ML20077M368
| ML20077M368 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/09/1991 |
| From: | Withers B WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| WM-91-0102, WM-91-102, NUDOCS 9108130312 | |
| Download: ML20077M368 (19) | |
Text
. - _.
I W8LF CREEK
' NUCLEAR OPERATING CORPORATION Bart D. WRhers Procedent and Chief Execuhvo Otheer August 9,1991 WM 91-0102 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Mail Station P1-137 Washington, D. C. 20555
Reference:
1)
Letter SAP-91-169 dated June 20, 1991 from D. S. Lipman, Westinghouse to F. T. Rhodes, WCNOC 2)
_ Letter NS-OPLS-OPL-I-91-380, dated July 11, 1991 from D. S. Lipman, Westinghouse to F. T. Rhodes, WCNOC
Subject:
Docket No. 50-482:
10 CFR 50.46 Annual Report of ECCS Model Revisions Gentlemen This letter provides Emergency Core Cooling System (ECCS) Evaluation Model (EM) revisions as they apply to Wolf Creek Generating Station (WCGS),
and information regarding the effects of ECCS Evaluation Model modifications and application to the WCGS USAR Chapter 15.6.5 LOCA analyses.
The changes in calculated Peak Clad Temperature (PCT) d"e to the revision of Westinghouse ECCS ems are reportable per 10 CFR 50.4t, idelines as follows.
1.
For Large Break LOCA (LB LOCA),
the net FCT increase due to EM revisions is 58.8 F, for a net PCT of 2163.5 F which remains less than the 10 CFR 50.46 limit of 2200 F.
2.
For Small Break. LOCA (SB LOCA),
the net PCT increase due to EM revisions is 37.0 F, for a net PCT of 1917.6 F which is bounded by the large break LOCA PCT.
The information contained in Attachment 1 was provided by Reference 1, and describes the resolution of ECCS EM issues and the impact of ECCS EM changes from August 1990 to May 1991.
Because these issues were under evaluation as of August 1990 they were not reportable under 10 CFR 50.46, which -is consistent with the direction given during a January 1991 meeting between Westinghouse and NRC Staff. contains the calculated LB and SB LOCA PCT margin allocations resulting from the permanent changes to ems, which were also provided by Reference 1.
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kM 91-0102 Page 2 of 2 Reference 2 provided revisions to the SB LOCA PCT changes identified in Reference 1.
Since the 10 CFR 50.46(b)(1) maximum PCT of 2200 F was not exceeded a reanalysis is not required.
If you have any questions concerning this submittal, please contact Mr. H. K. Chernoff of my staff.
Very truly yours, Bart D. Withers President and Chief Executive officer BDW/jd1 Attachments cc L. L. Gundrum (NRC), w/a A. T. Howell (NRC), w/a R. D. Martin (NRC), w/a D. V. Pickett (NRC), w/a
. to VM 91-0102 Page 1 of 14 ATTACHMENT 1 CHANGES TO THE WESTINGHOUSE ECCS EVALUATION MODELS AUGUST 1990 - MAY 1991 L
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. to WM 91-0102 Page 2 of 14
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 Vestinghouse ECCS EM in the period from August 1990 through May 1991, The current Westinghouse ECCS EH applicable to Wolf Creek Generating Station are named as listed in Table 1, and consist of several computer codes with specific functions.
Westinghouse has completed the evaluation of several items related to the Westinghouse ECCS Evaluation Models listed in Table 1.
Each of these items is discussed in the following sections, which include a description of the
- item, the assessment which was performed, the renulting change to the Evaluation Model, and the effect of the change on the PCT.
Some of the subjects discussed represent changes to program coding or to inputs directly related to the physical 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.
These are discussed in Section 3.
Also included, for information, are items for which a technical assessment is continuing, and items for which it was concluded that no change was necessary.
2.0 EVALUATION 110 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 ROD MODEL REVISIONS During the review of the original Westinghouse ECCS Evaluation Modsl following the promulgation of 10CFR50.46 in 1974 Westinghouse committed to maintain consistency between future loss-of-coolant accident (LOCA) fuel rod computer modele 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.
. to VM 91-0102 Page 3 of 14 Channe
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 Evaluation Models to maintain consistency with the latest approved version of the fuel rod design code.
In addition, it was determined that integration of the cladding strain rate equation used in the large break LOCA Evaluation Model, as described in Reference A, was being calculated twice each time step instead of once.
The coding was corrected to properly integrate the strain rate equation.
Effect of Changes:
The changes made to make the LOCA tuel rud models consistant with the fuel design codes were judged to be insignificant, as defined by 10CFR50.46(a)(3)(i).
To quantify the effect on the calculated peak cladding temperature (PCT),
calculations were performed which incorporated the
- changes, including the cladding strain model correction for the large break LOCA.
For the large break LOCA Evaluation Model, additional calculations, incorpo.ating only the cladding strain corrections were performed and the results supported the conclusion that compensating effects were not present.
The PCT effects reported below will bound the effects taken separately for the large break LOCA.
a)
Large Break LOCA The effect of the changes on the large break LOCA peak cladding temperature was determined using the BASH 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:
t ALtachment 1 to VM 91-0102 Page 4 of 14 1.
blowdown Analysis -
It was deter ined that the changes vill 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.
Ilot Assembly Rod Heatup Analysis -
The hot rod heatup calculations would typically show the largest offeet of changes.
Hot rod heatup computer analysis calculations were performed using the LOCBART computer code to assess the affect of the changes on the hot assembly average rod, hot roa and adjacent rod.
3.
Determination of the Effect on the Peak Cladding Temperature The effect of th thanges 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 10 F for the BASH Evaluation Model.
increased by less than by,F would bound the effect a tae peak It was judged that 25 cladding temperature for the BART Evaluation Model, while calculations performed for the Westinghouse 1981 Evaluation Model.showed that the peak cladding temperature could increase by approximately 41 F.
b)
Small Weak LOCA The effect of the changes on the smal'. break LOCA analysis peak cledding temperature calculations was determined using the 1985 small break LOCA Evaluation Model by performing computer analysis cale lations 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 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.
I datus:
Changes completed and implemented, w
.' to WM 91-0102 Page 5 of 14 2.2 SMALL BREAK LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSIMPTION Chance Deecription The Westinghouse small break loss-of-coolant accident (LOCA) emergency core cooling system (ECCS) Evaluation Model analyses assume that higher fuel rod initial fill pressure leads to a higher calculated peak cladding temperature (PCT),
as found in studies with the Westinghouse large break LOCA ECCS Evaluation Model.
H0 wever 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 then the reactor coulant system (RCS) pressure.
A lower fuel rod initial fill pressure could then result in a higher calculated peak cladding temperature.
The Westinghouse small break LOCA cladding strain model is based upon a correlation of Hardy's data, as described in Section 3.5.1 of Reference A.
Evaluation of the limiting fuel rod initial fill pressure assumption revealed that this model was used outside of the applicable range in the small break LOCA Evaluation Model calculations, allowing the cladding to expand and ccntract 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 Chanees Implementation of the corrected cladding creep equation results-in a small reduction in the pellet to cladding gap when the RCS Pressure exceeds the rod internal pressure and increases the gap after RCS pressure falls below the rod internal pressure.
Since the cladding typically demonstrates very little creep toward the fuel pellet prior to core uncovery when the RCS pressure exceeds the rod internal pressure, implementation of the correlation for the appropriate range has a negligible benefit on the peak cladding temperature calculation during this portion of the transient.
- However, after the RCS pressure falls below the rod internal pressure, implementation of an accurate correlation for cladding creep in small break LOCA analyses would reduce the expansion of the cladding away from the fuel compared to what was previously calculated and results in a PCT penalty because the cladding is closer to the fuel.
Calculations were performed to assess the effect of the cladding strain modifications for the limiting three-inch equivalent diameter 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 l
l
. to WM 91-0102 Page 6 of 14 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 e -~sure initial conditions, the combined effect ofthefuelrodinternagpi<atur and the cladding strain model revision is a
typically bounded by 40,F.
.Lwever.
In an extreme case the combined effect could be as large as 60 F.
_ Status Resolution of this issue does not impact WCGS SB LOCA analysis, since rod burst was calculated to occur in the limiting ECCS analysis for VCGS.
Therefore, there is no impact on the WCGS SB LOCA PCT.
3.0 EVALUATION MODEL APPLICATION CllAMCES 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 LO"A POWER DISTRIBUTION ASSUMPTION
Background:
Appendix K to 10CFR50 requires that the power distribution which results in tha 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 Vestinghouse large LOCA Evaluation Model is the chopped cosine power distribution.
This distribution is symmetrical and is defined by two quantities:
the ratio of peak linear power relative to the average (FQT),
and the ratio of hot rod integral power relative to the average (F Delta H).
This power distribution was found to produce the highest peak claddirt temperature (PCT) when compared to power distributions skewed to the top or bottom of the core in studies performed by Westinghouse and submitted to the NRC.
Typically, the power distributions were assumed to peak at discrete elevations in the core (4. 6 8 and 10 feet).
It was also assumed that the key parameters affecting PCT were the FQT.
F Delta H, the peak power location, and integral of power to the peak power elevation.
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. to VH 91-0102 Page '/ of 14 CalculatioAs performed with the advanced LOCA Evaluation Mode 19, BART and BASH, which examined peak power locations and power distributions whlet.a r e not considered in the original analyses, under some circumstances 3 to PCTs greater than those calculated with the cosine distributio
'his behavior was revealed when performing power distribution studies for core designs with relatively low FQT and relatively high F Delta H.
Further studies revealed that, in addition to FQT, F Delta H, and tha peak power location.
the nature of the axial distribution of power affected the results.
That is, two power distributions with the same FQT F Delta H and peak power location, but whose power was distributed differently along the rod could result in significantly different PCTs.
Westinghouse has completed an analysis effort to understand and properly account for the effect of skewed power distributions on the calculated large break LOCA PCT.
This effort 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.
Channe Descrintions-i As a result of these studies, revisions have been made to current reload i
and safety analysis methodology which accounts for the variability in power distributions'from cycle to cycle and plant to plant.
This revision l
provides a means of determining that the current licensing basis (i.e.,
the chopped cosine) is expected to remain. limiting, but also provides-for identifying and analyzing the most severe expected power distribution, if different from-the chopped cosine.
Statust In order to verify that a plant was not affected by this item, a large break LOCA power distribution surveillance factor was applied to confirm that the power shapes identified as potentially being more limiting are not present.
The owners of the affected plants were advised to temporarily apply this surveillance factor to their normal flux map measurements.
In some cases, a temporary 100 F PCT margin allocation was applied, rather than the surveillance factor.
This margin assured that -if limiting power shapes did occur, 10CFR50.46 limits would still be met.
The process described in Reference C will be used to assess specific core-designs.
In this process, each power distrih4 tion calculated in the core design will be evaluated to determine whether it is more limiting than the cosine-power distribution.
Adjustments will be made to the core design operating bands to eliminate these limiting distributions and surveillance factors will be defined to assure that plant safety limits are met.
This will assure that a change to the ECCS Evaluation Model is not required, since the chopped cosine power distribution will remain limiting.
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, to WM 91-0102 Page 8 of 14 3.2 LARGE BREAK LOCA BURST AND BLOCKAGE ASSUMPTION l
Backcround:
The cladding swelling and flow blockage models were reviewed in detail during the NRC's evaluation of the Westinghouse Evaluation Model.
!!owever, the use of the average rod in the hot assembly may not have been documented in a nihnner 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 blocksge resulting from the swelling and rupture of the fuel rods during a loss-of.
coolant accident (LOCA).
10CFR50 Appendix K Paragraph I.B states:
'...To be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not underestimated.'
In Westinghouse ECCS Evaluation Hodel 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 as'embly 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 rod producing the power of the average of rods in the hot assembly and is primarily used to calculate the enthalpy rise in the hot assembly.
Use of the average rod in the model assumes that the time at which blockage is calculated to occur is represented by the burst of the l
average rod. A review of current hot assembly power distributions indicates I
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 repressntation, howecc -
the true onset of blockage would likely begin earlier, as the highest power rods reach their burst temperature.
This time is estis.ated to be a few seconds prior to the time when the average rod bursts.
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. to VM 91-0102 Page 9 of 14 Large break LOCA Evaluation Models which use BART or BASH simulate the hot assembl,' rod with the actual average power, while order Evaluation Models use an Tverage rod power which ie adjusted downward to account for thimbles (this is described in detail in Addendum 3 to Reference 2).
If burst occurs after the flooding rate has fallen below one inch per second, the time at i
which the b\\ockage penalty is calculated will be delayed for these older Evaluation h9dels.
Westinghouse l as performed an evaluation of modeling the blockage based upon the burst of the higher power rods in the hot assembly for the VCGS EH.
The evaluation of the effects of earlier blockage predictJon resulted in a PCT penalty of 33.8 F for the LB LOCA analysis.
Channe 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. Vestinghouse concluded that further artificial changes to the ECCS Evaluation Models to f orce the calculation of an earlier bur",c t ime were not necessary.
In rare instances where burst has not occurred prio* to the flooding rate falling below 1.0-inch /second, the results sf the ECCS analysis calculation are supplemented by a permanent assessment of margin.
Typically this will only occur in cases where the calculated PCT is low.
Westinghouse concludes that no model change is required to calculate an earlier burst time.
Status Complete.
3.3 STEAM GENERATOR FLOW AREA Backnroundt 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) vhich addressed this requirement specifically for steam generator o bes in pressurized water reactors.
In that guide, the staff requireo analytical and experimental evidence that steam generator tube integrity will be maintained for the combinations of the loads resulting from a LOCA with the loads from a safe shutdown earthquake (SSE).
These loads are combined for added conservatism in the calculation of structural integrity.
This analysis provides the basis for establishing criteria for removing from service tubes which had experienced significant degradation.
i l
. to WM 91-0102 Page 10 of 14 Analyses performed by Westinghouse in support of the above requirement for various utilities, combined the most severe LOCA loads with the plant specific SSE, as delineated in the design criteria and the Regulatory Guide.
Generally, these analyses shovad that while tube integrity was maintained, the combined loads led to some tube deformation.
This deformation reduces the flow area through the steam generator.
The reduced flow area increases the resistance through the steam generator to the flow of steam from the core during a LOCA, 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 planto by Westinghouse in the late 1970's and early 1980's.
The combination of LOCA and SSE leads led to the following calculated phenomena 1.
LOCA and SSE loads cause the eteam generat:r 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 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 cora during the reflood phase of the LOCA, increasing the calculated peak cladding t emperature (PCT).
The ability of the steam generator to continue to perform its safety function was established by evaluating the effect of the resulting flow aree 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 te 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 was determined that the patential for flow area reduction due to the contribution of SSE loadings chould be included in other LOCA analyses.
With SSE loadings, flow area reduction may occur in all steam generators (not just the faulted loop).
Therefore, it was concluded that the effects of flow area reduction during the most limiting primary pipe break affecting LOCA PCT, i.e.,
the reactor vessel inlet break (cold leg break LOCA), had to be evaluated to confirm that 10CFR$0.4f> limits continue to be met and that the affected steam generatcts will continue to perform their intended safety function.
. to WM 91-0102 Page 11 of 14 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 vas considered equivalent to an increased level of steam generator tube plugging.
Typically, the area reduction was estimated to range from 0 to 7.51, depending on the magnitude of the seismic loads.
Since detailed non-linear seismic analyses are not available for Series 51 and earlier design steam generators, some area reductions had to be estimated based on available information.
For most of these plants, a 3 percent flow area reduction was assumed to occur in each steam generator as a result of the SSE.
For these evaluations, the contribution of loadings at the tube support plates from the LOCA cold leg break was assumed negligible, since the additional area reduction, if it occurred, would occur only in the broken loop steam generatet.
Vestinghouse recognizes that,lfor most plants, as required by GDC 2
' Design Basis for Protection against Natural Phenomena',
that steam generators must be able to withstand the effects of combined LOCA + SSE loadings end continue to perform their intended safety function.
It is judged that this requirement applies to undegraded as well as locally degraded steam generator tuben.
Compliance with ODC 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 judgement 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 gencrator tubes are to provide a flow path for the venting of steam generated in the core through the RCS pipe break ar.d 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 arca 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 after the combination of LOCA + SSE loads, meeting the requirements of GDC 2.
l
. to WM 91-0102 Tage 12 of 14 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 to primary inleakage during a LOCA event; it was noted that inleakage is not addressed in the existing ECCS analysis.
Westinghouse did not consider the potential for secondary to primary inleakage during resolution of the steam generator tube collapse item.
This is a
relatively new item, not previously addressed, since cracking at the tube support plate elevatione had been insignificant in the early 1980's when the tube collapse item was evaluated in depth.
There is ample data available which demonstrates that undegraded tubes maintain their integrity under collapse loads.
There is also some data which shows that cracked tubes do not behave significantly differently from uncracked tubes when collapse loads are applied.
- However, cracke.1 tube data is available only for round or ulightly ovalised tubes.
It is important to recognize that the core melt frequency resulting from a an significant steam subsequent tube collapse,ontheorderof10'g/RYorless.
ccTbined BOCA + SSE event, This geDerator tube inleakage is very low, estimate takes into account such facters as the possibility of a seismically induced LOCA, the expectcd cccurrence of cracking in a tube as a function of height in the eteam generator tube bundle, the localized effect of the tube eupport plate def orea tion,
and the possibility that a tube which is identified to deform during LOCA + SSE loadings would also contain a partial through vall 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 realons of the tube support plate elevations near vedges that are most susceptible t o collapse which may then lead to penetration of the primary preisure boundary and significant inleakage during a LOCA + SSE event.
Channe 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 flow area reduction for a particular steam generator design, based on the results of a different design, due to the differences in the design and materials used for the tube support plates.
. to WH 91-0102 Page 13 of 14 WCAP-10043, submitted by SLNRC 82-0047 dated 12-3-82 and supplemented by SLNRC 84-0017 dated 2-2-84, documents the Regulatory Guide 1.121 analyses for the SNUPPS plants.
As demonstrated in Section 4.1.2 of WCAP-10043, no tube distortions would occur in the intict steam generators subjected to SSE load only.
A 6.22 flow area reduction in the faulted steam generator was calculated for LOCA plus SSE loadh associated with a double-ended pump suction (DEPS)
- break, with a 02 contribution in flow area reduction from seismic considerations.
Further, generic studies have shown several hundred degrees of PCT margin between the double-ended cold leg (DECL) and DEPS breaks.
Due to this large margin, the DEPS break will remain non-limiting even with the addition of a PCT penalty due to the reduced flow area.
Therefore, no PCT penalty was applied.
Status:
Complete.
4.0 REFERENCES
A.
'LOCTA-IV Program Loss-of-Coolant Transient Analysis",
WCAP-8305, (Non-Proprietary), June 1974.
B.
'BART-Ali A Computer Code for the Best Estimate Analysis of Reflood Transients' WCAP-9695-A (Non-Proprietary), March 1984.
C.
Large Break LOCA Power Distribution Methodology',
WCAP-12935 (Non-Proprietary), May 1991.
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to WH 91-0102 Page 14 of 14 r
l TABLE 1
SUMMARY
OF VESTINGHOUSE ECCS EVALUATION H0DELS NAME:
1981 MODEL VITH BART APPLICATION:
Analysis of Large Break LOCA l
CODES USED:
PURPOSE:
SATAN-V1 Blowdown hydraulic-translent INTERIM-VREFLOOD Reflood hydraulic transient BAkT Hot assembly thermohydraulics INTERIM-LOCTA Fuel rod thermal transient I
COCO or LOTIC Containment pressure transient NOTE:
This model was developed to provide a more realistic calculation of heat transfer during the reflood porcion of the transient.
NAME:
1975 SBLOCA HODEL APP!ICATION:
Analysis of Small Break LOCA CODES USED:
PURPOSE VFLASH System hydraulic transient SELOCA Fuel rod thermal transient NOTE: 'This model is_no longer used, but some plants are licensed under this methodology, t
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, to W 91-0102 Page 1 of 3 ATTACle!ENT 2 l
ECCS EVALUATION HODEL PCT MARG 1H ASSESSMENTS J
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, to VH 91-0102 Page 2 01' 3 Large Break LOCA A.
Analysis of Record PCT = 2100 F B.
1989 LOCA Model Assessments
+0 F C.
1990 LOCA Model Assessments
+0 F l
D.
Current LOCA Model Assessments - June 1991 I
(Permanent Assessment of PCT Margin) 1.
Fuel Rod Model Revisions
+25.0 F (refer to Section 2.1 of Attachment 1) 2.
Burst & Blockage Assumption
+33.8 F (refe. to Section 3.2 of Attachment 1) 3.
SG Flow Area
+0 F (refer to Section 3.3 of Attachment 1)
E.
10CFR50.59 Safety Evaluations
- 1) Containment Purge
+10.5"F
- 2) Loose Parts
+20.2 F F.
Current LOCA Model Issues (Temporary use of PCT Margin)
- 1) LOCA Power Distribution Assumption
+100 F G.
Analysis Margins Used
-126 F Licensing Basic PCT + Margin Allocation PCT = 2163.5 F NOTES:
1.
A safety evaluation was performed at the time of the analysia to assess the effect of a delay in containment isolation due to containment purging at the time of a LOCA accident The penalty associated with 6
this evaluation was determined to be 10.5 F.
2.
A safety evaluation was performed to address the presence of a loose part detected in the lower reactor vessel area during Refuel IV.
3.
Margin between the ECCS analysis value for the core peaking factor of 2.42 and the technical specification limit of 2.32 and margin between the ECCS analysis value for steam generator tube plugging level of 10!
and the plant actual steam generatur tube plugging level of less than 1Z.
- to k'M 91-0102 Page 3 of 3 Small Break LOCA j
A.
Analysis of Record PCT = 3 790"F B.-
1989 LOCA Model Assessments
+0 F C.
1990 LOCA Model Assessments
+0 F 1
D.
Current LOCA Model Assessments (Permanents Assessment of PCT Margin)
- 1) Fuel Rod Model Revisions
+37 F 4
(refer to Section 2.1 of Attachment 1)
- 2) Rod Internal Pressure Assumption 40 F (refer to Section 2.2 of Attachment 1)
E.
10CFR50.59 Safety Evaluations
- 1) V5H Zirc Grids 42 F
- 2) AW Enthalpy I'elay
+44 F
- 3) Loose Parts
+44.6 F Licensitig Basis PCT + Margin Allocation PCT = 1917.6 F NOTES:
1.
A safety evaluation was performed to account for the effect of zircaloy gzids present with the Cycle 6 Vantage 5H Fuel Reload.
2.
It was determined that t.he appropriate Auxiliary Feedwater (AW) purge volume had not been used in calculating the delay time for A W enthalpy switchover.
Based upon 6ystem geometry and assumed AW flowtate, an enthalpy switchover time of 213 seconds would have been more appropriate than the 25 seconds assumed in the original analysis.
. During the
.sdditional swit. hover delay, higher enthalpy feedwater in the connon portions of the feed lines is purged by the incoming AW and delivered to the steam generators.
This higher enthlapy feedwater acts to reduce the heat transfer from the primary to secondary and results iD A higher PCT.
The PCT increase for SB LOCA roulting from the AW pu ge volume correction was conservatively calculated to be 44 i
3.
A safety evaluation was performed to address the pcesence of a loose part detected in the lower reactor vessel aran during Refuel IV.
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