ELV-02368, Discusses 10CFR50.46 ECCS Evaluation Model Significant Errors/Changes Rept & Provides Info Re Current Analysis Peak Cladding Temp Results & Provides Plant Safety Evaluations
| ML20082E092 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/26/1991 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ELV-02368, ELV-2368, NUDOCS 9107310202 | |
| Download: ML20082E092 (25) | |
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% ear Oy:rscos ELV-02368 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATlN: Document Control Desk Washington, D. C. 20555 Gentlemen:
Vogtle Electric Generating Plant 10 CFR 50.46 ECCS MODEL SIGNIFICANT ERRORS / CHANGES REPORT On June 28, 1991, Georgia Power Company became aware of significant errors and changes in the Westinghouse Emergency Core Cooling System (ECCS)
Evaluation Hodel for large and small break loss of coolant eccidents (LOCAs) permanently assessed against the current Vogtle Electric Generating Plant ECCS Models.
Enclosed is Georgia Power Company's report, in compliance with the 10 CFR 50.46 significant errors / changes reporting requirements, for Vogtle Electric Generating Plant Units 1 and 2.
For completeness, the effects of the significant errors / changes on the VANTAGE-5 ECCS Model for large and small break LOCAs currently under review by the NRC are included in this report.
Attachment A provides information regarding the effect of the ECCS Evaluation Model significant errors / changes on the current analysis peak cladding temperature (PCT) results reporte(i in section 15.6 of the Vogtle Electric Generating Plant Units 1 and 2 Final Safety Analysis Report (FSAR).
Attachment B provides a summary of the plant change safety evaluations performed under the-provisions of 10 CFR 50.59 that also affect the PCT results. Please note that the facili_ty change safety cvaluations included in Attachment B reflect only _those which result in nonzero PCT impact assessments.
It should also be noted that plant PCT margin between the FQ peaking factor assumption in the large break LOCA analysis (F(-2.32) and the current Vogtle Technical Specifications (Fg-2.30) was usec to demonstrate compliance with the 10 CFR 50.46 acceptance criteria on PCT (limit of 22000F).
Attachment C provides information regarding the effects of the VANTAGE-5 ECCS Evaluation Model errors / changes on PCT reported in the VANTAGE-5 r
j Technical Specifications submittal package to the NRC (ELV-02166, dated November 29,1990).
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Geory,ialher ah U. S. Nuclear Regulatory Commission ELV-02368 Page 2 Attachment D provides a summary of the plant changes not specifically included as part of the ECCS Evaluation Model results in Attachment C but reviewed in the VANTAGE-5 analysis, it should be noted that only the VANTAGE-5 small break LOCA errors and changes resulted in a PCT sum that is significant.
This information package constitutes Georgia Power Company's report to the NRC detailing the significant errors / changes per the reporting requiremeni.s of 10 CFR 50.46(a)(3)(ii).
For completeness, Attachments C and D ere included in this report for the VANTAGE-5 analysis currently under review by the NRC.
It has been determined that compliance with the requirements of 10 CFR 50.46 continues to be maintained when the effects of plant design changes performed under 10 CFR 50.59, plant margins which could affect the large break LOCA and small break LOCA analyses results, and the effects of the ECCS Evaluation Model significant errors / changes are combined.
If you have any questions regarding this report, please contact this office.
Sincerely, t.{l,yf/Yo
-DL W. G. Hairs +on, III WGH,Ill/ gps Attachment a
cc:
.Georaia Power Company hr. C. K. McCoy Mr. W. B. Shipman Mr. P. D. Rushton Mr. M. Sheibani NORMS U.S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle
ATTACHMENT A EFFECT OF WESTINGHOUSE ECCS EVALUATION MODEL SIGNIFICANT ERRORS / CHANGES ON THE LOCA ANALYSIS RESULTS
'FOUND IN SECTION 15.6 0F THE V0GTLE UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models within 30 days of discovery, when the errors and changes _ are significant.
Reference 1 defines a significant error or change as one which results in a calculated fuel peak cladding temperature (PCT) different by more than 500F from the temperature calculated for the limiting _ transient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 500F.
The following presents-an assessment of the effect of the significant errors and changes to the Westinghouse ECCS Evaluation Models on the loss L
of cool' ant cccident (LOCA) analysis results found in section 15.6 of the Vogtle Units 1 and 2 Final Safety Analysis Report (FSAR).
LARGE BREAX LOCA-ECCS EVALUATION M001L
- The large break LOCA analysis for Vogtle Units 1 and 2 was examined to assess the effect of the significant errors and changes to the Westinghouse large break LOCA ECCS Evaluation Model on PCT results reported in Section 15.6 of the FSAR.
The large break LOCA analysis-results were calculated using the 1981. version of the Westinghouse large break LOCA ECCS Evaluation Model (Reference 2).
The limiting size break analysis assumed the-following information important to the large break LOCA analyses:
-o 17x17 Standard fuel Assembly o'
Core Power - 1.02
- 3411 MWt o
Vessel Average Temperature - 589.60F I
o Steam Generator Plugging Level - 5%
o Fq - 2.32 o
F-delta-H = 1.55 o
i
' ATTACHMENT A Page 2 For Vogtle Units 1 and 2, the limiting size break resulted from the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD - 0.6 for the maximum safeguards condition.
The current analysis-of-record calculated PCT was 1995.80F.
The following errors and changes to the Westinghouse ECCS Evaluation Models would affect the 1981 Evaluation Model (Reference 5) large break LOCA analysis results found in section 15.6 of the Vogtle Units 1 and 2 FSAR, the sum of which are significant:
FUEL R0D MODFL REVISIONS During the review of the original Westingbouse ECCS Evaluation Medel following the promulgation of 10 CFR 50.46 in 1974, Westinghouse committed to maintain consistency between future 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.
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 pe!let surface roughness models.
l 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 l
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 6, was being calculated twice each time step instead of once.
The coding was corrected to properly integrate the strain rate equation.
The changes made to make the LOCA fuel rod models consistent with the fuel design codes were judged to be insignificant, as defined by 10 CFR 50.46 (a)(3)(i).
To quantify the effect on the calculated PCT, calculations were performed which incorporated the changes, including the cladding strain model correction for the large break LOCA.
For the large break LOCA Evaluation Model, additional calculations, incorporating only the cladding strain corrections, were performed, and the results supported l
i
' ATTACHMENT A Page 3 the conclusion that compensating effects were not present.
The PCT effects reported below will bound the effects taken separately for the large break LOCA.
The effect of the changes on the large break LOCA PCT 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:
1.
Blowdown Analysis It was determined that the changes will have a small effect on the core averege 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 rod heatup calculations would typically show the largest effect of the changes. Hot rod heatup computer analysis calculations were performed using the LOCBART computer code to assess the effect of the changes on the hot assembly average rod, hot rod, and adjacent rod.
3.
Determination of the Effect on the PCT The effect of the changes on the calculated PCT was determined by performing calculations using the 1981 Evaluation Model.
The analysis calculations confirmed that the effect of the ECCS Evaluation Model changes were insignificant as defined by 10 CFR 50.46(a)(3)(i).
The calculations showed that the PCT increased by less than 410F.
Therefore, a 410F-penalty has been assessed against the Vogtle large break LOCA results.
FUEL R00 BURST AND BLOCKAGE APPLICATION 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 detailea enough to allow the staff to adequately assess this aspect of the model.
' ATTACHMENT A Page 4 Appendix K to 10 CFR 50 requires consideration of the effects of flow blockage resulting from the swelling and rupture of the fuel rods during a LOCA.
Paragraph 1.8 of 10 CFR 50 Appendix K 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 inay 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 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 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 bursting of this rod adequately represents when most of the rods will l
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 (described in detail in Addendum 3 to Reference 7).
If burst occurs after tha flooding rate has fallen below 1.0 inch /second, the time at which the l
l blockage penalty is calculated will be delayed for these older Evaluation Models.
Ample experimental evidence currently exists which shows that flow blc,cxage does not result in a heat transfer penalty during a LOCA.
In addition, newer evaluation models have been oeveloped and licensed which demonstrate that the older evaluation models contain a substantial amount of l
i
' ATTACHMENT A Page 5 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 occurred 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, 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.
As part of the evaluation of the large break LOCA burst and blockage assumptions, the Plant Vogtle ECCS analyses were reviewed. The review of the analyses indicated that the hot assembly average fuel rod did not rupture until after the flooding rate was below 1.0 inch /second.
Since an interpretation of the requirements of Appendix K to 10 CFR 50 could require that the effects of flow blockage be represented as soon as the flooding rate fell below 1.0 inch /second, an estimate of the potential effect on the hot rod calculated PCT for Plant Vogtle was performed.
This resulted in a penalty assessment of 1650F to the PCT results, which is significant.
STEAM GENERATOR FLOW AREA APPLICATION 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 (GDC) 14 and 31]. The NRC issued Regulatory Guide 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 mairtained for the combination 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.
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 showed that while tube integrity was maintained, the combined loads led to some tube deformation.
This deformation reduces the flow area in the steam generator. The reduced flow area increases the resistance in the steam generator to the flow of steam i
from the core during a LOCA, which potentially could increase the i
calculated PCT.
The offect of tube deformation and flow area reduction in the steam generaw 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:
' ATTACHMENT A Pag 3 6 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 postulated large break LOCA, the primary side depressurizes to containment pressure. Applying the resulting pressure differential to the deformed tubes causcs some of the these tubes to collapse and reduces the effective flow area in 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 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.
In April 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 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 10 CFR 50.46 limits continue to be met, and that the affected steam generators will continue to perform their intended safety function.
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 percent, depending on the magnitude of the seismic loads. Since detailed nonlinear 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 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 negligible, since the additional area reduction, if it occurred, would occur only in the broken loop steam generator.
l
' ATTACHMENT A
)
Page 7 Westinghouse recognizes that for most plants, (as required by GDC 2,
" Design Basis for Protection Against Natural Phenomena") the steam generators must be able 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 Westing!,ouse'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 et conflict with the requirements of GDC 2.
During a large break LOCA, the intended safety functions of the steam geneiator tubes are to provide a flow path for the venting of steam generated in the core through the RCS pipo 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 stoam generator tube deformation, per se, with loss of safety function.
Crosssectional 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.
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 elevations had been insignificant in the early 1980s when the tube collapse item was evaluated in depth. There is ample data available which demonstrates that undegraded tubes maintain their integrity under collapse
' ATTACHMENT A Page 8 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, cracked tube data is available only for round or slightly ovalized tubes.
It is important to recognize that the core melt frequency resulting from a combined LOCA + SSE event, subsequent tube collapse, and significant steam generator tube inleakage is very low, on the order of 10-8/RY or less.
This 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 possibility that a tube identified to deform during LOCA + SSE 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.
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.
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 5 percent of the tubes are susceptible to deformation.
The effect of potential steam generator flow area reduction on the cold leg large break LOCA PCT has been estimated for Vogtle to be a 100F-penalty on PCT.
PLANT MARGIN UTILIZATION Based upon the cumulation of the calculated PCT with the permanent and temporary assessments of PCT margin, the Plant Vogtle estimated ECCS analysis results for the 1981 ECCS Evaluation Model case would exceed the 22000F-limit of 10 CFR 50.46.
1
l
' ATTACHMENT A Page 9 However, plant margin is available to reduce the estimated PCT. The difference between the ECCS analysis assumption for the total core peaking factor (FQT) of 2.32 and the current Vogtle Units 1 and 2 Technical Specifications limit for FQT of 2.30 found in the Core Operating Limits Report was credited.
For the 1981 ECCS Evaluation Model calculation for Plant Vogtle, the 0.02 difference in FQT is equivalent to approximately 470F.
Therefore, a 470F PCT benefit is credited to the Vogtle large break LOCA results.
RESULTANT LARGE BREAK LOCA PCT As discussed above, significant errors and changes to the Westinghouse large break LOCA ECCS Evaluation Model and credit for available Vogtle plant PCT margin will result in the following PCT results:
1.
Current Analysis-of-Record 1995.80F 2.
Prior LOCA Model Assessments-1989
+ 16.00F (Reference 3) 3.
Current LOCA Model Assessments-June 1991 a) Fuel Rod Model Revisions
+ 41.00F b)
Fuel Rod Burst and Blockage Application
+ 165,00F c) Steam Generator Flow Area Application
+ 10.00F d)
Plant Margin on FQT
- 47.00F 2180.80F 4.
ECCS Model Errors / Changes Resultant PCT CONCLUSION An evaluation of the effect of significant errors and changes to the Westinghouse large break 1981 ECCS Evaluation Model was performed for the large break LOCA analysis results found in sectior 15.6 of the Vogtle Units 1 and 2 FSAR. When the effects of the large breat ECCS model errors / changes were combined with the current plant analysis results and use of available Vagtle plant PCT margin, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained.
SMALL BREAK LOCA ECCS EVALUATION MODEL The small break LOCA analysis for Vogtle Units 1 and 2 was also examined to assess the effect of the errors / changes to the ' Westinghouse ECCS Evaluation Models on PCT results (the sum of which are significant) reported in section 15.6 of the FSAR. The small break LOCA analysi.; results were calculated using the October 1975 version of the Westinghouse small break LOCA ECCS Evaluation Model incorporating the WFLASH computer code (Reference 8). The analysis assumed the following information important to the small break LOCA analyses:
i I
l
' ATTACHMENT A Page 10 o
17x17 Standard Fuel Assembly o
Core Power - 1.02
- 3411 MWt o
Vessel Average Temperature - 589.60F o
Steam Generator Plugging Level - 5%
Fg - 2.20 at 10 ft o
o F-delta-H - 1.55 For Vogtle Units 1 and 2, the limiting size small break resulted from a 4-inch equivalent diameter break in the cold leg. The current analysis-of-record calculated PCT was 15370F.
The following errors and changes te the Westinghouse ECCS Evaluation Models would affect the WFLASH small break LOCA analysis results found in section 15.6 of the Vogtle Units 1 and 2 FSAR, the sum of which are significant:
FUEL R00 MODEL REVISIONS As discussed earlier in the fuel rod model revisions for large break LOCA section, there are fuei rod model errors and changes in the ECCS Model.
The effect of these changes on the wall break LOCA analysis PCT 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 was insignificant as defined by 10 CFR 50.46(a)(3)(i).
The calculations showed that 370F would bound the effect on the calculated PCT for four-loop plants. Therefore, a 370F PCT penalty has been assessed against the Vogtle small break LOCA results.
SMALL BREAK LOCA R00 INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION The Westinghouse small break LOCA ECCS Evaluation Model analyses assume that higher fuel rod initial fill pressure leads to a higher calculated PCT, as found in studies with the Westinghouse large break LOCA ECCS Evaluation Model. However, lower fuel rod internal pressure could result in decreased cladding creep (rod swelling) away from the fuel pellets when the fuel rod internal pressure was higher than the reactor coolant system (RCS) pressure. A loaer fuel rod initial fill pressure could then result in a higher calculated PCT.
8
' ATTACHMENT A i
Page 11 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 6.
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 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.
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 PCT calculation during this portion of the transient. However, after the RCS pressure f alls 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 calculsted 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 PCT resulting from the cladding strain model change would be less than 200F, The effect on the calculated PCT depended upon when the PCT occurs and whether the rod internal pressure was above or below the system pressure when the PCT 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 400F.
Therefore, a 400F-penalty has been assessed against the Vogtle small break LOCA results.
NOTRUMP CODE SOLUTION CONVERGENCE This error and change to the NOTRUMP code solution convergence reliability does not affect the Vogtle WFLASH code small break LOCA results.
- However, the error and change affects the VANTAGE-5 NOTRUMP results, which is discussed in Attachment C.
Therefore, this issue is not applicable to the Vogtle WFLASH ECCS Model.
' ATTACHMENT A l
Page 12 RESULTANT SMALL BREAK LOCA PCT As discussed above, errors and changes to the Westinghouse small break LOCA ECCS Evaluation Model will result in the following PCT results, the sum of which is significant:
1.
Current Analysis-of-Record 1537,00F 2.
Prior LOCA Model Assessments-1990
+
0.00F (Reference 4) 3.
Current LOCA Model Assessments-June 1991 a) Fuel Rod Model Revisions
+ 37.00F b) Rod Internal Pressure Assumption
+ 40.00F 4.
-1614.00F CONCLUSION An evaluation of the effect of errors and changes to the Westinghouse small break WFLASH, ECCS Evaluation Model was performed for the small break LOCA analysis results found in section 15.6 of the Vogtle Units 1 and 2 FSAR.
When the effects of the small bretk ECCS model errors / changes (the sum of which is significant) were combined with the current plant analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained.
REFERENCES 1.
" Emergency Core Cooling Systems; Revisions to Acceptance Criteria,"
Federal Register, Vol. 53, No.180, pp. 35996-36005, dated September 16, 1988.
2.
WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 (Non-Proprietary), " Westinghouse ECCS Evaluation Model - 1981 Version,'
1981, Eicheldinger. C.
3.
ELV-01184, "Vogtle Electric Generating Plant,10 CFR 50.46 Annual ECCS t
Model Changes Report," letter from W. G. Hairston (GPC) to USNRC, dated December 22, 1989.
4.
ELV-02368, "Vogtle Electric Generating Plant,10 CFR 50.46 Annual ECCS Model Changes Report," letter from W. G. Hairston (GPC) to USNRC, dated December 20, 1990.
5.
" Westinghouse ECCS Evaluation Model: 1981 Version," WCAP-9221-A, Revision 1, (Non-Proprietary).
6.
"LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, (Non-Proprietary), June 1974.
I
' ATTACHMENT A Page 13 7.
"BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9695-A (Non-Proprietary), March 1984.
8.
" Westinghouse Emergency Core Cooling System Small Break Octobe-1975 Model," WCAP-8971-A (Non-Proprietary).
vJ,
ATTACHMENI B EFFECT OF SAFETY EVALUATIONS PERFORMED ON THE LOCA ANALYSIS RESULTS FOUND IN SECTION 15.6 0F THE V0GTLE UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT j
LARGE BREAK LOCA DESCRIPTION OF PLANT MODIFICATIONS' The large break loss of coolant. accident (LOCA) analysis results have been
. supplemented by safety evaluations of changes which could affect the peak cladding temperature (PCT) as follows:
1.
A safety evaluation to determine the effect for a change of the charging flow rates used in the FSAR section 15.6 large break LOCA analysis due to-increased runout flow of the charging pumps was
- performed for Vogtle Units 1 and 2.
This evaluation determined that the large break LOCA analysis PCT results could be affected by a 20F-increase.
2.
A safety evaluation to determine the effect of a change in safety injection. flow was performed for the Vogtle Units 1 and 2 FSAR sectionL 15.6 large break LOCA analysis.
This evaluation determined that the 'large break LOCA analysis PCT results could be affected by a 30F-increase.
3.
A safety evaluation to determine the effect of containment purging during a LOCA was performed for the Vogtle Units 1 and 2 FSAR section 15.6 large break LOCA analysis. This evaluation determined that the large break LOCA analysis PCT results could be affected by a 1
BESULTANT LARGE BREAK LOCA PCT As discussed above, the plant modifications could affect the resultant FL.
as follows:
Resultant PCT from ECCS Evaluation Model Significant Errors / Changes Reported in Attachment A 2180.80F
- 1. -Safety Evaluation for Charging Pump Increased Runout
+
2.00F 2.
Safety Evaluation for Safety Injection Flow Changes
+
3.00F 3.
Safety Evaluation for Containment Purging
+ 10.00F Licensing Basis PCT-
=2195.80F C.0NCLUSIONS It was determined that compliance with the requirements of 10 CFR 50.46
-would be maintained when safety evaluations for changes which affected the J
.~..
r 2.
I.:
" ATTACHMENT B
,Page 2.-
Y i
F large break LOCA analysis _results_were combined with the effect of the
-large break ECCS Evaluation Model. significant errors and changes -(reported 4
-in Attachment -A) applicable to Vogtle Units 1_and 2.
r SMALL BREAK LOCA E
' DESCRIPTION OF PLANT MODIFICATIONS The small. break LOCA analysis results have been_ supplemented by a safety evaluation which could affect the PCT as follows:
1.
A safety evaluation to determine the effect of changing instrumentation uncertainties due to Veritrak transmitters was performed for the Vogtle Units 1 and 2 FSAR-section 15.6 small break LOCA analysis.
This
[~
evaluation determined that the small break LOCA analysis PCT results
. could = be affected by a _3.70F-increase.
!~
- 2. -A safety evaluation was performed for the Vogtle Units 1 and 2 FSAR
.section 15.6 small break LOCA analysis to determine-the effect of relocating each: steam generator instrumentation line lower level' tap and associated changes in' each steam generator initial nominal water
' level
_This evaluation determined that the small break LOCA analysis
- PCT results could be affected by an ll.00F-increase.
. RESULTANT SMALL BREAK LOCA PCT As discussed above, _the plant modifications could affect the resultant PCT
. as follows:
i:
Resultant PCT from ECCS Evaluation Model Significant Errors / Changes Reported in Attachment A 1614.00F
- 1. ' Safety Evaluation for Veritrak Transmitters
+
3.70F-I'_
2.
Safety Evaluation for Steam Generator Lower Level Tap Relocation
+- 11.00F Total: Resultant _ PCT 1628.70E
' CONCLUSIONS It was determined that compliance with the requirements of 10 CFR 50.46 would be maintained when safety evaluations for changes which affected -the small break LOCA analysis results were combined with the effect of the -
small bre1k ECCS Evaluation Model. errors and changes (the. sum of which are L
-significant) applicable to Vogtle Units 1 and 2.
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e ATTACHMENT C EFFECT OF WESTINGHOUSE ECCS EVALUATION MODEL SIGNIFICANT ERRORS / CHANGES ON THE LOCA ANALYSIS RESULTS FOUND IN THE V0GTLE UNITS 1 AND 2 VANTAGE-5 FUEL DESIGN LICENSING AMENDMENT REQUEST BACKGROUND Provisions in 10 CFR 50.46 required applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models within 30 days of discovery, when the errors and changes are significant.
Reference 1 defines a significant error or change as one which results in a calculated fuel peak cladding temperature (PCT) different by more than 500F from the temperature calculated for the limiting transient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 500F.
The following presents an assess:.ent of the effect of the significant errors and changes to the Westinghouse ECCS Evaluation Models on the LOCA analysis results fcund in the Vogtle Units 1 and 2 VANTAGE-5 Fuel Design Licensing Amendment Request (Reference 2).
LARGE BREAK LOCA ECCS EVALUATION MODEL The large break LOCA analysis for Vogtle Units 1 and 2 was examined to assess the effect of errors and changes to the Westinghouse large break LOCA ECCS Evaluation Model on PCT results reported in Reference 2.
The large break LOCA analysis results were calculated using the Westinghouse BASH large break LOCA ECCS Evaluation Model (Reference 3). The limiting size break analysis assumed the following information important to the large break LOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly o
Core Power - 1.02
- 3565 MWt o
Vessel Average Temperature - 587.30F o
Steam Generator Plugging level - 10%
Fg - 2.50 o
o F-delta-H - 1.65 l
ATTACHMENT C Page 2 For Vogtle Units 1 and 2, the limiting size break resulted from the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD - 0.6.
The VANTAGE-5 analysis calculated PCT was 20370F, The following errors and changes to the Westinghouse ECCS Evaluation Models J
would affect the BASH Evaluation Model large break LOCA analysis results:
FVEL R0D MODEL REVISIONS As discussed earlier in Attachment A, there are fuel rod model errors and changes in the ECCS Model.
The effect of the changes / errors on the calculated PCT was determined by performing calculations using the BASH Evaluation Model.
The analysis calculations confirmed that the effect of the ECCS Evaluation Model changes was insignificant as defined by 10 CFR 50.46(a)(3)(i).
The calculations i
showed that the PCT increased by less than 100F, Therefore, a 100F PCT l
penalty has been assessed against the Vogtle VANTAGE-5 large break LOCA l
results.
FUEL R0D BURST AND BLOCKAGE APPLICATION As discussed earlier in Attachment A, there ar" eno:
and changes to the large break LOCA fuel rod burst and blockage applicat.on.
In the Vogtle VANTAGE-5 BASH analysis, the hot assembly average fuel rod was calculated to burst prior to the time at which the flooding rate was calculated to fall below 1.0 inch /second.
Since the effect of flow blockage was not underestimated in the steam cooling period, the fuel rod burst and blockage application concern is not applicable to the BASH large break LOCA l
VANTAGE-5 analysis.
STEAM 6t'NERATOR FLOW AREA APPLICATION As discussed earlier in Attachment A, there is a steam generator flow area application error / change to the large break LOCA ECCS Evaluation Model.
The effect to the VANTAGE-5 large beak LOCA results are the same as for the 1981 Evaluation Model results.
Thus, a 100F PCT penalty has been assessed to the VANTAGE-5 large break LOCA PCT results.
RESVLTANT LARGE BREAK LOCA PCT As discussed above, errors and changes to the Westinghouse large break LOCA ECCS Evaluation Model will result in the following PCT results for VANTAGE-5:
1.
VANTAGE-5 Analysis Results 2037.00F 2.
Current LOCA Model Assessments-June 1991 a)
Fuel Rod Model Revisions
+- 10.00F b) Steam Generator Flow Area Application
+ 10 QOF 2
3.
ECCS Model Errors / Changes Resultant PCT 2057.00F
' ATTACHMENT C Page 3 CONCLUSID!(
An evaluation of the effect of errors and changes to the Westinghouse large break BASH ECCS Evsluation Model was performed for the large break LOCA analysis results found in the Vogtle Units 1 and 2 VANTAGE-5 Fuel Design Licensing Submittal (Reference 2). When the effects of the large break ECCS model errors / changes were combined with the VAN 1 AGE-5 plant analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained.
SMAll. BREAK LOCA ECCS EVALUATION MODEL The small break LOCA analysis for Vogtle Units 1 and 2 was also examined to assess the effect of the errors / changes to the Westinghouse ECCS Evaluation Models on PCT results (the sum of which are significant) reported in the VANTAGE-5 Licensing Submittal (Reference 2). The small break LOCA analysis results were calculated using the NOTRUMP version of the Westinghouse small break LOCA ECCS Evaluation Model (Reference 4). The analysis assumed the following information important to the small break LOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly o
Core Power - 1.02
- 3565 MWt o
Vessel Average Temperature - E87.30F o
Steam Generator Flugging Level - 10%
o FQ - 2.47 at 9.5 ft o
F-delta-H - 1.70 For Vogtle Units 1 and 2, the limiting size small break resulted from a 3-inch equivalent diameter break in the cold leg. The VANTAGE-5 analysis calculated PCT was 20370F.
The following errors ah! changes to the Westinghouse ECCS Fvaluation Models would affect the NOTRUMP small break LOCA analysis results found in Reference 2, the sum of which are significant:
1
,1
' ATTACHMENT C Page 4 f.UEL R00 MODEL REVIS10NJ As discussed earlier in Attachment A, there are fuel rod model errors and changes in the ECCS Model.
The effect of these changes on the small break LOCA analysis PCT calculations was determined using the 1985 small brehk LOCA Evaluation Model (Reference 4) 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 was insignificant as defined by 10 CFR 50.46(a)(3)(i). The calculations showed that 370F would bound the effect on the calculated PCT for four-loop plants. Therefore, a 370F PCT penalty has been assessed against the Vogtle small break LOCA VANTAGE-5 results.
SMALL BREAK LOCA ROD INTERNAL PRESSURE INITIAL CONDITION ASSUMPTION As discussed earlier in Attachment A, there are errors / changes in the small break LOCA rod internal pressure initial condition assumptions in the ECCS Model. The effect of these changes on the VANTAGE-5 NOTRUMP small break LOCA results 15 a 400F-penalty on PCT.
NOTRUMP CODE SOLUTION CONVERGENCE 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 by Appendix K to 10 CFR 50.
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 pressura, mass, and internal energy.
Standard input values were specified for the input parameters which govern the time step size selection.
However, since tha initial studies, modifications were made to the NOTRUMP computer program to enhance code performance and a
j implement necessary modifications. Subsequent to the modifict.tions, solution convergence was not reconfirmed.
/
To analyze chang
'n plant operating conditions, sensitivity studies were performed with the NOTRUMP computer code for variations in initial RCS pressure, auxiliary feedwater flow ntes, power distribution, etc., which
')
resulted in PCT variations greater than anticipated based upon engineering x
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 size selection criteria which culminated in a revision to the recommended time step size selection criteria inputs.
Fixed input values originally recommended for the steady-state and all bred transient calculations were modified to l
'And o-J ATTACHMENT C Page 5 assure converged results. The NOTRUMP code was reverified against the
'SVT-08 Semiscale experiment, and it was confirmed that the code adequately predicts key small break phenomena.
-Generally, the modifications result in small shifts in timing of core uncosery anel recovery. However, these changes may result in a change in the calculated PCT which exceeds 500F for some plants.
Based on representative calculations,.however, this change will most likely result in a reduction in the calculated PCT.
Since the potential beneficial effect of a nonconverged solution is 31 ant specific, a generic PCT effect cannot be provided.. However, it has )een concluded that current licensing basis results remain valid since the results are conservative relative to the change.
The VANTAGE-5 NOTRUMP small break LOCA analysis for Vogtle incorporated the improvement to the NOTRUMP solution convergence criteria, which is a change to the solution convergence criteria of the last NRC-reviewed and approved i
generic ECCS Evaluation Model. Therefore, the Vogtle VANTAGE-5 small. break LOCA analysis may be assumed to be adequately converged. The potential benefit' to Vogtle's small bre.1k LOCA analysis is then already included in
-the_ VANTAGE-5 PCT analysis results and is not affected by this change / error.--
RESULTANT SMALL B.REAK LOCA PCT As discussed above, errors and changes to the Wes+1nghouse small break LOCA ECCS Evaluation Model (the sum of-which are significant) will result in the following PCT results for VANTAGE-5:
1.
VANTAGE-5 Analysis 2037.00F l
2.
Current LOCA Model Assessments-June 1991 L
a) Fuel Rod Model Revisions
+ 37.00F b) Rod Internal Pressure Assumption
+ 40.00F 3.
-2114.00F CONCLUSION An-evaluation of the effect of errors and changes to the Westinghouse small break NOTRUMP ECCS Evaluation Model was perro,med for the small break LOCA
. analysis results found in the Vogtle Units 1 and 2 VANTAGE-5 Fuel Design E
Licensing Submittal (Reference 2).
When the effects of the small break-ECCS model errors /changs; (the sum of which are significant) were combined with-the VANTAGE-5 plant analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained.
J o
ATTACHMENT C Page 6 REFERENCES 1.
" Emergency Core Cooling Systems; Revisions to Acceptance Criteria,"
federal Register, Vol. 53, No.160, pp. 35996-36005, dated Septer..ber 16, 1988.
2.
ELV-02166, "Vogtle Elect:
Generating Plant, Request for Technical Specifications Changes, VANTAGE-5 Fuel Design," letter from W. G. Hairston (GPC) to USNRC, dated November 29, 1990.
3.
"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-il524-A (Non-Proprietary), March 1987.
4.
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10081-A (Non-Proprietary).
J ATTACHMENT D EFFECT OF SAFETY EVALUATIONS PERFORMED ON THE LOCA ANALYSIS RESULTS FOUND IN THE V0GTLE UNITS 1 AND 2 VANTAGE-5 FUEL DESIGN LICENSING AMENDMENT REQUEST I
I l
LARGE BREAK LOCA 1
' DESCRIPTION OF PLANT MODIFICATIONS The VANTAGE-51arge break loss of coolant accident (LOCA) analysis results have been supplemented by safety evaluations of changes which could affect the PCT a.; follows:
1.
A safety evaluation to determine the effect of containment purging during a LOCA was included with the Vogtle Units 1 and 2 VANTAGL15 large break LOCA analysis. This evaluation determined that the large break LOCA analysis PCT results could be affected by a 100F-increase.
2.
A safety evaluation to determine the effect of a +/- 60F-uncertainty band for the RCS operating average temperature was included with the VANTAGE-5 analysis. This evaluation determined that the large break LOCA analysis PCT results could be affected by a 110F-increase.
3.
A transition cycle (LOPAR to VANTAGE-5 fuel) penalty assessment of 500F in PCT.
RESULTANT LARGE BREAK LOCA PCT l
As discussed above, the plant modifications could affect the resultant VANTAGE-5 PCT as follows:
Resultant PCT from ECCS Evaluation Model Errors / Changes Reported in Attachment C 2057.QOF 1.
Safety Evaluation for Containment Purging -
+ 10.00F 2.
Safety Evaluation for +/- 60F Uncertainty Band
+ ll.00F 3.
Transition Cycle Penalty
+ 50.00F Licensing Basis PCT
-2128.00F CONCLUSIONS It was determined that compliance with the requirements of 10 CFR 50.46 would be maintained when safety evaluations for changes which affected the
.large break LOCA analysis results were combined with the effect of the L
large break ECCS Evaluation Model errors and changes applicable to Vogtle
' Units 1 and 2 VANTAGE-5 fuel.
l
\\
_-_ ~ -..-. -.-
0 5 *'
- ATTACHMENT D Page 2' SMALL BREAK LOCA DESCRIPTION OF PLANT MOD!FICATIONS The VANTAGE-5 small break LOCA analysis results have been supplemented by a safety evaluation which could affect the PCT as follows:
1.
A safety evaluation was included with the Vogtle VANTAGE-5 small break LOCA analysis to determine the effect of relocating each steam-generator instrumentation line lower level tap and associated changns in each steam generator initial nominal water level.
This evaluation determined that the small break LOCA analysis PCT results could be affected by a 150F-increase.
2.
A safety evaluation to determine the effect of a +/- 60F-uncertainty factor band for the RCS operating average temperature was included with the VANTAGE-5 analysis. This evaluation determined that the small break LOCA analysis PCT results could be 'affected by a 40F-increase.
RESULTANT SMALL BREAK LOCA PCT As discussed.above, the plant modifications could affect the resultant PCT as follows:-
Resultant PCT from ECCS Evaluation Model Errors / Changes Reported in Attachment C 2114.00F Ll. Safety Evaluation for Steam Generator Lower Level Tap j'
Relocation
+ 15.00F 2.
Safety Evaluation for +/- 60/ Uncertainty Band
+
4.00F Total Resultant PCT
=2133.00F CONCLUSIONS It was determined that. compliance with the requirements of 10 CFR 50.46 would be maintained when safety evaluations for changes which affected the
. small break LOCA analysis results were combined with the effect of the small break ECCS Evaluation Model errors and changes (the sum of which are significant) applicable to Vogtle Units 1 and 2 VANTAGE-5 fuel.
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