ML20077E366
| ML20077E366 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 05/31/1991 |
| From: | Fineman C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20077E355 | List: |
| References | |
| CON-FIN-D-6039 EGG-EAST-9410, NUDOCS 9106100077 | |
| Download: ML20077E366 (44) | |
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Attachment EGG EAST.9410 TECHNICAL EVALUATION REPORT FINAL REVIEW Of LOCA ANALYSES FOR THE DUKE POWER COMPANY CA1AWBA AND MCGUIRE UNITS WITH BABCOCK & WILCOX FUEL COMPANY RELOAD FUEL BAW-10174, REVISION 1 C. P. Fineman 4
May 1991 Idaho National Engineering I.aboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. 06039 00 SD0 K k9
e ABSTRACT Babcock & Wilcox Fuel company (BWFC) will supply reload fuel to the Duke Power Company Catawba and McGuire units beginning in 1991.
To support operation of the Catawba and McGuire units with BWFC fuel, loss of coolant accident (LOCA) analyses for Catawba and McGuire were performec with the BWFC recirculating steam generator (R5G) LOCA Evaluation Model (EM) and reported in BAW 10174 Rev.1, Mark BW Reload LOCA Analysis for the Catawba and McGuire Units.
Duke Power Company also supplied information in the above report to justify operation of the Catawba and McG,uire units during the transition period while both Westinghouse Optimized Fuel Assemblies and BWFC Mark BW fuel assemblies reside in the core.
The information was reviewed to determine whether the acceptance criteria of 10 CFR 50.45 are met with the Mark-BW reload fuel and during the transition period.
The information was also reviewed to determine that Nuclear Regulatory Commission approved methods were ' sed in the LOCA analyses.
The review u
found accepted methods were used, and the LOCA analyses provided by Duke Power Company demonstrated the acceptance criteria of 10 CFR 50.46 were met.
Therefore, the INEL recommends the report be accepted to support operation of the Catawba and McGuire units with BWFC reload fuel provided certain conditions are met.
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SUMMARY
Babcock & Wilcox Fuel Company (BWFC) will supply reload fuel to the Duke Power Company Catawba and McGuire units beginning in 1991. To support operation of the Catawba and McGuire units with BWFC fuel, loss of coolant accident (LOCA) analyses for Catawba and McGuire were performed sith the BWFC retirculating steam generator (R5G) LOCA Evaluation Model and reported in BAW 10174, Rev.1, Mark BW Reload LOC A Analysis for the Catawba and McGuire Units. Duke Power Company also supplied information in the above report to justify operation of the Catawba and McGuire units during the transition period while both Westinghouse Optimized Fuel Assemblies and BWFC Mark BW fuel assemblies reside in the core.
Information on For small break LOCAs (SBLOCAs) was also provided by Duke Power Company.
SBL",As, Duke Power Company justified that previous SBLOCA analyses performed for Westinghouse fuel remained bounding for the BWFC reload fuel.
The information provided by Duke Power Company was reviewed to determine whether the Catawba and McGuire units meet the acceptance criteria of 10 CFR 50.45 with the Mark-BW reload fuel and during the transition period.
This was done for both large-and small-break LOCAs.
The review also determined whether Duke Power Ccipany used Nuclear Regulatory Ccemission approved analysis methods to determine compliance.
The review found that approved methods were used by Duke Po,<er Company.
Review of the information provided by Duke Power Company found mixed core operatio:, was adequately justified and the acceptance criteria of 10 CFR 50.46 were met for large-and small-break LOCAs:
0 1.
Peak cladding temperatures were less than 2200.
2.
Maximum local cladding oxidation was less than 17%.
3.
Core wide oxidation was less than 1%.
4.
The core geometry remains amenable to cooling.
5.
Long-term cooling was assured.
Therefore, the INEL recommends the report be accepted to support operation of the Catawba and McGuire units with BWFC reload fuel provided certain conditions are met.
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PREFACE This report was prepared for the U.S. Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, by EG&G Idaho, Inc., Energy and Systems Technology Group.
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CONTENTS ABSTRACT..............................................................- 11 S U MMA R Y............................................................... i i i PREFACE...............................................................
iv 1.
INTRODUCTION..............................................
1 2.
ANALYSIS METHODS.................................................
3 2.1 Overall BWFC RSG LOCA EM...................................
3 2.2 RELAP5/M002 B&W..........................................
11 2.3 FRAP T6.B&W................................................
12 2.4 REFLOD3B...................................................
13 2.5 BEACH......................................................
14 2.6 S u mm a ry....................................................
15 3.
SENSITIVITY STUDIES AND SPECTRUM ANALYSIS.......................
16 3.1 Accumulator Configuration.................................
17 3.2 Ereak Spectrum Analysis....................................
17 3.3 Break Type.................................................
18 3.4 Ma x imum ECCS Fl ow An aly s i s.................................
18 3.5 Summary....................................................
19 4.
COMPLIANCE WITH 10 CFR 50.46.....................................
20 4.1 Large. Break LOCAs..........................................
20 4.1.1 Temperature and Oxidation Criteria..................
20 4.1.2 Cool abl e Geomet ry Criterion.........................
21 4.1.3 Long Term Cooling Criterion.........................
22 4.1.4 L a rg e - B re a k LOC A S unna ry............................
23 4.2 Small-Break LOCAs..........................................
23 4.2.1 Duke Power Company Justification....................
24 4.2.2 INEL Evaluation.....................................
26 4.2.3 Small-Break LOCA Summary............................
28 m
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4.3 Mixed Core Operation.......................................
28 4.3.1 Mixed Core Operation - LBLOCAs......................
29 4.3.2 Mixed Core Operation - SBLOCAs......................
30 4.3.3 M i x e d C o r e S um a ry..................................
32 4.4 O v e r a l l L OC A S u m a ry........................................
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C 0sC t u S i 0n S.....................................................
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6.
REFERENCES.......................................................
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+7 TECHNICAL EVALUATION REPORT FINAL REVIEW OF LOCA ANALYSES FOR THE i
DUKE POWER COMPANY CATAWBA AND MCGUIRE UNITS O
j BABCOCK & WILCOX FUEL COMPANY RELOAD FUEL BAW-10174. REVISION 1 1.
INTRODUCTION Babcock & Wilcox-Fuel Company (BWFC) will supply reload fuel to the Duke Power Company Catawba and McGuire units beginning in 1991. To support operation of the Catawba and McGuire units with BWFC fuel, large break loss of coolant accident (LBLOCA) analyses for Catawba and McGuire were
. performed with the BWFC recirculating steam generator (RSG) LOCA Evaluation j
Model (EM)I and.eported in Reference 2.
Plant performance during small break LOCAs'(SBLOCAs) was addressed by demonstrating the previous
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SBLOCA analyses for Westinghouse. fuel still apply to BWFC Mark CW reload fuel.
In: Reference 2, Duke Power Crmpany also supplied information-to justify operation of the Catawba and McGuire units during the transition period when both Westinghouse Optimized Fuel Assemblies (OFA) and BWFC Mark BW fuel assemblies reside'in the core.
Duke Power Company submitted Reference'2 to the Nuclear Regulatory Commission (NRC), Office of Nuclear.
Reactor Regulation (NRR) for review and approval-by NRR to' justify continued operation of the Catawba and McGuire units with BWFC reload fuel.
The Office of Nuclear Reactor' Regulation requested the assistance of the Idaho National Engineering Laboratory (INEL) in reviewing Duke Power.
Company's--submittal-for Catawba and McGuire.
Specifically, the INEL was requested to review:
. hether Duke Power Company used NRC approved analysis methoc's.
1.
W 2.
Whether the Catawba and McGuire units meet the acceptance
- criteria of 10 CFR 50.46 with the Mark-BW reload fuel and-during the transition period.
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Related to the above reviews, NRR also requested INEL review and evaluate Duke Power Company's responses to NRC questions regarding the Catawba and McGuire LOCA analyses. The NRC questions were transmitted to Duke Power Company in Reference 3.
The responses from Duke Pcwer Cempany are contained in References 4 to 8.
This technical evaluation report documents the review of the Catawba and McGuire LOCA analyses with BWFC reload fuel.
Section 2 of this report documents the review of the methods used by Duke Power Company to complete the LOCA analyses for Catawba and McGuire.
Sensitivity studies and the spectrum analysis are discussed in Section 3.
Cotiformance of the analysis results to the acceptance criteria of 10 CFR 50.46 for large-and small break LOCAs is discussed in Section 4.
Justification of the operation of Catawba and McGuire with mixed cores is also discussed in that section. The conclusions of the review and references are listed in Sections 5 and 6, respectively..
l 2
a 2.
ANALYSIS METHODS This section discusses the analysis methods used by Duke Power Company to perform the Catawba and McGuire LOCA analyses.
The methods were reviewed for NRC approval and conformance to conditions of approval imposed as a result of the NRC review.
The methods reviewed include the everall BWFC RSG LOCA EH and the following computer programs:
RELAPS/M002 B&W,9_FRAP T6 B&W,10 REFLOD3B,Il and BEACH.I2 The above methods will be discussed in the listed order.
2.1 Overall BWFC RSG LOCA EM The overall BWFC RSG LOCA EM i:
scribed ir, Reference 1.
As 1
described in that report, the BWFC RSG LOCA EM uses a series of codes to calculate the system response to a large break LOCA.
The methodology was approved by the NRC.13 A number of the NRC conditions of approval in Reference 13 are related specifically to SBLOCA analyses.
See items 2.j,k,n; 4, 5, and 9 to 12, in Section 4 of the technical evaluation report (TER) enclosed with Reference 13.
Because Duke Power Company justified the Westinghouse SBLOCA analyses remain Dounding for Catawba and McGuire with Mark BW reload fuel, those requirements are not discussed in this report.
The following conditions were imposed on the LBLOCA methodology as a result of the NRC rr iew.
1.
The following items need to be addressed by Duke Power Company in the plant-specific submittal:
a.
Justification for taking the initial rod pressure from the steady state fuel code or a use of a bounding pressure.
Duke Power Company used the initial rod pressure from the NRC approved steady state fuel code TAC 0314,15 in their evaluations, b.
Data to be used for the fuel temperature and moderator void reactivity void coefficients will be provided and justified.
This information was provided by Duke Power Company in their response to question 2, Reference 4.
The Doppler temperature 0
coefficient used was -2.2 pcm/ F, This value is the most 3
negative bounding value for beginning of-life (BOL) c' ore conditions. According to Reference 4, the most negative Doppler 0
coefficist,t at BOL is greater than -1.9 pcm/ F, At end of life 0
(EOL), the Doppler coefficient may approach 2.2 pcm/ F, but according to Duke Power Company the moderator density coefficient at EOL compensates for this variation. Duke Power Company stated the moderator reactivity effect is dominated by the density variation and the moderator temperature effect on reactivity is not used (the r.ost positive temperature coefficient allowed by 0
the Technical Specifications is 0.0 pcm/ F).
The moderator density reactivity is also based on the most positive coefficient allowed by the Technical Specification limits.
Using the NULIF code,16 the boron concentration at BOL was varied until the 0
temperature coefficient was 0.0 pcm/ F; then the density was varied to obtain the density effect on reactivity. This resulted in a curve, provided in. Duke Power Company's response to question 2, that varied the moderator density coefficient from 0.0 % delta-k/k at a relative density of 1.0 to -20.854 % delta k/k at a relative density of 0.2.
Provided the plants are operated within the limits set by the above analysis parameters, the Doppler and Moderator density coefficients used by Duke Power Company are adequate.
c.
Effect of pump trip on L8LOCAs will be evaluated. Duke hwer Company responded to this issue in question 16, Reference 5.
They provided the results of a study that evaluated the effects of leaving the pumps on or tripping them off during the blowdown phase.
The results of the study showed that, although there were differences between the two cases early in the blowdown, the results for the two cases had converged by the end-of-blowdown.
The pumps tripped case gave slightly higher fuel and cladding temperatures relative to the case with the pumps left on. With respect to the reflood phase, Duke Power Company noted that a study in the BWFC RSG LOCA EM topical report (see Section A.2.4, Reference 1) had already shown that a free spinning pump rotor l
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was a benefit to the analysis. Therefore, Duke Power Company concluded, and the INEL agrees, that the pumps tripped case is the most limiting for LBLOCAs.
To clarify the response to question 16, a telecon was held on December 14, 1990 with BWFC, who performed the LOCA ana_ lyses for Duke Power Company, to discuss the effect of pump two phase degradation on the pumps on/off study.
Babcock & Wilcox Fuel company noted the study discussed above was performed using the pump degradation m;1tipliers identified in the BWTC RSG LOCA EM report. That study assumed the pump were tripped.
Based on the EM work and the study for Catawba and McGuire, BWFC concluded that pump operation has only a small effect on the overall transient relative to other parameters.
Therefore, BWFC felt no further work was needed. The INEL reviewed the analyses provided by BWFC and Duke Power Company that considered pump effects and concluded the effect of improved pump performance (leaving the pumps on or decreased two phase degradation) in the BWFC EM is such that it reduces the calculated PCT.
In the pump two phase degradation study in Reference 1, the PCT was approximately 50 to l
75 F lower for the case with the least two phase degradation 0
(see Figure A-36),
in the Catawba and McGuire pumps on/off 0
study, the PCT was approximately 22 F lower for the pumps on case (see Figure 16-6, Reference 5).
Therefore, INEL considers use of the two-phase degradation multipliers identified assuming the pumps were off and pump trip at the start of the transient adequate for use in the pamps nn/off study.
d.
The containment pressure for the LBLOCA analyses will be provided l
and justified on a plant-specific basis. The Catawba containment pressure reported in the Final Safety Analysis Report (FSAR) associated with the maximum emergency core cooling system (ECCS) injection case was used for the LBLOCA analyses reported in i
Reference 2.
Duke Power Company provided additional information l
to support this choice in their response to question 8, Reference 6.
They noted the mass and energy release to the l
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9 containment for the current LBLOCA evaluations was greater than for the Catawba FSAR analysis.
Thus, a containment pressure analysis based on the current mass and energy releases would have been higher than the FSAR curve.
The higher pressure would result in lower
- peak cladding temperatures (PCTs) during the reflood phase of the analyses.
Duke Power Company stafed the containment heat structures and heat transfer surfaces were recalculated in 1987, the recalculated values were used in the minimum containment pressure analysis for the reload calculations, and they periodically review the results to verify the calculations remain conservative with respect to subsequent plant modifications. Duke Power Company also compared the Catawba and McGuire minimum containment pressure analyses and verified the Catawba analysis was lower.
Therefore, INEL considers use of the Catawba FSAR containment pressure adequate to ensure a conservativt containment pressure was used in the LOCA analyses, e.
The ECCS single failure applied to the LBLOCA EM calculation will be justified on a plant-specific basis.
The ECCS single failure chosen for the Catawba and McGuire analyses was the no failure option.
The spillage to the containment from the additional injection resulted in a lower containment pressure, lower reflood rate, and higher PCT.
To clarify the information provided on the single failure in Reference 2, a telecon was held with BWFC on December 14, 1990 to clarify the effect of pump operation, pumps on or locked, on the single failure chosen for analysis.
In the call, BWFC noted the single failure assumption is dominated by reflood effects, not blowdown effects. Therefore, the maximum ECCS flow case still is
.the dominant failure because it results in a lower containment pressure. Also, BWFC pointed out the pump locked rotor assumption already has been shown to be the most severe condition for the reflood phase. With regard to the minimum containment pressure calculation, BWFC noted that the NRC Containment Systems Branch (CSB) requires that the containment sprays and fans be on 6
at their maximum capa' city to maximize the effect on containment i
pressure. The INEL reviewed the pump rotor study in Reference 1, and this confirmed the s'atement by BWFC; the locked rotor is the worst case for reflood. Branch technical position, CSB 61, Minimum Containment Pressure Model for PWR ECCS Performance Evaluation, was also reviewed. This review confirmed BWFC's statement on the NRC CSB required assumptions for the containment sprays and fans for all containment pressure analyses.
Therefore, INEL finds the single f ailure analysis by Duke Power Company adequate for LOCA analyses.
f.
The values used in the REFLOD3B calculations for the fuel and cladding conductivities and the gap heat transfer coefficient will be justified in the plant-specific application:,
Duke Power Company addressed this item in their response to question 1, Reference 4.
They stated the gap conductance and the fuel and cladding conductivities were selected to maximize heat transfer in order to minimize the reflood rate. The gap conductance and the cladding conductivity increase with temperature.
Because the average gap and cladding temperatures during reflood did not 0
exceed 1600 F, that temperature was used to determine the gap conductance and cladding conductivity, in addition the gap conduttance was set assuming the minimum cold gap width, 3.25 mils.
Based on these parameters, the gap conductance was 2
set to 827 BTU /ft -hr OF and the cladding conductivity was set to 13.698 BTU /ft-hr 0F The fuel conductivity decreases with increasing temperature. Therefore, the fuel conductivity was determined assuming a lower bound average fuel temperature of 0
800 F; this resulted in a fuel conductivity of 2.628 BTV/ft-hr OF.
Because the average fuel temperature 0
dropped below 800 F late in the reflood analysis, Duke Power Company reported the results of a study where the fuel 0
conductivity was based on a temperature of 400 F, They noted the effect on the reflood rate was less than 1%. Therefore, Duke Power Company concluded use of a fuel conductivity based on 0
800 was acceptable.
Because of the small sensitivity for the 1
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fuel conductivity ~and_the conservative gap conductance and cladding conductivity used, these parameters are. adequate, g.
_The effects of the maximum pumped ECCS flow versus_ the minimum pumped ECCS flow will be studied on a plant specific basis. The results of this study will be provided to the NRC for r.tview with Duke Power Company's justification for the pumped ECCS flow rates used in the LBLOCA EM analyses.
See item e above.
h.
Justification. should be provided in the LBLOCA EM plant specific submittal that an end-of-life (EOL) blowdown rupture will not occur if an EOL calculation is not supplied. Also, the core crossflow resistance study is not considered valid for blowdown rupture s.-- In Reference 2, it was stated the EOL volume-average fuel-temperature for Catawba and McGuire was slightly less than 1800 F and the rod pressure was 2280 psia. This compares to an 0
0 EOL volume average fuel temperature of 1911 F and a rod pressure of 2280 psia for the EOL case in the BWFC RSG LOCA EM report. Because a blowdown rupture did not occur during the analysis reported in the EM topical report and because those conditions bound the conditions at Catawba and McGuire,-Duke Power Company concluded that a blowdown rupture would not occur at Catawba and McGuire. The INEL agrees with Duke Power Company's conclusion and considers the information supplied in Reference 2 adequate to' resolve the question of a blowdown rupture occurring at EOL, i.
A separate REFLOD3B nodalization study will be completed prior to the first application of the LBLOCA EM to a Combustion
-Engineering (CE)' 2 x 4 plant to confirm the nodalization scheme.
Because Catawba and McGuire are Westinghouse four-loop plants,
-this requirement does 'not apply, J.
Unless justification can be provided that the current licensing base and procedures are still applicable to BWFC reload fuel, the details of the long term cooling methodology will need to be
i justified with the first application of the method to a RSG pl ant.
If a plant-specific long term cooling analysis needs to be performed by Duke Power Company, the analysis methods and results should be justified as discussed in Sections 2.1.6 and 2.2.6 of the TER enclosed with Reference 13. Technical
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specifications that could affect the long term cooling analyses for Catawba and McGuire were not changed with the relond fuel; therefore, new long term cooling analyses were not performed for the plants.
k.
In their RSG LOCA EM, BWFC credited control rod insertion after the end of-blowdown in justifying their approach to calculating fission power durino the refill and reflood portions of a II requires an LBLOCA.
Because the NRC Standard Review plan analysis justifying control rod insertion at the time it is credited, Duke Power Company should justify on a plant specific basis the control rod insertion credited in BWFC RSG LOCA EM.
This should include justification that the core and upper vessel geometry will permit control rod insertion at the time credited in the analysis. Duke Power Company discussed this item in their response to question 4 Reference 6.
A review of the loads on the upper vessel structures for Catawba determined that the insertion of all control rods during reflood could not be guaranteed. However, they noted there was sufficient negative reactivity inserted by sources other than the control rods to provide the negative reactivity assumed in BWFC's analysis in the BWFC RSG LOCA EM without having to credit control rod insertion.
1.
The plant models used by BWFC for SBLOCA EH analyses use dual flow path junction models for the vessel connections to the het and cold legs. This modeling approach may also be used for LBLOCA EM analyses. Use of dual flow path junction models is not recommended by the RELAP5/ MOD 2 code developers because of the possibility of unphysical recirculation flows.
However, BWFC stated in their RSG LOCA EH that this approach will be used in RELAP5/M002 - B&W plant analyses. Therefore, BWFC committed to 9
plotting all junctions within duel flow path groupings and examining the calculated results for unphysical flow patterns.
If any unphysical results are found, then corrective measures will be taken, justification for any EH deviation provided, and NRC approval for the deviation sought. This is adequate to ensure unphysical recirculation due to the dual flow path model will be identified and corrected.
However, the use of two junctions at the hot and cold leg connections also decouples the interaction of the steam and liquid. Therefore, BWFC should also review the results to ensure liquid draining is calculated correctly in cases where the steam flow is sufficient to impede or prevent liquid drainback.
For Catawba and McGuire, Duke Power Company justified the current SBLOCA analyses for the plants are still applicable.
Therefore, for SBLOCAs this requirement does not apply.
For LBLOCAs, the dual flow path model was used for the hot leg to vessel connections, in a call with Duke Power Company and BWFC on November 19, 1990, it was stated that the junction flows were reviewed and no unphysical results fot.
In a call with BWFC on December 20, 1990 it was stated that the hot leg to vessel ficws were reviewed and countercurrent flow was correctly treated with the dual flow path model.
Babcock &
Wilcox Fuel Company reached this conclusion by looking at the steam velocity in the lower path when countercurrent flow was calculated in the lower path r.nd comparing it to the steam velocity in the upper path.
In all cases, the steam flow in the upper path did not exceed the : team flow in the lo'.
path.
Therefore, DWFC concluded the dual flow path model d6d not result in accelerated draining of the hot leg for the large-break LOCA limits analyses. Thus, this requirement was met for LBLOCAs.
m.
The proposed method to meet Criterion 4 of 10 CFR 50.46 is
" adequate for LOCA EH analyses and the effect of PCT on core geometry. However, core geometry concerns for a combined seismic and LOCA loads event also need to be addressed in the plant-specific submittal. This concern was addressed by Duke Power Company in Chapter 10 of Reference 2.
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2.
The BWFC RSG LOCA EM does not account for prerupture cladding strain in-RELAPS/ MOD 2 B&W and BEACH. This position was accepted in the review of the LOCA EM. However, BWFC was required to justify the acceptability of any licensing analyses where cladding swell exceeded 20% but rupture was not calculated, in the Catawba and McGuire LBLOCA analyses, fuel rod rupture was calculated.
For SBLOCAs, the, current SBLOCA were justified as still being applicable.
3.
During the BWFC RSG LOCA EM review, the INEL noted the FRAP T6 B&W analysic of Semiscale LBLOCA experiment S 04 6 had temperatures 0
approximately 250 F different at two elevations that had the same pos.e r.
In a telephone call with BWFC on the EM on August 24, 1990, BWFC stated the differences were the result of the data transfer from RELAPS/ MOD 2 B&W to FRAP-T6 B&W and the interpolation of the core conditions passed between the codes with different core axial noding.
Because this situation has the potential to result in the calculation of lower and, thus, nonconservative PCTs, the NRC required BWFC to review LBLOCA EM analyses where the codes involved in the analysis use different core axial noding to detect oiscrepancies in heat transfer regimes that may result from the data transfer.
In a call with Duke Power Company and SWFC on November 19, 1990, BWFC stated they reviewed each LOCA limit run and the node temperature patterns made sense; the temperature patterns were explainable by power differents., between nodes, location of rupture, and grid locations.
For example for the 2.9 ft LOCA limits case, prior to rupture, nodes 8 and 9 where within 0
0 50 F and nodes 11 and 12 were within 80 F.
Therefore, it appears heat transfer regime discrepancies did not occur in the Catawba and McGuire LBLOCA analyses.
This satisfies the NRC concern.
2.2 RELAPS/M002 - B&W The BWFC RSG LOCA EM used RELAP5/M002 - B&W to calculate the system LBLOCA. thermal-hydraulic blowdown response. The code was approved by the NRC.18 A number of the NRC conditions of approval in Reference 18 are related specifically to SBLOCA analyses.
See Items,5,and 6 in Section 4 of Reference 18.
Because Duke Power Company justified,the Westinghouse SBLOCA
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e analyses remain bounding for Catawba and McGuire with Mark BW reload fuel, those requirements are not discussed in this report. The RELAPS/ MOD 2 - B&W code was approved with the following LBLOCA restrictions:
1.
The CSO film boiling correlation in the core heat transfer model should not be used for licensing applications without additional review and approval by the NRC. According to the BWFC RSG LOCA EM, the Condie B'engston film boiling correlation is used in the core heat transfer model. Therefore, this requirement was met.
2.
Prerupture cladding swell is not modeled because B'WFC indicated it is generally 1tss than 20% with insignificant flow diversion effects.
The NRC safety evaluation report (SER) stated the acceptability of neglecting pretupture cladding swell wotid be resolved as part of the LOCA EH review. That review concluded the neglect of prerupture cladding swell is acceptable, Therefore, the NRC condition was met.
3, for use in LOCA licensing calculations, the NRC SER required the use of static properties for the Extended Henry Fauske and Hoody critical flow models.
The BWFC LOCA EM requires the use of static properties with the above models.
Therefore, the NRC requirement was met.
4.
The NRC SER required BWFC to ensure the decay heat used in LOCA l
licensing analyses complied with Appendix K, i.e., use 1.2 times the l
1971 ANS standard for decay heat.
In response to question 8, L
Reference 19, BWFC compared the RELAPS/M002 - B&W built-in decay heat l
curve to the 1971 standard. The comparison showed that the built-in data adequately. represented the 1971 curve. Therefore, if the built-in data was used in the Catawba and McGuire submittal, it satisfies the NRC concern.
2.3 FRAP-T6-B&u FRAP-T6-B&W was used in the BWFC RSG LOCA EM to calculate the hot rod response (PCT and local oxidation) from the beginning of the accident 20 with the through the end of refill. The code was approved by the NRC following restriction.
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I The NRC required the time steps used in LOCA analyses be consistent with BWFC's response to question 3, Reference 21.
Babcock & Wilcox fuel Company's response to question 3 stated the time steps used comply with 22 INEL guidelines and were validated by sensitivity studies discussed in the BWFC LOCA EM report.
The sensitivity studies were discussed in BWFC's response to question 28. Reference 23, where BWIC provided the specific time steps required to be used in LOCA analyses.
Use of these time steps satisfies the NRC concern.
2.4 REFLOD3B R(FLOD3B was used to calculate the system refill and reflood
- response, in particular, the core reflood rate calculated by REFLOD3B is used as a boundary condition by BEACH in calculating the reflood hot rod response (PC1 and local oxidation).
Revision 2 of the code was approved by the NRC24 with the following restrictions:
1.
The c.EfLOD3B SER required the parameter C be 1 1.05 in REFLOD3B sup analyses.
However, the current version of the BWFC R$G LOCA EM requires C be 2 1.025.
Babcock & Wilcox fuel Company provided 3yp sufficient information regarding this difference in their response to a question (question 4, Reference 19) on the BWFC RSG LOCA EM that the NRC accepted C 1 1.025 for use in LOCA analyses in Reference 13.
sup 2.
The NRC SER required the parameters CT and Cg be equal to 1.0 sec'I and 1.0988 ft'l, respectively, for LOCA analyses.
The B',FC RSG LOCA EM requires that CT and Cg be set at the required values.
3.
A value of 1.0 for the vent valve steam condensation efficiency was required by the NRC SER when the vent valve model is used to represent upper head leakage paths.
This is the value required by the BWFC RSG LOCA EM.
4 The CRf 4 carryover rate fraction correlation and the BWF reflood heat transfer correlation are not to be used for EH calculations. Also, optiun 1 of the REFLOD3B cladding surf ace heat traasfer correlations i
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is not to be used for LOCA analyses; rather, option 2 which includes t
the CRFCKN carryover fraction correlation and the ANC/FLECHT heat transfer correlations are required to be used.
Babcock & Wilcox Fuel Company's RSG LOCA EM requires the NRC specified options be used for LOCA analyses.
I Also, Revision 3 of the REFLOD3B code was approved in Reference 13 i
without additional restrictions.
2.b ALA.CB BEACH was used to calculate the reflood hot rod response (PCT and local oxidation).
Because of BWFC RSG LOCA EM changes, the version of the BEACH code used in the Catawba and McGuire analyses also changed.
The final LOCA limits calculations used to justify conformance to the acceptance criteria of 10 CFR 50.A6 (see Chapter 8 of Reference 2 and 25 Section 4 of this report) used the code version approved by the NRC with the following restrictions:
1..
The acceptability of any licensing analyses where cladding swell exceeded 20% but rupture was not calculated should be justified.
I for the LBLOCA analyses ~, rupture was calculated for the LOCA limits calculations.
Therefore, this requirement does not apply.
- 2. -
In using the revised grid and rupture models with the recommended empirical values of the droplet breakup number, n, and the volume
-length constant, cg, given in Reference _12, the user should ensure the models are applied to plant conditions within the applicable ranges for which the empirical constants were assessed or.must supply.
additional justification for their use.- The applicable ranges were given in the NRC SER.
Based on the information in Reference 2 and References 4 to 8 and obtained in a call with BWFC, the company that performed the LOCA limits analyses for Duke Power Company, on
-December 14, 1990, the INEL concluded the code was applied within the ranges specified in the SER.
1 14
3.
If the multiplier gap conductance model in the previous version of BEACH is used in licensing calculations rather than the acceptable models for dynamic gap conductance and cladding swell and rupture, further justification will be required as specified in the the staff SER on Revision 1 of BAW 10166P. Based on the call with BWFC on December 14, 1990, the multiplier gap conductance model discussed in the NRC SER was not used.
In the December 14, 1990 call, BWFC noted there is the possibility of confusion over the meaning of the gap multiplier model as it applies to BEACH analyses. This is because there are two gap multiplier models in the code, one which is used in BEACH analyses and one which is not used based on the NRC SER. The first model is used to initialize the fuel rod model in BEACH to give the same initial stored energy as the TACO 3 steady state fuel code. This is the same multiplier model used in REL'AP5/M002 B&W, The second model was implemented in an earlier version of BEACH and allowed the code user to vary the gap conductance axially to simulate rupture effects.
This was the model discussed in the NRC SER.
The INEL notes that use of the first model, to initialize the rod model to the TACO 3 results, was accepted in the RELAPS/ MOD 2 BtW review.
Therefore, it is not a concern to use it with BEACH. Also, BWFC indicated the second model is no longer used in BEACH analyses because of the addition of the dynamic gap conductance model from RELAPS/M002 P&W.
2.6 lym)ty Based on the review described above, the Catawba and McGuire LOCA analyses were completed using NRC approved methods and computer programs.
Also, use of the the methods and codes conformed to any NRC conditions of approval.
15
3.
SENSITIVITY STUDIES AND SPECTRUH ANALYSIS Because of the similarities between Catawba and McGuire, the LOCA analyses reported in Reference 2 were performed with a single model representing both plants. However, the INEL noted that the Westinghouse 0
FSAR LBLOCA analyses indicated the PCT was 137 F lower for Catawba relative to McGuire. To clarify the reasons for this difference between the previous LOCA results and the current singic model approach for Catawba and McGuire, Duke Power Company noted that plant specific differences between the two plants could account for the difference in PCT (see question 9. Reference 5).
The major differences between the two plants are the accumulator flow (Catawba higher than McGuire) and pumped injection flow rate late in the reflood period (McGuire higher than Catawba).
These differences were related back to the Westinghouse PCT calculations by noting that the initial temperature rise during reflood for Catawba turned over at a lower temperature relatJve to McGuire because of the larger accumulator flow.
However, because of the lower pumped emergency cort coolant (ECC) flow late in the reflood period, Catawba experienced a second temperature increase where the PCT was calculated.
For McGuire, however, once the cladding terrperatures peaked early in the reflood phase and started to decrease, they continued to decrease for the rest of the reflood phase.
Because the PCT occurred early for McGuire and late for Catawba, the INEL notes that differences in decay heat levels would also be a factor in calculating the different PCTs.
Duke Power Company noted in their response to question 9 that sensitivity studies performed by BWFC were used to account for the Catawba and McGuire accumulator and pumped ECC injection dif ferences in a worst case maMier. However, because of differences in ems between Westinghouse 'and BWFC, this resulted in different results relative to the PCT calculation as discussed in Section 3.1 of this report.
A number of sensitivity studies and the break spectrum analysis were provided by Duke Power Company to justify the initial and boun'dary conditions and break size used in the LOCA limits analysis.
The studies covered the accumulator configuration, break spectrum analysis, break type (double ended or split break), and the maximum versus minimum ECCS flow study. They will be discussed 'in that order.
16
--,r-.
- - - ~
..,,,_,,-.,..-,y,_,
g-...__,m.
,..-,.y,,,
y,--,-
3.1 Accumulator Confiauration One area where Catawba and McGuire differ is in the accumulator configuration.
The differences are due to different gas and liquid volumes and different resistance f actors in the accumulator lines.
Catawba has a 3
larger liquid volume (1050 versus 950 f t ), smaller gas volume (350 3
versus 450 ft ), and lower line resistance factor (5.7 versus 12.4) compared to McGuire.
To determine which plant's accumulator configuration is most limiting with respect to PCT, a study was performed where both 0
The PCT was 4 f higher with the Catawba configurations were analyzed.
accumulator configuration, and the Catawba configuration was chosen for further review.
This result is different from that discussed above; the Catawba accumulator configuration resulted in a lower PCT for Catawba relative to McGuire in the Westinghouse FSAR inalyses.
Duke Power Company noted there were two dif ferences between the Westinghouse and BWFC ems that could account for this (question 9. Reference $).
First, the BWTC EM only allows the fCC to fill the downtomer to the bottom of the cold leg nozzles while the Westinghouse model allowed the downcomer level to increase above the bottom of the cold leg nozzles.
Thus, the additional accumulator injection in the Westinghouse analyste for Catawba was able to provide additional downcomer driving head for the core reflood that was not possible in the BWFC analysis.
Second, because the BWFC analysis assumed complete condensation of steam on the accumulator injection in the cold leg, the additional accumulator flow, once the downcomer was full, represented a penalty and not a benefit as in the FSAR analysis.
3.2 Break Spectrum Analysis The break spectrum analysis was performed for a double ended guillotine break with discharge coefficients of 1.0, 0.8, and 0.6 applied to the break flow. The case with a discharge coefficient of 1.0 resulted in the highest PCT.
Because the break spectrum analysis was completed before the BWFC evaluation model was finalized, Duke Power Company discussed why they 17 m
considered the results of this spectrum analysis valid even for the final tviluation model.
In reviewing the effects of the EM changes Duke Power Company concluded the evaluation model changes had the effect of moving the PCT location one grid span from the ruptured location rather than having the PCT occur in the node adjacent to the ruptured node.
They compared cladding temperatures in the next grid span for the cases with discharge coefficients of 1.0, 0.8, and 0.6.
Although the range of cladding temperatures in the next grid span was smaller, the discharge coefficient of 1.0 case still had the highest PCT.
Because the case with the discharge I
coefficient of 1.0 had the highest PCT in both the ruptured grid span and the adjacent grid span, that case was chosen for the next step in the sensitivity study.
I 3.3 Break Tvee The next study determined whether the guillotine or split break type was most limiting.
A split break with an area equal to twice the cold leg pipe size and discharge coefficient of 1.0 was analyzed and compared to the corresponding guillotine break.
The guillotine break had the highest PCT.
In response to question 12, Reference 6, it was stated the guillotine break is the worst break type because during reflood it allows steam to flow out an area equivalent to one half the break area.
For a split break, the effective area for steam flow from the vessel is less than half the total break area.
This allows more steam to vent from the downcomer for the guillotine break, reducing the downcomer pressure, and lowering the core reflood rate.
Thus, a higFer PCT is calculated for the guillotine break.
To verify the results of the study apply to B'JFC's final version of the EM, Duke Power Company compared the cladding temperatures in the grid span above the one with the rupture location.
The guillotine break also had the highest cladding temperature in this grid span.
Therefore, it was chosen for the next step in the study.
3.4 Maximum ECCS Flaw Analysis To determine the single f ailure that results in the highest PCT, Duke Power Company compared the results of analyses that used the maximum and 18
i minimum [CCS flows.
Related to the discussion at the beginning of Section 3 of this report on the differences between the Westinghouse LM and BWFC EH PC1 results, ;5e minimum pumped injection flow curves used in the study were the smallest of the four units and the maximum pumped injection flow curves used in the study were the largest of the four units.
The maximum [CCS flow case resulted in the highest PCl and was chos,en for use in the LOCA limits analyses.
To verify the results of the study apply to BWFC's final version of the EM, Duke Power Company compared the cladding temperatures in the grid span above the one with the rupture location. The maximum ECCS flow case also had the highest cladding temperature in this grid span.
Therefore, it was chosen for use L. the LOCA limits analysis.
3.5 suamary As a result of the studies performed, the following boundary conditions were chosen for the LOCA limits analysis:
Catawba accumulator configuration, guillotine break with dtscharge coefficient of 1.0, and the maximun, ECCS flow.
Based on the fact the conditions that resulted in the highest PC1 were chosen, IN[L considers the conditions chosen appropriate for use in the Duke Power Company LOCA limits analysis.
~
19
4.
COMPLIANCE WITH 10 CFR 50.46 i
l 4.1 Larae Break LOCAs 4.1.1 Ten erature and 0xidation Criteria Compliance with the temperature and oxidation criteria of 10 CFR 50.46 for LBLOCAs was determined by a direct application of the BWFC R5G LOCA EM.
The analyses were perforrned for a guillotine break with discharge coefficient of 1.0 and the maximum ECCS flow. The total core power was 3479 MW, 102% of 3411 MW.
The large break LOCA limit calculations g
g were run with a total peaking factor (F ) of 2.32 at all elevations up to q
8.0 ft. Above 3.0 f t, the peaking was reduced linearly until F was q
approximately 2.1 at the 12 ft elevation (see Figure B 1, Reference 2).
Duke Power Com; any also imposed a burnup limit on F in the form of a q
multiplier that was 5 1.0.
This yas done to ensure the EOL conditions would not result in a blowdown rupture.
The multiplier was 1.0 up to a burnup of approximately 50,000 MWD /Mtu.
For burnups greater than 50,000 MWD /Mtu, the multiplier reduced linearly until it was approximately 0.85 at a burnup of 60,000 MWD /Mtu (see figure B 2, Reference 2),
Five axial power shapes, with peaks at 2.9, 4.6, 6.3, 8.0, and 9.7 ft were studied.
The limiting axial power shape was the case with the 4.6 ft For this case, the PCT was 1945 f (Reference 7), the peak local 0
peak.
exidation was 4.9%, and the whole core oxidation was 0.55% (question 13.
Reference 5).
All these values are well below 10 CFR 50.46 acceptance criteria of PCT less than 2200 F, local oxidation less than 17%, and 0
whole core oxidation less than 1%.
The response to question 15, Reference 8, showed the uncertainty in PCT due to the possibility of the axial power shape in the plant being different from the ones analyzed was less than 50 F.
Therefore, within the limits allowed by 10 CFR 50.46, J
the first three 10 CFR 50.46 acceptance criteria are met, in the LBLOCA limits analyses, the cold leg temperature is set to a nominal cold leg temperature based on the core flow and the control system response to the core flow (Table 9 2, Reference 1).
The core flow used in the LBLOCA limits cases was the flow determined from the at power 20
4 departure from nucleate boiling (DNB) anilysis.
Because the calculated PCT is sensitive to the cold leg temperature used in the analysis, additional analyses may need to be performed for one of the Catawba or McGuire units if it is operated outside the normal operating mode.
For example, if Duke Power Company extends the operating cycle for one of the units and in doing so the cold leg temperature decreases to below the analyzed tempe,rature, additional analyses are needed to support this operation because it is outside the normal mode of operation.
l l
4.1.2 Coolable Geometry Criterion l
l The fourth criterion of 10 CFR 50.46 requires the core remain in a coolable geometry when both thermal and mechanical loads are considered.
Duke Power Company noted the mechanical loads affect only the outer two or three points in the lattice structure of the core.
Because those fuel rods operated at relatively low power..they did not rupture during the LOCA analyses.
Additional information, provided in Duke Power Company's response to question 19, Reference 5, demonstrated the average core temperature remained below the rupture temperature.
Therefore, Duke Power Company concluded, and the INEL agrees, the thermal and mechanical effects can be considered separately.
Duke Power Company noted the hottest bundle in the core was able to 0
cool successfully and maintain a PCT below 2200 f.
The analyses included the effects of cladding deformation and rupture (the maximum hot channel blockage calculated for Catawba and McGuire was 60%).
Duke Power Company argued that because rupture locations on different rods are not coplanar but distributed within the upper part of a grid span, the bundle retains a coolable rod coolant channel rod arrangement.
The INEL reviewed data on tests with cladding rupture 26,27 and found that it supported Duke Power Company's argument on non coplanar rupture.
With respect to the effects of rod bowing and fuel assembly damage from external forces (pipe breaks and seismic events), the NRC reviewed the
. adequacy of the BWFC Mark BW fuel design in Reference 28.
The NRC found the Mark BW design adequately accounted for rod bowing and maintained a coolable geometry when subjected to a combined LOCA and seismic load.
~
21
t Based on the above considerations. Duke Power Company has adequately demonstrated the cores for Catawba and McGuire retain a coolable geometry following a LOCA.
4.1.3 Lena Term Coolina Criterion i
Criterion five of 10 CFR 50.46 requires the long term cooling of the core be demonstrated after the initial success of the ECC system.
Duke Power Company noted the initial, successful operation of the ECCS is demonstrated by the LOCA limits analyses. Those analyses demonstrated the core would be cooled and the core returned to low cladding temperatures.
They also noted that once the cladding has returned to low temperatures, i
maintaining the low temperature condition requires being able to provide a continuous supply of coolant to the core.
Duke Power Company stated the capability to provide long term coolant supply to the core is demonstrated in the Catawba and McGuire safety. analysis reports (SARs) and is not dependent on the fuel design; therefore, they concluded the previous licensing base remained valid with Mark BW reload fuel.
The INEL agrees with Duke Power Company's conclusion that the previous licensing base remains valid for the Mark BW fuel because it is independent of fuel design.
Also, fuel design was the only change for Catawba r 1 McGuire proposed by Reference 2; that is, no other system design or s'
' c,i specification changes were proposed in addition to the Marl %
> id fuel.
The final long term cooling consideration is preventing boric acid precipitation in the core for cold leg breaks.
In Reference 2, Duke Power Company noted this was accomplished by establishing hot leg recirculation i
core cooling within 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of accident initiation. Demonstrating the sufficiency of the 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> limit is dependent only on showing boric acid concentrations remain below solubility limits.
in turn, this depends only on injection rate, reactor coolant system (RCS) geometry, and core power level.
That is, it is independent of fuel design.
In Reference 2, Duke Power Company stated that, because none of the above factors changed, they concluded the previous evaluation for boric acid precipitation remained i
valid with Mark BW reload fuel.. However, the INEL noted in a conference call with Duke Power Company on September 26, 1990 that, because of the larger diameter of the Mark BW fuel rod, the core volume is reduced by 22
t 3
approximately 38 ft with Mark BW reload fuel.
In a subsequent conference call on October 8,1990, it was noted the boric acid precipitation time was recalculated by-Ouke Pom r Company and the time to hot leg recirculation was now set at nine hours.
During a call on November 5, 1990, it was indicated this calculation would be provided with the Catawba and McGuire updated SARs in 1991 and the operating procedures when the plants restart with BWFC reload fuel, Review of this calculation will need to be completed at that time.
Based on the above findings, the INEL concludes Duke Power Company has adequately addressed the NRC concerns stated in the long term cooling criterion of 10 CFR 50.46.
4.1.4 Larac Preak LOCA Sum *ary The limiting LBLOCA was a gu.1110 tine break with discharge coefficient of 1.0 and the maximum ECCS flow.
The limiting axial power shape was the case with the 4.6 ft peak and total core power was 3479 MW, 102% of t
0 3411 MW.
for this case, the PCT was 1945 F, the peak local oxidation g
was 4.9?.. and the whole core oxidation was 0.55%.
Therefore for LBLOCAs, Duke Power Company provided analyses and sufficient information to demonstrate:
0 1.
Peak cladding temperatures were less than 2200 F.
2.
Maximum local cladding oxidation was less than 17%.
3.
Core wide oxidation was less than 1%.
4 The core geometry remains amenable to cooling.
5.
Long term cooling was assured.
4.2 im_all Breek LOCAs Duke Power Company did not use the BWFC R$G LOCA EH to reanalyze the small break soectrum for Catawba and McGuire with Mark BW reload fuel to show compliance with the acceptance criteria of 10 CFR 50.46 for SBLOCAs.
Rather, Duke Power Company provided information to justify the previous licensing base remained valid for the plants with Mark BW fuel.
A summary of Duke Power Company's justification and the !NEL evaluation follows.
4 23
4.2.1 Duke Power Company Jntification First, Duke Power Company noted previous SBLOCA analyses were not limiting for Catawba and McGuire.
Second, the differences between Mark BW fuel and the Westinghouse fuel are small.
Fuel assembly differences were discussed by Duke Power Company in detail (see below).
Westinghouse and BWFC fuel assemblies differ in the following areas:
assembly unrecoverable pressure drop, initial fuel temperatures, initial internal rod pressures, the axial power shape, and fuel rod diameter.
The effect of each of these areas was evaluated by Duke Power Company, and their justification that the effect of these areas on Catawba's and McGuire's SBLOCA response is small is discussed below.
The difference in assembly pressure drops is less than 1 psi, with the Mark BW assecbly having the smaller pressure drop.
The effect of this difference nn the overall loop flow would be less than l'..
Because the flew and also the core power are the same as in previous analyses, the initial hot leg temperatures for Catawba and McGuire with BWFC reload fuel 0
will be less than 1 F different from the hot leg temperatures with Westinghouse fuel, Therefore, the subcooled depressurization phase of a SBLOCA will be the same regardless of the fuel, and the reactor and pump trip signals will occur at the same time in the transient.
During the flow coastdown and natural circulation phase of the accident, the flow rates are smaller relative to the steady state flow.
Because the difference in assembly presssre drop is small, the effect of the pressure drop difference will be even less d rirg this phase of the accident than during the subcooled depressurization phase.
Duke Power Company noted the flow during this phase of the SBLOCA is sufficient to prevent critic 9l heat flux and transfer all the initial stored energy to the steam generators.
Differences in initial fuel temperature add or subtract overall energy from the RCS.
Information was provided in Reference 2 to show the initial fuel energy is removed from the fuel rod during the pump coastdown phase; therefore: (a) this difference has no impact beyond the pump coastdown 24
__---. = _
phase and (b) the core energy content during the loop draining and boiloff phases will be identica! to the current licensing calculations based on Westinghouse fuel. Add *tional information was provided by Duke Power Company in Reference 5. response to question 20. This response stated that once the resctor trips the energy in the rods decreases and establishes the temperature distri>ution needed to remove decay heat with an average temperature close to the saturation temperature.
In a telecon with Duke
. Power Company ot September 25, 1990, the statement that the core energy content during.'he loop draining and boiloff phases would be identical to the current licensing calculations based on Westinghouse fuel was revised to state that the core energy content would be similar. Also, the response to cuestion 20 points out it is the decay heat during the boiloff phase which will determine the PCT not the stored energy in the core.
Catawba and McGuire will be operated at the same power level witn Mark BW fuel as with 0FA fuel; therefore, the decay heat for the two fuel types will be the same.
Because the mixture level js also expected to be similar for the two fuel types during a SBLOCA (question 23, Reference 5), Duke Power Company concluded the impact of differences in initial stored energy on PCT is minimal.
Fuel rod internal pressures for the two fuel types differ only slightly.
The difference in internal pressure could affect the rod gap dimensions and the time to rupture.
Prior to core uncovery when fuel temperatures are low, the gap differences will be negligible. Duke Power Company also noted that the internal pressure change due to burnup of the two fuel types is such that the Mark BW fuel internal pressure will increase more slowly relative to the Westinghouse OFA fuel.
As a result, by the time the fuel burnup reaches a point where rupture is possible during a SBLOCA, the OFA fuel rod pressure is higher and the OFA fuel would tend to rupture sooner than the Mark BW fuel.
However, because the SBLOCA PCT is approximately 1500, the impact of a difference in rupture timing 0
would be negligible.
Finallys Duke Fcwer Company noted the SBLOCA imposed plant operating limits, including maximum allowable total peaking, are not being changed when BWFC fuel is introduced into the core.
Therefore:
(a) the axial power profile used by Westinghouse in the SBLOCA analyses remains bounding, 25
a 8
e (b) the thermal load imposed on the fuel during a temperature excursion is conservative, and (c) the PCT for the current SAR evaluations are conservative for the Mark BW fuel. Additional information provided by Duke Power Company noted that the axial power shape depends more on plant design and operation rather than fuel assembly design (question 21, Reference 4).
Also, the power shape chosen for evaluation in the SBLOCA analysis is done to ensure a bounding evaluation is performed rather than anticipating that any fuel assembly would experience that axial profile. Also, the Westinghouse axial power shape, which is the same axial power shape that Duke Power Company would use in a direct evaluation of SBLOCAs with Mark BW fuel, will be used as the bounding power shape for the Mark BW fuel n plant operating limits, in response to question 22, Reference 5 Duke Power Company noted that they reviewed the ECCS pump flows, ECC water temperatures, and boron concentrations in the existing SB10CA FSAR analyses.
They stated any technical specification changes needed to ensure plant operation consistent with the analysis assumptions will be proposed for the first Catawba and McGuire reload with Mark BW fuel.
Current SBLOCA analyses also assure 10'.
steam generator tube plugging for both plants. The loop resistances used in current FSAR SBLOCA analyses was reviewed and approved and there have been no changes in the system flow or total loop differential pressure since that analysis. Also, loop resistance will not be altered by the change in fuel type.
Based on the information above, Duke Power Company concluded the current small break analyses in the Catawba and McGuire SARs remain valid as a licensing base even after Mark BW reload fuel is introduced into the Cores.
4.2.2 INEL Evaluation The LBLOCAs at Catawba and McGuire are clearly more * 'iting. Based on updated SARs for the plants,29,30 the PCTs for Catawb. are 1703.7 F 0
0 for the LBLOCA and 1304 F for the SBLOCA.
For McGuire, the PCTs are 0
0 1841 F for the LBLOCA and 14BB F for the SBLOCA.
For Catawba the 4 inch break was limiting, while the 3-inch break was limiting for 26
+
4 McGuire.
This comparison confirms Duke Power Company's statement that, based on past analyses, the LBLOCA is limiting for Catawba and McGuire.
Regarding the difference in assembly pressure drop, INEL concurs with Duke Power Company's judgement that the effect should be small. With the 0
end result being less than a 1 F difference in hot leg temperatute, the initial subcooled blowdown of the plants with Mark BW reload fuel should not be any different than with Westinghouse fuel.
The INEL also agrees that, because of the small differences in assembly pressure drop, differences during the pump coastdown and natural circulation phase would also be negligible.
As discussed in the previous section, infermnion was provided by Duke Power Company to show the initial fuel enert,) is removed from the fuel rod j
during the pump coastdown phase.
The !NEL agrees that the stored energy will be similar rather than identjcal.
This is because differences in rod dimensions for the fuel pellets and cladding would result in differences in stored energy even with the same decay heat temperature distribution for i
the two fuel types.
The INEL also agrees that equality of decay heat, not equality of stored energy, is the important factor in determining the PCT.
Therefore, the difference in initial stored energy is nc,t significant during the core uncovery phase.
Because the fuel rod internal pressures for the two fuel types differ only slightly, it was argued in Reference 2 that the effect on SBLOCAs should be negligible.
Duke Power Company also stated that the OFA rod should rupture slightly earlier than the Mark BW rod in SBLOCAs that occur from fuel burnups where rupture is possible.
The INEL reviewed the Catawba and McGuire33 FSARs and notes the Westinghouse SBLOCA analyses did not calculate any fuel rods to rupture.
Based on the information suppliec by Duke Power Company and the information in the Catawba and McGuire FSARs, the 1NEL concludes that Mark-BW fuel also would not rupture in an SBLOCA analysis based on current Westinghouse calculated hydraulic conditions anti SBLOCA assumptions (e.g., fuel burnup, initial power, steam generator tube plugging ECCS flows, et.c.).
Therefore, the INEL agrees that the effect on SBLOCA analyses will be negligible.
This includes the effect on gap dimensions and rupture.
27
)
Finally, INEL agrees with Duke Power Company's statement that the actual power shape imposed on the fuel is more dependent on plant design Also, because the and operation rather than fuel assembly design.
operating limits for the Hark BW fuel will be based on the power shape used to evaluate OFA fuel and the same axial power shape would have been used by Duke Power Company to evaluate SBLOCAs directly, the INEL concluded sufficient information was provided to justify that the axial power profile used by Westing'toute in the SBLOCA Analyses remains bounding and the PCI for the current SAR evaluations are conservative for the Mark BW fuel.
The information provided by Duke Power Company in response to question 22, Reference 5 regarding the analysis assumptions in the current FSAR SBLOCA analyses and the current operating status of the plants was Because Duke Power Company essentially stated that the also reviewed.
plants aould be operated within the bounds of the current licensing analysis base, that response is considered acceptable.
4.2.3
$*all. Break LO(f Suv ary Based on the above information, the INEt recommends the current SELOC licensing base for Catawba and McGuire be used to license the plants with BWFC reload fuel.
Based on the previous Westinghouse analyses, the 0
limiting PCT for Catawba is 13040 (4-inch break) and 14BB F for McGuire (3 inch break). However, should Duke Power Company modify Catamba or McGuire 50 that the FSAR SBLOCA analysis assumptions (e.g., ECCS flows, steam generator tube plugging, etc.) no longer remain valid, SBLOCA results based on the BWFC R5G LOCA EH will need to be provided to the NRC for Also, the INEL notes that should Duke Power Capany reanalyze review.
SBLOCAs for Catawba and McGuire, the entire SBLOCA spectrum will need to be reanalyzed to identify the worst break size and location (i.e., severed injection line and breaks at the bottom and top of the cold leg pipe which could be sensitive to loop seal phenomena) with the BWFC RSG LOCA EM.
4.3 tib3d Core Operation Because Catawba and McGuire are currently loaded with a, full core of Westinghouse fuel, the plants will operate several cycles with mixed cores, 28
i I
i.e., cores containing both Westinghouse and BWFC Mark BW fuel.
To support operation of Catawba and McGuire with the mixed core, Duke Fower Company discursed the effects of mixed core operation as they relate to LBLOCAs and 5BLOCAs at the plants. The effects of a mixed core on LBLOCAs are discussed first fo' lowed by SBLOCAs.
4.3.1 Mixed Core Operation LBLOCAs Duke Power Company identified the following differences between the Westi'ighouse and BWFC fuel assemblies:
guide tube and instrument tube differences (the upper portion of the guide tube and the whole instrument tube is larger for Mark BW fuel); fuel and cladding dimensions (the Mark BW fuel rod is larger than the Westinghouse fuel, including fuel, gap, and cladding dimensions); and fuel assembly pressure drop (the pressure drop for the Mark BW assembly is slightly lower than the Westinghouse assembly).
To determine the effect of these differences on mixed core operation, Duke Power Company analyzed two cases, one with a Mark BW assembly as the hot assembly and Westinghouse fuel in the average core and one wi'ih a Westinghouse assembly as the hot assembly and Mark BW fuel in the average core.
During blowdown for the case with the Mark BW hot assembly, Duke Power Company stated flow diversion from the Westinghouse assemblies to the Mark BW assembly will decrease the end of blowdown temperature by 30 to 50 f.
However, during reflood, the effect en PCT is in the opposite 0
direction due to the higher core pressure drop of the Westinghouse fuel.
Relative to a core with only Mark BW assemblies, the core refloods slower 0
and the PCT for the Mark BW assembly increases by approximately 30 F.
0 The overall effect is a PCT less than 50 f different from the case with a full Mark BW core.
Because the PCT calculated with the BWFC EM for a full Mark BW core is 1945 F, the PCT for a Mark BW fuel assembly during any 0
period of mixed core operation will also be less than the 10 CFR 50.46 0
accepta'nce criterion of 2200 F.
For the case with the Westinghouse fuel as the hot assembly, the effetts are the reverse of those discussed above for the Mark BW assembly, but '.he overall change is still small.
During blowdown, flow diversion 29
l 11 assembly to the Mark BW assemblies will cause from the Westinghoust the end of blowdown tei;erature of the Westinghouse fuel to increase *0 to However, during reflood, the core will reflood f aster than a core 0
50 f.
with only Westinghouse fuel assemblies and the PCT will be reduced by 0
approximately 30 F.
The overall effect is a PCT less than 50 f 0
Because the PCT different from the case with a full Westinghouse core.
0 calculated by Westinghouse for a full Westinghouse core is 1704 f at Catawba and IB41 F at McGuire, the PCT for a Westinghouse fuel assembly 0
during any period of mixed core operation will also be less than the 0
10 CFR 50.46 acceptance criterion of 2200 F.
For mixed core operation, Duke Power Company showed the LBLOCA PCT for the BWFC fuel may decrease slipMiy from the 1945 f calculated with the 0
0 0
BWFC R5G LOCA EM and may increase slightly from the 1704 F and 1841 f calculated by Westinghouse for Westinghouse fuel in Catawba and McGuire.
Ho ever, the increase for Catawba,and McGuire will be small encugh that the Ig45 F calculated for BWFC reload fuel is still limiting.
0 Based on the above considerations, lHEL concludes Duke Dower Company has aceouately demonstrated the LBLOCA cladding temperatures will remain acceptably low during the transition period with mixed core operation.
4.3.2 Mixed Core Operation SBLOCAs Duke Power Company addressed the effects of mixed core operation on The SBLOCAs in response to questions 23 and 24, References 5 and 6.
differences addressed include unrecoverable pressure drop, initial fuel temperature, initial. fuel rod internal oressure, and the axial profile (question 24) plus the effect of the larger Park BW rod diameter on core levels (question 23).
Concerning the difference in unrecoverable pressure drop, Duke Power Company noted that the pressure drop difference is small and occurs at the Because the entrance is covered with water during the assembly entrance.
core boiloff and the flow rates are so low, any flow related pressure The INEL changes are inconsequential with respect to mixed core operation.
i agrees with this conclusion.
30
l i
Duke Power Company noted the Mark BW assembly will have an initial stored energy content approximately M greater than the Westinghouse OFA.
The mechanisms to remove the differences in initial stored energy were discussed in the response to question 20. Reference $.
That response showed the differences in stored energy would easily be removed during the flow coastdown period of the SBLOCA and any differences at the beginning of the core boilof t would be minimal.
Therefore, there will be no adverse consequences from the difference in initial stored energy during mixed core operation.
The ItiEl agrees with this conclusion and does not see a mechanism by which the difference in stored energy would affect mixed core operation.
Duke Powtr 9 mpany noted the differente in the fuel rod gas internal pressure could iEpact the timing of rupture slightly, but the impact of a rupture or rupture timing difference would be negligible.
Rupture does not affect vessel inventory or cause jnteractions between the two fuel designs.
Duke Power Company in Reference 2, also stated that for SBLOCAs that occur from fuel burnups where rupture is possible the OFA fuel would tend to rupture earlier than the Mark BW fuel.
Therefore, the liiEl agrees that the effect on SBLOCA analyses with mixed cores is negligible.
For the axial power profile, Duke Power Company noted the same axial power profile is used in the SBLOCA evaluation of 0FA fuel by Westinghouse and Mark BW fuel by BWFC. Therefore, the potential difference in assembly power shapes not approaching allowable limits will not have an adverse effect on mixed core operation. The INEL concurs.
Finally, Duke Power Company noted that the geometrical differences between assemblies could have a slightly positive effect on mixed core operation because of a higher minimum core level.
Because the OFA calculations tre the calculations of record for the Mark BW fuci, Duke Power Company concluded there is not an adverse effect on mixed core
-operation.
After reviewing the information from Duke Power Company, the ItiEL agrees.
Duke Power Company showed mixed core operation should not have an impact on SBLOCAs.
Therefore', based on the Westinghouse SBLOCA analyses, 31
~
1 l
)
l 0
0 the SBLOCA mixed core PCTs should remain 1304 F and 1488 F for Catawba and McGuire, respectively.
Based on the above considerations, INEL concludes Duke Power Company has adequately demonstrated the SBLOCA cl* adding temperatures will remain acceptably low during the transition period with nixed core operation.
4.3.3 Mixed Core Su v ar_y for mixed core operation Duke Power Company showed LBLOCAs remain limiting for Catawba and McGuire.
The mixed core PCT for Catawba and 0
McGuire during a LBLOCA should be less than or equal to 1945 F but greater than or equal to 1841 F which exceeds the SBLOCA mixed core PCTs 0
of 14BB F and 1304 F for McGuire and Catawba, respectively.
0 0
For mixed core operation, Duke Power Company stated (Appendix A.
Referente 2) technical specifications for each type of fuel assembly will be based on the full core analyses for each fuel type.
That is, technical specifications for the Mark BW fuel will be based on Reference 2 analyses, and the technical specifications for the OFA fuel will be based on FSAR analyses.
Also, operational limits or technical specifications not directly applied to a fuel assembly will be based on the analysis with generated the most stringent limit.
4.4 Overall LOCA Sumary The above analysis results and information show that Catawba and McGuire will meet 10.CFR 50.46 criteria with BWFC reload fuel.
For LBLOCAs, SBLOCAs, and mixed core operation, Duke Power Company showed LBLOCAs are limiting.
The limiting break was a cold leg guillotiae break with a discharge coefficient of 1.0 and an axial peak at 4.6 ft.
Total core power was 3479 MW, 102% of 3411 MW.
For this break with a full t
t 0
core of BWFC reload fuel, the PCT was 1945 F, peak local oxidatian was 4.9%, and the peak whole core oxidation was 0.55%.
All these values are below the criteria of 10 CFR 50.46.
32
I 5.
CONCLUSIONS 1he information provided by Duke Power Company in BAW 10174, Rev.1.
Mark FW Relogl0C A Analysis for the Cahda and Mdgire Units, and in their responses to NRC questions was reviewed to determine whether the acceptance criteria of 10 CFR 50.46 were met for the Catawba and McGuire plants with Mark BW reload fuel.
Based on our review, the INEL reached the following conclusions and recomends BAW 10174. Rev.1, be accepted to support operation of the Catawba and McGuire units with BWFC reload fuel provided the conditions identified below are met.
1.
For LBLOCAs, the limiting break was a guillotine break with discharge coefficient of 1.0 and the maximum ECCS flow.
The total core power was 3479 MW, 102% of 3411 MW.
The maximum total peaking f actor t
t (F ) was 2.32 at elevations up to 8.0 ft.
Above 8.0 ft, the peaking q
was reduced linearly until Fg was approximately 2.1 at the 12 ft elesation, lhe pC1 was calculated for the case with the axial peak at 4.6 ft.
The BWFC RSG LOCA EM was used in these analyses.
Therefore, Duke Po.er Cotpany provided analyses and sufficient information to ctmenstrate the acceptance criteria of 10 CFR 50.46 were met:
1.
Peak cladding temperatures were less than 2200 f (19450f).
0 2.
Maximum local cladding oxidation was less than 17% (4.9%).
3.
Core wide oxidation was less than 1% (0.55%),
4.
The core geometry remains amenable to cooling.
5.
Long term cooling was assured.
2.
For SBLOCAs, Duke power Company did not use the BWFC R5G LOCA EM to reanalyze the small break spectrum.
Rather, they provided sufficient information to demonstrate the current SBLOCA licensing base for Catawba and McGuire is adequate to license those plants with Mark BW reload fuel.
Based on the previous analyses, the SBLOCA PCTs for 0
0 Catawba and McGuire are 1304 F (4 inch break) and 148B F (3-inch break), respectively.
Therefore, 10 CFR 50.46 acceptance criteria were met for SBLOCAs.
3.
Based on the above, LBLOCAs are limiting at Catawba and McGuire.
33
>l.
s l
4 For a full Mark BW core, the calculated PCT was 1945 F, and for a 0
full OfA core, the calculated PCTs were 1704 F and 1841 f for 0
0 Catawba and McGuire, respectively. During mixed core operation, Duke Power Company showed the calculated PCT would be bounded by 1945 F 0
as calculated for a full Mark BW core during a large break L,0CA.
Thertfore, the acceptance criteria of 10 CFR 50.46 will be met for Catawba and McGuire during the period of mixed core operation when both Westinghouse OFAs and BWFC Mark BW fuel assemblies will reside in the core.
5.
The methods used to perform the LOCA analyses for Catawba and McGuire were approved by the NRC, and the methods were applied in a manner consistent with any NRC conditions of acceptance.
6.
Because of the larger diameter of the Mark BW fuel rod, the core volume is reduced by by approximately 38 ft3 with Mark BW reload fuel.
In a conference call en October 8, 1990, Duke Power Company noted the boric acid precipitation time was recalculated and the time to hot leg recirculation was now let at nine hours.
During a call on Novenber 5, 1990, it was indica'ed this calculation would be provided withtheCatatpaandMcGuireupdatedSARsin1991andtheoperating procidures whel. the plants restart with BWFC reload fuel.
Review of thiscalculatijnwillneedtobecompletedatthattime.
7.
Information on the fuel temperature and moderator void reactivity void coet'ficients used in the LOCA analyses was provided by Duke Power Com; any. The Doppler temperature coefficient used was -2.2 pcm/ f.
O The moderator density reactivity is shown in a curve, provided in Duke Power Company's response to question 2, Reference 4, that varied the mederator density coefficient from 0.0 % delta-k/k at a relative density of 1.0 to 20.854 % delta k/k at a relative density of 0.2.
The plants should be operated within the bounds set by the above analysis limits for the Doppler and Moderator density coefficients.
34
8.
As discussed in Section 4.2.of this report, Duke Power Company provided sufficient information to justify the current licensing base for SBLOCAs was still applicable to Catawba and McGuire with BWFC reload fuel. However, should Duke Power Company modify Catawba or McGu{re so that the FSAR SBLOCA analysis assumptions (e.g., ECCS flows, steam generator tube plugging, etc.) no longer remain, valid, SBLOCA analysis results based on the BWFC RSG LOCA EM will need to be provided to the NRC for review. Also, the INEL notes that should Duke Power Company reanalyze SBLOCAs for Catawba and McGuire, the entire SBLOCA spectrum will need to be reanalyzed to identify the worst break size and location (i.e., severed injection line and breaks at the bottom and top of the cold leg pipe which could be sensitive to loop seal phenomena) with the BWTC RSG LOCA EM.
9.
In the LBLOCA limits analyses, the cold leg temperature is set to a nominal cold leg temperature, based on the core flow and the control system response to the core-flow. The core flow used in the LBLOCA limits cases was the flow determined from the at power DNB analysis.
Because the calculated PCT is sensitive to the cold leg temperature used in the analysis, additional analyses may need to be performed for one of the Catawba or McGuire units if it is operated outside the normal operating mode.. For example, if Duke Power Company extends the operating cycle for one of the units and in doing so the cold leg temperature decreases to below the analyzed temperature, additional analyses are needed to support this operation because it is'outside the normal mode of operatlon.
10.
For mixed core operation, Duke Power Company stated technical specifications for each type of fuel assembly will be based on the full core analyses for each fuel type.
That is, technical specifications for the Mark-BW fuel will be based on Reference 2 malyses, and the technical specifications for the OfA fuel will be based on FSAR analyses. Also, operational limits or technical specifications not directly applied to a fuel assembly will be based on the analysis with generated the most stringer. limit.
4 35
,i,.,
6.
REFERENCE'S 1.
Babcock & Wilcox Nuclear Power Group, gly loss of Coolant Acciden L Evaluation Model for Recirculatino Steam._Ging11er Plants. BAW 10168P, Rev. 1, September 1989.
2.
Babcock & Wilcox Fuel Company, Mark BW Reload LOCA Analysis for the
(_atawba and McGuire Units, BAW 10174 Rey, 1, November 1990.,
3.
Letter from D. S. Hood, NRC, to H. B. Tucker, Duke Power Company,
" Request for Additional Information on BAW 10174, Mark BW Reload LOCA Analysis for the Catawba and McGuire Units (TACs 75138/75139/75140/75141),* Harch 27. 1990.
4.
Letter from H. B. Tucker, Duke Power Company, to USNRC Document Control Desk, "McGuire Nuclear Station, Docket Numbers 50 369 and 370, Catawba Nuclear Station Docket Numbers 50 413 and 414 Response to Request for Additional Information Regarding BAW 10174 (TACS 75138/75139/75140/75141)," June 7, 1990.
5.
Letter from H. B. Tucker. Duke Power Company, to USNRC Document Control Desk, 'McGuire Nuclear Station, Docket Numbers 50 369 and -370, Catawba Nuclear Station, Docket Numbers 50 413 and 414, Response to Request for Additional Information Regarding BAW 10174 (TACS 75138/75135/75140/75141)," July 25, 1990.
6.
Letter from H. B. Tucker, Duke Power Company, to USNRC Document Control Desk, McGuire Nuclear Station, Docket Numbers 50 369 and.370, Catawba Nuclear Station, Docket Numbers 50 413 and -414, Response to Request for Additional Information Regarding BAW 10174 (TACS 75138/75139/75140/75141)," August 8, 1990, 7.
Letter from H. B. Tucker, Duke Power Company, to USNRC Document Control Desk, "McGuire Nuclear Station, Docket Numbers 50 369 and.370, Catawba Nuclear Station Docket Numbers 50 413 and 414 Response to Request for Additional Information Regarding BAW 10174 (TACS 75138/75139/75140/75141)," September 4, 1990.
B.
Letter from M. S. Tuckman, Duke Power Company, to USNRC Document Control Desk, 'McGuire Nuclear Station, Docket Numbers 50 369 and 370, Catawba Nuclear Station, Docket flumbers 50-413 and 414, Topical Report BAW 10174, Mark BW Reload LOCA Analysis for Catawba and McGuire, Revised Responses to Questions (TACS 75138 141)," November 7,
- 1990, 9.
J. A. Klingenfus et al., E[ LAP 5/M002 B&W:
An Advanced Computer Procram for licht Watn Reactor LOCA and Non LOCA Transient Analysis, BAW 10164P, Rev. 1, Babcock & Wilcox Nuclear Power Group, October 1988.
10.
J. H. Jones, FRAp-T6 B&W - A Computer Code for the Transient Analysis of Licht Water Reactor fuel Reds BAW 10165P, Rev.1,. ctober 1988.
0 11.
C. K. Nithianandan, REFLOD3B. Model for Multinode Core Refloodino _
Analytis, BAW 10171P, Rev. 3 Babcock & Wilcox Nuclear Power Group, September 1989.
36
12.
N. H. Shah, BEACH:
Best Estimate Analysis Core Heat Transfer. A Computer Procram for Reflood Heat Transfer Durina LOCA, BAW 10166P, Rev. 2, Babcock & Wilcox Nuclear Power Group, September 1989.
13.
Letter from A. Thadani, NRC, to J. H. TayTor, B&W, ' Acceptance for Referencing of Licensing Topical Report, BAW 10168P, Revision 1. RSG LOCA B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants " January 22, 1991.
~
1 t..
Letter from J. H. Taylor, B&W, to C. Berlinger, NRC, ' TACO 3 - fuel Pin Thermal Analysis Computer Code,' JHT/87-75, April 10,1987.
15.
Letter from A. Thadani, NRC, to J. H. Taylor, B&W, ' Babcock & Wilcox Topical Report BAW 10162P, TACO 3
- Fuel Pin Thermal Analysis Computer Code," August 14, 1989.
16.
Letter from S. A. Varga, NRC, to K. E. Suhrke, B&W, " Evaluation of BAW 10115," December 8, 1976.
17.
USNRC, Division of Licensing, Standard Review Plan, NUREG 0800, July 1981.
18.
Letter from A. Thadant, NRC, to J. H. Taylor, B&W, " Acceptance for Referencing of Licensing Topical Report BAW 10164P, RELAPS/M002 B&W, An Advanced Computer Program for Light Water Reactor LOCA and Non LOCA Transient Analysis," April 18, 1990.
19.
Letter from J. H. Taylor, B&W, to V. Wilsen, NRC, JHT/90 37, March 12.
1990.
20.
Letter from-A. Thadani, NRC, to J. H. Taylor, B&W, Acceptance for Referencing of Licensing Topical Report BAW 10165P, Revision 1, A Computer Code for the Transient Analysis of Light FRAP T6 B&W Water Reactor Fuel Rods," May 19, 1989.
21.
Letter from J. H. Taylor, B&W, to J. A. Norberg, NRC, 'fRAP T6 B&W Topical Report BAW 10165P,' JHT/88 207, November 11, 1988.
22.
L. J. Stefken et al., FRAP T6:
A Computer Procram for the Transient--
Analysis of Oxide Fuel Rods, NUREG/CR-2148. EGG 2104, May 1981.
23.
Letter from J. H. Taylor, B&W, to J. A. Norberg, NRC, "RSG LOCA Topical Report, BAW 10168P," JHT/89 66, March 31,1989.
24.
Letter from A. Thadant, NRC, to J. H. Taylor, B&W, " Acceptance for Referencing of Licensing Topical Report BAW 10171P, Revision 2.
REFLOD3B Model for Multinode Core Reflooding Analysis," May 19,-
1989.
25.
Letter A. Thadani, NRC, to J. H. Taylor, B&W, " Acceptance for Referencing of Licensing Topical Report BAW 10166P, Revision 2.
BEACH Best Estimate Analysis Core Heat Transfer, A Computer Program for Reflood Heat Transfer During LOCA* August 13, 1990.
26.
K. Wieher and U. Harten, Datenbericht REBEKA 6, KfK 3986, March 1986.
37
v-
., v, 27.
F. J. Erbacher and S. Leittikow, 'Zircaloy Fuel Cladding Behavior in a loss of Coolant Accident: A Review,' ASTM STP 939, 1987, pp. 451 488, 28.
Letter from A. Thadant, NRC, to J. H. Tay'ir, B&W, ' Acceptance for Referencing of Licensing Topical Report 8, 10172P, Mark BW Hechanical Design Report (TAC NO. 68873), December 19. 1989.
29.
H. B. Tucker, Duke Power Company, to USNRC Document Control, Desk,
' Catawba Nuclear Station Units 1 and 2 Docket Nos. 50 413 and 50 414, 1987 Final Safety Analysis Report Annual Update,"
June 30, 1988.
30.
H. B. Tucker, Duke Power Company, to USNRC Document Control Desk,
'HeGuire Nuclear Station, Docket Nos. 50 369, 370, 1988 FSAR Update,"
June 22, 1989.
31.
H. B. Tucker, Duke Power Company, to USNRC Document Control Desk,
'McGuire Nuclear Station Docket Nos. 50 369, 370, Annual FSAR Update,' July 23, 1987.
e D
=
38
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