for the LOCA analysis.
Chance
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.
- 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 codes to maintain consistency with the latest approved versinn 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 code, as described in Reference 3, was being calculated twice each time step instead of once. The coding was corrected to properly integrate the strain rate equation.
Affected Evaluation 'iodels:
1981 Large Bret.k LOCA Evaluation Model, With BASH 1985 Small Break LOCA Evaluation Model Effect of Chances:
The changes made to make the LOCA codes consistent with the fuel design codes were judged to be insignificant, as defined by 10CFR50.46(a)(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 code, additional calculations, incorporating only the cladding strain corrections were performed and the results show that no compensating effects are present.
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l The PCT effects reported below will thus bound the effects taken separately fo" the large break LOCA, a) Large Break LOCA The effect of the changes on the large break LOCA peak cladding temperature was generally determined using the BASH large break LOCA Evaluation Model. Several calculations were performed to assess the effect of the changes on the calculated results as follows:
- 1. Slowdown Analysis -
It was de+ ermined that the changes will have a small effect on the core average rod and hot assembly average rod performance during the blowdown analysis. The effect of the changes on the blowdown analysis was determined by performing a blowdown depressurization computer calculation for a typical three-loop plant and a typical four-loop plant using the SATAN-VI computer code.
- 2. Hot Assembly Rod Heatup Analysis -
The hot assembly rod heatup calculations would typically show the largest effect of the changes. Hot assembly rod heatup computer analysis cciculations were performed using the LOCBART computer code to assess the ef fect of the changes on the hot assembly average rod, hot rod and adjacent rod.
- 3. Determination of the Effect on the Peak Cladding Temperature The effect of the changes on the calculated peak cladding temperature was determined by performing a calculation for typical three-loop and four-loop plants using the BASH Evaluation Model. The analysis calculations confirmed that the -
effect of the ECCS Evaluation Model changes were insignificant as defined by 10CFR50.46(a)(3)(1). The calculations showed that the peak cladding temperatures increased by less than 10"F for the BASH Evaluation Model, b) Small Break LOCA The effect of the cr anges on the small break LOCA analysis peak cladding temperature calculations was determined using the 1985 small break LOCA Evaluation Model by performing computer analysis calculations for a typical three-loop plant and a typical four-loop plant. The analysis calculations confirmed that the effect of the changes on the small break LOCA ECCS Evaluation Model were insignificant as defined by 10CFR50.46(a)(3)(1). The calculations showed that 37*F would bound the effect on the calculated peak cladding temperature for three-loop plants.
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- i Status:
These changes have been implemented and utilized in VCSNS specific analysis submitted to NRC in Reference 17 (Appendix G).
2.2 SMALL BREAK LOCA R0D INTERNAL PRESSURE INITIAL CONDITION ASSUMP110N Change
Description:
The Westinghouse small break loss rf-coolant accident (LOCA) emergency core cooling system (ECCS) Evaluation Model analyses assume that higher fuel rod initial fill pressure leads to a higher calculated peak cladding 'M erature (PCT), as found in studies with the Wistinghouse large break ;. A ECCS Evaluation Model. However, lower fuel rod internal pressure could result in decreased cladding creep (rod swelling) away from the fuel pellets when the fuel rod internal pressure was higher than the reactor coolant system (RCS) pressure. A lower fuel rod initial fill pressure could then result in a higher calculated peak cladding temperature.
The Westinghouse small break LOCA cladding strain model is based upon a correlation of Hardy's data, as descrDed in Section 3.5.1 of Reference 3.
Evaluation of the limiting fuel rod initial fill pressure assumption reveaTed that this model was used outside of the applicable r?nge in the small break LOCA Evaluation Model calculations, allowing the cladding to expand and contract more rapidly than it should. The model was corrected to fit applicable data over the range of small break LOCA conditions. Correction of the cladding strain model affects the small break LOCA Evaluation Model calculations through the fuel rod internal pressure initial condition assumption.
Affected Evaluation Model:
1985 Small Break LOCA Evaluacion Model Effect of Changes:
i Implementation- of the corrected cladding creep equation results in a small reduction in the pellet to cladding gap when the RCS pressure exceeds the rod
, internal pressure and increases the cap af ter RCS pressure f alls below the rod internal pressure. Since the cladding typically demonstrates very little creep toward the fuel pellet prior to core uncovery when the RCS pressure exceeds the rod internal pressure, implementation of the correlation for the appropriate range has a negligible benefit on the peak cladding temperature calt.ulation during this portion of the transient, However, after the RC$
pressure falls below the sod internal pressure, implementation of an accurate corr 21ation for cladding creep in small break LOCA analyses would reduce the -
expansion of the cladding away from the fuel compared to what was previously l calculated and results in a PCT penalty because the cladding is closer to the j fuel.
l Generic calculations were performed- to assess the effect of the cladding strain modifications for the limiting three-inch equivalent diameter cold leg break in typical three-loop and four-loop plants. The results indicated that l Page 5 of 22 1
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 initial conditions, the combined effect of the fuel rod internal pressure and the cladding strain model revision is typically bounded by 40*F. However, in an extreme case the combined ef fect could be as large as 60*F.
, Status-j Modi ' cations to the small break LOCA cladding strain model for application to appropriate range of conditions have been implemented and the effect of the rod internal pressure initial condition assumption assessed. Changes to the Strain model affects assumptions concerning the limiting time in the core cycle. The effects have been examined in burn up studies for VCSNS and are documented in Reference 17 (Appendix G).
2.3 NOTRUMP CODE SOLUTION CONVERGENCE Chance
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 convergence as required ty Appendix K to 10CFR50. Temporal solution convergence sensitivity studies were performed by varying input parameters which govern the rate of change of key process variables, such as changes in the pressure, mass, and internal energy. Standard input values were specified for the input parameters which govern the time step size selection.
- However, since the initial studies, modifications were made to the NOTRUMP computer program to enhance code performance and implement necessary modifications (Reference 15). Subsequent to the modifications, solution
- convergence was not re-confirmed.
To analyze changes in plant operating conditions, sensitivity studies were performed with the NOTRUMP computer code for variations in initial RCS pressure, auxiliary feedwater flow rates, power distribution, etc., which resulted in peak cladding temperature (PCT) variations which were greater than anticipated based upon engineering judgment. In addition, the direction of the PCT variation - conflicted with engineering judgment expectations in i some cases. The unexpected variability of the sensitivity study results
. indicated that the numerical solution may not be properly converged.
Sensitivity studies were performed for the time step size selection criteria which culminated in a revision to the recommended time step size selection L criteria inputs. Fixed input values originally recommended for the steady state and all break transient calculations were modified to assure converged results. The NOTRUMP code was re-verified against the SUT-08 Semiscale experiment and it was confirmed that the code. adequately predicts key small break phenomena.
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1985 Small Break LOCA Evaluation Model i l
Generic Effect of Changes:
Generally, the modifications result in small shifts in 19.i ng of core uncovery and recovery. However, these changes may result in a change in the calculated peak cladding temperature which exceeds 50*F for some plants.
Based on representative calculations, however, this change will most likely result in a reduction in the calculated peak cladding temperature. Since the potential beneficial effect of a non-converged solution is plant specific, a generic PCT effect cannot be provided. However, it has been concluded that current licensing basis results remain valid since the results are conservative relative to the change.
Status:
This change has been implemented and was utilized in the VCSNS analysis submitted to the NRC via Reference 17 (Appendix G).
2.4 GAMMA ENERGY DEPOSITION MODEL REVISION Change
Description:
The Westinghouse small break LOCA ECCS Evaluation Model incorporating the NOTRUMP analysis technology was revised to permit specification of the gamma energy deposition factors appropriate to the fuel assembly type, in the hot assembly fuel rod heatup calculations performed with the small break version of LOCTA-IV. Gamma energy deposition factors calculated in accordance with the methods defined in Reference 16 can be specified by the analyst for the specific type of fuel and core design. parameters being analyzed.
Affected Evaluation Model:
1985 Small Break LOCA Evaluation Model Generic Effect of Chanqes:
! There is no effect for the change unless the gamma energy deposition factors I differ from those specified in Reference 3. Gamma energy deposition factors which differ from those specified in Reference 3 are justified by specific
, calculations of the fuel and core design parameters and the various energy l- sources in the core. The affect of values which differ from those specified in Reference 3 depends upon the thermal-hydraulic transient response and the magnitude of the difference. A calculation was performed for a typical l- three-loop plant in which the gamma energy deposition factors were reduced by approximately 2% for burnup conditions higher than the beginning of life (BOL). The effect on the cladding temperature calculation was determined by examining the hot assembly average rod which indicated that the temperature was reduced by approximately 30*F.
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Status:
Modifications to the small break LOCA hot assembly fuel rod heatup calculations performed with the small break version of LOCYA-IV were made to permit specification of gamma energy deposition factcrs as input by the l analyst. Gamma energy deposition factors will default to the values gecified in Reference 3 if specific gamma energy deposition input is not l provided. This modification has been implemented. Gamma energy deposition !
factors calculated in accordance with the methods defined in Reference 16 l have been applied in the VCSNS analysis provided in Reference 17 (Appendix G),
2.5 REVISIONS TO REPRESENT ZIRLO MATERIAL PROPERTIES Chance
Description:
The cladding models which could affect the LOCA ECCS Evaluation Model analyses include the cladding specific heat, the high-temperature creep (swelling), the rupture criteria, and the circumferential strain following rupture. A brief summary of each of the models is provided in the following.
Clad Specific Heat Phase changes have a pronounced effect on clad spec'fic heat capacity, and the phase changes in ZIRLO cladding material occur at different temperatures than for the Zircaloy-4 claddir.g material. Models for the specific heat were developed as a function of temperature for use in the safety analyses. The development of the 21RLO specific heat model is discussed in detail in Appendix A of Reference 17. The resulting specific heat model was incorporated as part of a ZIRLO cladding model option.
Hioh-Temperature Creep
- The high-temperature creep behavior of ZIRLO cladding material was developed based upon tests for the various phases of the material. The measured creep rates were correlated as functions of hoop stress and i temperature. Details of the creep tests and the verification of the creep model for the constant temperature and constant pressure conditions of the tests are given in Appendix C of Reference 17. This
, model was incorporated as part of a ZIRLO cladding model option.
l Rupture Temperatures
, ZlRLO clad rod burst tests were performed and a model devtloped to l represent the rupture behavior pnenomena in the LOCA analyses. The test
- results were correlated in the form of burst temperature as a function l of engineering hoop stress. Octails of the burst tests and the development of the burst temperature correlation are given in Appendix D of Reference 17.
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Circumferential Strain Followina Rupture The circumferential strains at the rupture elevations were measured in
.the ZlRLO_ rupture tests and a correlation of rupture strain as a function of rupture temperature was developed and used if burst is calculated to occur.
Affected Evaluation Models:
1981 Large Break LOCA Evaluation Model with BASH 1985 Small Break LOCA Evaluation Model Effect of Chances:
The effect of the use of ZIRLO cladding on the VCSNS ECCS analysis results is described in Reference 17.
4 Status:
ZIRLD was approved for use at VCSNS in Amendment 105 to the Technical c - Specifications. To support this Technical Specification change, the ZIRLO
- modifications were implemented and utilized in the VCSNS specific analysis
~ described in Appendix G to Reference 17.
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4 3.0 EVALUAlION H0 DEL APPLICATION CHANGES The following section describes changes in the way the LOCA evaluation model is applied, or provides additional information on the method of application.
3.1 LARGE BREAK LOCA POWER DISTRIBUTION ASSUMPTION
Background:
Appendix K to 10CFR50 requires that the power distribution which results in the most severe calculated consequences be used in the ECCS Evaluation Model calculations. The power distributions to be studied are those expected to occur during the core lifetime.
The current basis for all Westinghouse large LOCA Evaluation Models is the chopped cosine power distribution. This distribution is symmetrical and is defined by two quantitles: the ratio of peak linear power relative to the average (FQT), and the ratio of hot rod integral power relative to the average (FAH). This power distribution was found to produce the highest peak cladding temperature (PCT) when compared to power distributions skewed to the top or bottom of the : ore in studies performed by Westinghouse and submitted to the NRC. Typically the power distritutions were assumed to peak at discrete elevations in the core (4, 6, 8, and 10 feet). It was also assumed that the key parameters affecting PCT were the FQT, FAH, the peak pce.er location, and integral of power to the peak power elevation.
Calculations performed with the advanced LOCA Evaluation Models, BART and BASH, which examined peak power locations and power distributions which were not considered in the original analyses, under some circumstances lead to PCTs greater than those calculated with the cosine distribution. This behavior was revealed when performing power distribution studies for core designs with relatively low FQT and relatively high FaH. Further studies revealed that, in addition to FQT , FAH, and the peak power location, the nature of the axial distribution of power affected the results. Inat is, two power distributions with the same FQT, FAH, and peak power location, but whose power was distributed differently along the rod could result in significantly different PCTs.
- Westinghouse has completed an analysis effort to understand and properly account for the effect of skewed power distributions on the calculated large break !0CA PCT. This effort included the identification of the worst power distributions that could occur during core life with full consideration of the current generation of reload core designs.
Change
Description:
As a result of these studies, revisions have been made to the current reload and safety analysis methodology which accounts for the variability in power distributions from cycle to cycle and plant to plant. This revision provides a means of determining that the current licensing basis (i.e., the chopped cosine) is expected to remain limiting, but also provides for identifying and analyzing the most severe expected power distribution, if dif ferent from the chopped cosine.
Page 10 of 22
Affected Evaluation Model:
1981 ECCS Evaluation Model with BASH Status:
In Order to verify that VCSNS was not affected by this item, a reduction in the Core Operating Limits Report K(Z) curve was implemented for fuel cycle 6 and 7. This assures that more limiting power shapes do not occur, thus assuring that 10CFR50.46 limits would be met.
The process descr'ibed in Reference 18 will be used to assess future core designs. In this process, each power distribution calculated in the core design will be evaluated to determine whether it is more limiting than the cosine power distribution. Adjustments will be made to the core design operating bands to eliminate these limiting distributions and surveillance factors will be defined to assure that plant safety limits are met. This will assure that a change to the ECCS Evaluation Model is not required, since the choppad cosine power distribution will remain limiting.
3.2 LARGE BREAK LOCA BURST AND BLOCKAGE ASSUMPTION
Background:
The cladding swelling and flow blockage models were reviewed in detail during
- the NRC's evaluation of the Westinghouse Evaluation Model. However, the use of the average rod in the hot assembly may not have been documented in a manner detailed enough to allow the staff to adequately assess this aspect of the model.
Appendix K to 10CFR50 requires consideration of the effects of flow blockage rcsulting from the swelling and rupture of the fuel rods during a loss-of-coolant accident ;LOCA), 10CFR50 Appendix K Paragraph I.B states:
"...To be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not underestimated."
In Westinghouse ECCS Evaluation Model calculations, the average rod in the hot assembly is used as the basis for calculating the effects of flow blockage. If a significant number of fuel rods in the hot assembly are operating at power levels greater than that of the average rod, the time at which cladding swelling and rupture is calculated to occur may be predicted later in the LOCA transient, since the lower power rod will take longer to heat up to levels where swelling and rupture will occur.
A review of the Westinghouse model used to predict assembly blockage was performed. This model was developed from the Westinghouse Multi-Rod Burst Tests (MRBT) and was the model used to determine assembly wide blockage until replaced by the NUREG-0630 model starting in 1980. These models provide the means for determining assembly wide blockage once the mean burst strain has been established. Implementation of these burst models has relied upon the average rod to provide the mean burst strain. The average rod is a low power Page 11 of 22
rod producing the power of the average of rods in the hot assembly and is 1
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 BASH simulate the hot assembly rod with the actual average power, while older Evaluation Models use an average rod power which is adjusted downward to account for thimbles (this ~
is described in detail in Addendum 3 to reference (7)). If burst occurs af ter the flooding rate has fallen below one inch per second, the time at which the blockage penalty is calculated will be delayed for these older Evaluation Models.
Change
Description:
Ample experimental evidence currently exists which shows that flow blockage does not result in a heat transfer penalty during a LOCA. In addition, newer Evaluation Models have been developed and licensed which demonstrate that the older Evaluation Models contain a substantial amount of conservatism.
Westinghouse concluded that further artificial changes to the ECCS Evaluation Models to force the calculation of an earlier burst time were not necessary.
In rare instances where burst has not occur red prior to the flooding rate falling below 1.0-inch /second, the results of the ECCS analysis calculation are supplemented by a permanent assessment of margin. Typically th-is will only occur in cases where the calculated PCT is low. Westinghouse concludes that no model change is required to calculate an earlier burst time.
Affected Evaluation Model:
1981 ECCS Evaluation Model with BASH Status:
Complete.
3.3 STEAM GENERATOR 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 (RG 1.121) which addressed this requirement specifically for steam generator tubes in pressurized water reactors. in that guide, the staff required analytical and experimental evidence that steam generator tube integrity will be maintained for the combinations of the loads resulting from a LOCA with the loads from a safe shutdown earthquake Page 12 of 22
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(SSE). _These loads are combinad 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 i degradation.
Analyses performed by Westinghouse in support of the above requirement for various utilities.. combined the most severe LOCA loads with the plant specific SSE, as delineated in the design criteria and the Regulatory Gaide.
Generally, these analyses showed that while tube integrity was maintained, the combined loads led to some tube deformation. This deformation reduces the flow area through the steam generator. The r sced flow area increases the resistance through the steam generator to the t ow of steam from the core during a LOCA, which potentially could increase the calculated PCT..
The effect of tube deformation and flow area reduction in the steam generator was analyzed and evaluated for some plants by Westinghouse in the late 1970s and early 1980s. 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 of lateral loads at the wedge supports at the periphery of the plate. The tube support plate deformation may cause tube deformation.
- 3. During a postula'ted 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 function was established by evaluating the effect of the resulting flow area reduction on the. LOCA PCT.
The postulated break examined was the steam generator
- ._ outlet - break,- because this break was judged to result in the greatest loads on the steam generator, and thus the greatest flow area reduction. It was concluded that the steam generator would continue to meet its safety function because 'the degree of flow area. reduction was small, and the postulated break at the steam generator outlet resulted in a low PCT.
i; In April of 1990, in considering the effect of the combination of LOCA + SSE loadings on the _ steam generator component, it was determined that the potential for flow area reduction due to the contribution of SSE loadings should be included in other LOCA analyses. With SSE loadings, flow area l1 reduction may occur' in all steam generators (not just the faulted loop).
Therefore, it was concluded that the effects of flow area reduction during l
the most limiting primary pipe break affecting LOCA PCT, i.e. , the reactor l: v:5 del inlet break (cold leg break LOCA), had to be evaluated to confirm that l 10CFR' 50.45 limits continue to be met and that the affected steam generators will continue to perform their intended safety function.
I i Page 13 of 22 i
l Consequently, the action was taken to address the safety significance of steam generator tube collapse during a cold leg break LOCA. The effect of flow area reduction from combined LOCA and SSE loads was estimated. The magnitude of the flow area reduction was considered equivalent to an increased level of steam generator tube plugging. Typically, the area reduction was estimated to range from 0 to 7.5%, depending on the magnitude of- the scismic loads. Since detailed non-linear seismic -analyses are not availt.ble for all steam generators, some area reductions had to be estimated based on available information.. For most of these plants, a 5 percent flow area reduction was assumed to occur in each steam generator as a result of the SSE. For these evaluations, the contribution of loadings at the tube support plates from the LOCA cold leg break was assumed egligible, since the additional area reduction, if it occurred, would occur only in the broken loop steam generator.
Westinghouse recognizes that, for most plants, as required by GDC 2, " Design Basis ici protection against Natural Phenomena", that steam generators must be role to withstand the effects of combined LOCA + SSE loadings and continue to perform their intended safety function. It is judged that this requirement applies to undegraded as well as locally degraded steam generator tubes. Compliance with GDC 2 is addressed below for both conditions.
For _ tubes which have not experienced cracking at the tube support plate elevations, it is Westinghouse's engineering judgment that the calculation of steam generator tube deformation or collapse as a result of the combination of LOCA loads with SSE loads does not conflict with the requirements of GDC 2.
During a large break LOCA, the intended safety functions of the steam generator tubes are to provide a flow path for the venting of steam generated in the core through the RCS pipe break and to provide a flow path such that the other plant systems can perform their intended safety functions in mitigating the LOCA event.
Tube deformation has the same effect on the LOCA event as the plugging of steam generator tubes. The effect of tube deformation and/or collapse can ~ be taken into account by assigning an appropriate PCT penalty, or accounting for the area reduction directly in the analysis. Evaluations completed to date show that tube deformation results in acceptable LOCA PCT. From a steam generator structural integrity perspective,Section III of the ASME Code recognizes that inelastic deformation _can occur for faulted condition loadings. There are no requirements that equate -steam generator tube deformation, per se, with loss of safety function. Cross-sectional bending stresses in the tubes at the tube support plate elevations are considered secondary stresses within the definitions of the ASME Code and need not be considered in establishing the limits for allowable steam generator tube wall degradation._ Therefore, for undegraded - tubes, for the expected degree of flow area reduction, and despite the calculation showing potential tube collapse for a limited number of tubes, the steam generators continue ' to perform their required safety functions af ter the combination- of LOCA + SSE loads, meeting the requirements of GDC 2.
p 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
- l- support plate elevations could progress to through-wall cracks during tube
! deformation. This may result in the potential for significant secondary to l
l Page 14 of 22
primary inleakage during a LOCA event; it was noted that inleakag is not. l addressed in the existing ECCS analysis. Westinghouse did not consider the potential for secondary to primary inleakage during resolution of the steam generator tube collapse item. This is a relatively new item, not previously addressed, since cracking at the tube support plate elevations had been insignificant in the early 1980's when the tube collapse i tem was evaluated in depth. There is ample data available which demonstrates that undegraded
, tubes. maintain their integrity under collapse loads. There is also some data which shows that cracked tubes do not behave significantly dif ferently from
, uncracked tubes when collapse loads are applied. However, cracked tube data is available only for round or slightly ovalized tubes.
It is important to recognize that the core melt frequency resulting from a combined LOCA + SSE event, subsequent tube collapse, and significant steam I generator tube inleakage is very low, on the order of 10-8/ Reactor Year or l
less. This estimate takes into account such factors as the possibility of a seismically induced LOCA, the expected occurrence of cracking in a tube as a function of height in the steam generator tube bundle, the localized effect of the tube support plate deformation, and the possibliity that a tube which
- is identified to deform during LOCA + SSE loadings would also contain a partial through-wall crack which would result in significant inleakage.
Change
Description:
As noted above, detailed analyses which provide an estimate of the degree of flow area reduction due to both seismic and LOCA forces are not available for all steam generators. The information that does exist indicates that 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 ficw area reduction for a particular steam generator design, based on the results of a different design, due to the j differences in the design and materials used for the tube support plates.
While a specific flow area reduction has not been determined for some earlier design steam generators, the risk associated 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 five percent of the tubes are susceptible to deformation.
The effect of potential steam generator area reduction on the cold leg break LOCA peak cladding temperature has been either analyzed or estimated for each Westinghouse-plant.- A value of 5 percent area reduction has been-applied for
-VCSNS. The effect of tube deformation and/or collapse will be taken into account by allocating _an appropriate PCT margin.
- Affected Evaluation Model:
L 1981 Large Break ECCS Evaluation Model with BASH l - Status:
Complete. A PCT margin allocation of 11*F has been applied to VCSNS to cover
__up to 5% deformation of steam generator tubes.
Page 15 of 22 at g y- +
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3.4 BROKEN LOOP SAFETY INJECTION (SI) ft0W IN SMALL BREAKS Back a rou n_d :
In the Westinghouse NOTRUMP small break Evaluation Model, it is assumed that the safety injection water which flows to the loop in which the break is postuiated to occur is entirely discharged tu the containment. The ptactice of not taking credit for safety injection into the broken loop preceded the development of calculational models used to satisfy the requirements of 10CFR50.46 or the older Interim Acceptance Criteria (IAC).
It was assumed that neglecting safety injection flow to the broken loop would reduce the capability for core cooling because the flow would not contribute to the reactor coolant system inventory. It was also assumed ihat the interactive ef fects on the break flow and the condensation of s eam would overall result in better core cooling. The besis for these assumptions was questioned.
The model for the calculation of the amount of steam condensation as a result of interaction with the safety injection water was selected for conservatism in the Westinghouse 1985 small break LOCA Evaluation Model. This model was reviewed in detail by the NRC and approved. However, this model is not applicable to the situation of SI injection into the broken loop. To evaluate the ef fect of safety injection flow into the broken loop, a change was made to the ECCS Evaluation Model to provide a more realistic model for the interaction of the safety injection flow with steam in the RCS, A more realistic model was developed based upon test data obtained from the COSI test facility. The COSI test facility is a 1/100 scale representation of the cold leg and SI injection ports in a Westinghouse Pressurized Water Reactor.
The more realistic model was incorporated into a modified version of the Evaluation Model and analysis calculations were performed for a typical three-loop plant.
Analysis calculations which included safety injection flow into the broken lonp with the mora realistic steam-SI condensation model showed a 54*F benefit over the current model analysis calculation, in which SI into the broken loop is not modeled. However, an increase in PCT was noted when SI was modeled in the broken loop for the realistic model, when compared to a realistic model case without SI injection (see summary of results below).
Although incorporation of safety injection flow into the broken loop shows a penalty on the peak cladding temperature calculation, it is Westinghouse's ment that the penalty results from the required mudels of Appendix K to L ,,fR50 regarding break flow. It is Westinghouse judgement that the best estimate response of the system to a small break LOCA event would demonstrate that safety injection flow into the loop containing the break would mitigate the consequences of the event to a greater extent than if safety injection flow were not directed into the broken loop.
Westinghouse concluded that the practice of neglecting safety injection flow into the broken loop in combination with a conservative condensatica model as in the current version of the Westinghouse 1985 small break LOCA Evaluation Model is conservative and in compliance with the regulatory requirements.
Therefore a model change is unnecessary. In order to reach this conclusion.
Page 16 of 22
however, a change had to be made to the Evaluation Model in order to make it applicable to this analysis scenario.
SUMMARY
OF RESULTS PCT *F Current model without safety injection into the broken loop 2037 Revised model with safety injection into the broken Icop 1983 Revised model without safety injection ir' the broken loop 1806 While no change to the Evaluation Model is contemplated as a result of this evaluation, it is possible to view the effect of safety injection flow into the broken loop 9s significant, since the revised condensation model significantly reduces the PCT overall. In accordance with 10CFR50 Appendix K, 11.3: __
" Appropriate sensitivity studies shall be performed for each evdluation model to evaluate the effect on the calculated results of variations in ,
noding phenomena assumed to predominate, including Dump operation or locking, and values of parameters over their applicable ranges. For '
items to which the results are shown to be sensitive, the choices made shall be justified."
The existing model is justified as adequately conservative under the '
requirements of Appendix K to 10CFR50 and will therefore not be revised.
Chanae
Description:
Upon evaluation, it was determined that no change to the ECCS Evaluation Model was necessary.
Affected Evaluation Modei:
1985 Small Break LOCA Evaluation Model Status:
Complete.
Page 17 of 22
4.0 REFERENCES
- 1. " SATAN-VI Program: Comprehensive Space Time Analysis of Loss-of-Coolant" WCAP-8306 (Non-Proprietary), June 1974.
- 2. " Calculational- Model for Core Reflooding after a loss of Coolant Accident (WREFLOOD Code)", WCAP-8171 (Non-Proprietary), June 1974.
- 4. " Containment Pressure Analysis Code (C0CO)", WCAP-8326 (Non-Proprietary),
June 1974.
- 5. "Long Term Ice Condenser Containment Code - Lotic Code", WCAP-8355 (Non-Proprietary), July 1974.
- 6. " Westinghouse ECCS Evaluation Model: 1 981 Version," WCAP-9221-A, Revision 1, (Non-Proprietary).
- 7. "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients", WCAP-9695-A (Non-Proprietary), March 1984.
- 8. "The 1981 Versicn of the Westinghouse ECCS Evaluation Model Using the BASH Code", WCAP-11524-A (Non-Proprietary), March 1987.
.9. Intentionally left blank.
- 10. - Intentionally left blank.
- 11. Intentionally left blank.
~12. "NOTRUMP: A Nodal Transient- Small Break and General Network Code",
WCAP.10080-A (Non-Proprietary).
- 13. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10081-A (Non-Proprietary).
l' 14. Intentionally.left blank.
, 15. "10CFR50.46 Annual Notification for 1989 of Modifications in i Westinghouse ECCS Evaluation Models," NS.4RC-89-3633 Letter from h
W. J. Johnson (Westinghouse) to T. E. Murley (NRC), Dated October 5,1988.
l 16. " Westinghouse Large Breck LOCA Best Estimate Methodology, Volume 1:
l Model Description and Validation, Addendum 4: Model Revisions,"
l WCAP-10924-P, Revision 1, Volume . . Addendum 4, Nissley, M. E. , et. al. .
l August 1990 L
l
- 17. " VANTAGE +- Fuel Assembly' Reference Core Report", WCAP-12610 (Proprietary), June 1991.
- 18. "Large Break LOCA Power Distribution Methodology", WCAP-12935 (Non-Proprietary), May 1991.
Page 18 of 22
TABLE 2 10CFR50.46 MARGIN UTILIZATION FOR LARGE BREAK LOCA
=================================================================
PLANT NAME: VIRGIL C. SUMMER (CGE)
UTILITY NAME: SOUTH CAROLINA ELECTRIC & GAS A. PREVIOUS ANALYSIS OF RECORD PCT = 2141"F Evaluation Model: 81 BASH, FQT= 2.45, FAH= 1.62, SGTP = 15 % Other: Vantage-S Fuel B. PRIOR LOCA MODEL ASSESSMENTS - 1989 APCT= 0*F C. PRIOR LOCA MODEL ASSESSMENTS - 1990 APCT= 0*F D. CURRENT LOCA MODEL ASSESSMENTS - 06/1991
- 1. FUEL ROD INITIAL CONDITION INCONSISTENCY APCT= + 10"F
- 2. ZlRLO MATERIAL PROPERTY CHANGES APCT= 0*F
- 3. LB-LOCA BURST & BLOCKAGE ASSUMPTION APCT= 0"F
- 4. SG TUBE SEISMIC /LOCA ASSUMPTION APCT= + 11*F E. OTHER MARGIN ALLOCATIONS
- 1. ANALYSIS MARGINS USED: Analysis with revised Kfz) APCT= - 27"F V5 Transition core penalty APCT= + 50"F see Note 1.
- 2. PLANT MARGINS USED: APCT= 0*F
- 1. This item was suomitted to NRC as part of the Vantage 5 analyses approved for use at VCSNS in Amendment 75 to the Technical Specifications.
Page 19 of 22
+
TABLE 3 10CFR50.46 MARGIN UTILIZATION FOR LARGE BREAK LOCA
============= ===================================================
PLANT NAME: VIRGIL C. SUMMER (CGE)
UTILITY NAME: SOUTH CAROLINA ELECTRIC & GAS A. CURRENT ANALYSIS OF RECORD PCT = 2083.5*F Evaluation Model: 81 BAS _H, FQT= 2.45, FAH= 1.62, SGTP = 20 %, Other: Vantage + Fuel B. PRIOR LOCA MODEL ASSESSMENTS - 1989 APCT=__ 0*F C. PRIOR LOCA MODEL ASSESSMENTS - 1990 APCT= 0*F D. CURRENT LOCA MODEL ASSESSMENTS - 06/1991
- 1. FUEL ROD INITIAL CONDITION INCONSIS'TNCY APCT= 0*F
- 2. ZIRLO MATERIAL PROPERTY CHANGES APCl= 0*F
- 3. LB-LOCA BURST & 9 LOCKAGE ASSUMPTION APCT= 0*F
- 4. SG TUBE SEISMIC /LOCA ASSUMPTION APCT= + ll*F E. OTHER MARGIN ALLOCATIONS
- 1. ANALYSIS MARGINS USED: V5 Transition core penalty APCT= + 50*F see Note 1
- 2. PLANT MARGINS USED: APCT= 0*F
- 1. This item was submitted to NRC as part of the Vantage 5 analyses approved for use at VCSNS in Amendment 75 to the Technical .
Specifications.
Page 20 of 22
i TABLE 4 10CFR50.46 MARGIN ~ UTILIZATION FOR SMALL BREAK LOCA
...===.................=........................................ ............
PLANTLNAME: VIRGIL C. SUMMER (CGE' g UTILITY.NAME:~50VTH CAROLINA ELECTR: C & GAS A. PREVIOUS ANALYSIS OF RECORD PCT- 2095*F Evaluation Model: NOTRUMP , FQT= 2.45 FAH= 1.68 (See note in E1) SGTP = 15 %, Other: Vantage-5 Tu,el
-B. PRIOR LOCA MODEL ASSESSMENTS - 1989 APCT= 4 37'F C. PRIOR LOCA MODEL ASSESSMENTS - 1990 APCT= 0*F D. CURRENT LOCA MODEL ASSESSMENTS - 06/1991
- 1. FUEL R00 INITIAL CONDITION INCONSISTENCY APCT= + 37'F
- 2. NOTRUMP SOLUTION CONVERGENCE RELIABILITY APCT= 0*F
- 3. SB-LOCA ROD INTERNAL PRESSURE ASSUMPTION APCT= + 26*F
- 4. GAMMA ENERGY DEPOSITION MODEL APCT= 0*F
- 5. ZIRLO MATERIAL-PROPERTY CHANGES APCT= 0*F
- 6. BROKEN-LOOP SAFETY INJECTION FLOW - APCT= 0*F E.- 10CFR50.59 SAFETY EVALUATIONS
- 1. MSSV SETPOINT TOLERANCE RELAXAT!ON 2 Note 1) - APCT= + 61*F ,
(SECL-89-939, FAH constrained to 1.bc)
- 2. EMERGENCY FEEDWATER ENTHALPY DELAY (see Note 2) APCT= + 1*F F. OTHER MARGIN. ALLOCATIONS
- l. ANALYSIS MARGINS USED: Reanal_ysis FAH=I.62. MSSV's APCT= - 120*F 20% SGTP. +20% A.0, (see Note 3)
- 2. PLANT MARGINS USED: APCT= 0*F
- 1. - This item was . submitted to NRC on February 4, 1991 for the Technical Specification Change Request to relax the MSSV setpoint from 11% to 13%.
1This Technical Specification change-has not been approved by NRC.
- 2. This. item was a potential issue evaluated by Westinghouse to account for the_ delay in purging hot water from . the _ Emergency Feedwater line upon initiation of injection.
- 3. The limiting' break was reanalyzed to support ar. increase in steam
-generator _ tube plugging margin from 15_to 20%.-- To gain additional PCT '
_ margin, two additional changes were made. FAH was reduced from 1.68 to the Technical Specification limit of 1.62 L and the core's initial axial of fset - was . reduced f rom +30%- to +20%. - The - overall decrease -in PCT partially offsets _the cumulative effect of the other'ECCS model. changes.
s Page 21 of 22 z-- , . - . - - - . . . =. -
f TABLE 5 10CFR50.46 MARGIN UTILIZATION FOR SMALL BREAK LOCA
====================================_=_==============_,==========
PLANT NAME: VIRGIL , SUMMER (CGE)
UTILITY NAME: SOUTH CAROLINA ELECTRIC & GAS A. CURRENT ANALYSIS OF RECORD PCT = 2007*F Evaluation Model: NOTRUMP , FQT= 2.45, FAH= 1.62, SGTP = 20 %, Other: Vantaae+ Fuel B. PRIOR LOCA MODEL ASSESSMENTS - 1989 APCT= 0*F C. PRIOR LOCA MODEL ASSESSMENTS - 1990 APCT= 0*F _
D. CURRENT LOCA MODEL ASSESSMENTS - 06/1991
- 1. FUEL R0D INITIAL CONDITION INCONSISTENCY APCT= 0*F
- 2. NOTRUMP SOLUTION CONVERGENCE RELIABILITY APCT= 0*F *
- 3. SB-LOCA R0D INTERNAL PRESSURE ASSUMPTION APCT= 0*F
- 4. GAMMA ENERGY DEPOSITION MODEL APCT= 0*F
- 5. ZIRLO MATERIAL PROPERTY CHANGES APCTu '}
- F
- 6. BROKEN LOOP SAFETY INJECTION FLOW APCT= 0*F E. OTHER MARGIN ALLOCATIONS
- 1. ANALYSIS MARGINS USED: APCT= 0*F
- 2. PLANT MARGINS USED: APCT= 0*F
- 3. FUEL MARGINS USED: APCT= 0*F LICENSING BASIS PCT + MARGIN ALLOCATION PCT = 2007*F Page 22 of 22 l
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