ML19259B305
| ML19259B305 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe, Maine Yankee |
| Issue date: | 01/17/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19259B304 | List: |
| References | |
| NUDOCS 7901260039 | |
| Download: ML19259B305 (8) | |
Text
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Topical Report Evaluati m Report Number: YAEC-ll60 Report
Title:
Application of Yankee-WREM-Based Generic PWR ECCS Evaluation Model to Maine Yankee (3 Loop Sample Problem)
Report Date: July 1978 Originating Organization: Yankee Atomic Electric Company Sumary of Topical Report The topical report presents results of a representative LOCA analysis for the Maine Yankee plant using YAEC's modified WREM-Based ECCS Evaluation This model is based on Exxon's ENC /WREM(I) model, and was modified Model.
for a previous reload licensing application for the Yankee Rowe plant.
Documentation of these modifications (b5) has been reviewed and found acceptable by the NRC staff. The only variation from previously accepted analytical procedures used in this analysis consists of an updating of the hot wall delay model used for manually computing the refiil interval between the End-of-Bypass (E0BY) to the time refill water begins to enter the core, or Bottom-of-Core Recovery (B0CREC). The hot wall delay model employed by YAEC is identical to that reviewed and accepted by the hRC for use in Exxon's Evaluation Model(6), and is based on recent experimental data obtained from CREARE tests (7,8) which results in slight variations in hot wall delay from the previous CREARE model.
The plant nodalization detail used in the blowdown and reflood models are identical to those used in a previous analysis by YAEC on the Yankee Rowe plant,I9) and by Exxon on the H.B. Robinson plant (10)
Containment pressure used in the sample calculation presented in this report was based on a Combustion Engineering analysis performed for Maine Yankee Cycle 3 loading, but the licensing calculations to be submitted for the Maine Yankee reload license and subsequent licensing submittals will use containment pressure 790126CC39
. calculations performed with the CONTEMPT-LT Version 26 code.
Topical Report Evaluation The YAEC Evaluation Model used in this calculation for the Maine Yankee plant consists essentially of the same model used in previous analyses of the Yankee Rowe plant which has been reviewed and accepted by the NRC staff. The only variation to the previous analytical process consists of using CREARE's updated hot wall delay model in the manually computed refill interval between blowdown and refill.
This model was reviewed and found acceptable in a previous Exxon Nuclear Company ECCS model update (6)
The analysis performed for this sample calculation considered a double ended guillotine LOCA on the discharge side of one of the primary loop pumps.
Comparison of results is shown on Table 1, however; the comparison is only general due to an important difference between assumptions used in the calculations. The previous reload analysis for Maine Yankee was performed by Combustion Engineering, and used a top-skewed axial power distribution. The YAEC analysis presented in this report assumed a chopped cosine distribution which would be expected to result in earlier hot spot quench in reflood and reduced PCT as is shown on the table.
The sample calculation presented in the report was chosen to provide the most rapid depressurization transient for use in the time step sensitivity studies conducted as part of the model confirmation studies, and was not intended for use in fuel cycle comparisons.
Direct compari-sons between fuel cycle LOCA characteristics will be possible when the licensing calculations have been submitted.
. A time step sen'sitivity study was perfonned for the blowdown, hot channel, reflood, and hot pin calculations by reducing the nominal anaximum and minimum time step limits used in the sample calculation to one-half the nominal values for each time interval used in the transient, Because of the interval time step adjustments made in the codes, these time step limit changes will not necessarily impose reduced or increased time steps throughout the LOCA calculation, but do demonstrate that the time step range allowed within the nominal limits produces convergent results with a lower limit range.
PCT sensitivity to axial power shape was also assessed for a peak linear power rate of 14.5 kw/ft. A comparison between a chopped cosine axial distribution and a top-skewed peak having the maximum axial offset permitted for Cycle 4 demonstrated a 370"F higher PCT for the offset distribution.
Sensitivity of the LOCA response transient and PCT to break noding, steam generator noding, passive heat slab noding, and lower plenum noding were not performed by YAEC for Maine Yankee. Such studies have been performed for the Yankee Rowe plant or the H.B. Robinson plant by Ixxon in previous reload licensing applications,:and have-demonstrated that the nodalization models
-used for these plants and for Maine Yankee have provided converged results with the more detailed models.
Break spectrum and fuel burnup LOCA sensitivity have not been included in this <focument as they will be the subject of the licensing application
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to be submitted in the near future.
.a.
Several errors in T00DEE2 were identified by consultants to the Swedish Nuclear Power Inspectorate.
Five of these apply to the YAEC model.
YAEC corrected four of the errors which resulted in less than 1 degree F change in peak cladding temperature (PCT).
(References 11 & 13).
Correcting the fifth problem, resulted in a reduction in PCT of about 6 F.
This last item only applies during periods of reflood steam cooling.
Since the PCT was reduced, it was necessary to determine that the steam cooling model was still conservative compared to FLECHT data.
This comparison was performed for the same FLECHT experiments used by ENC for this model. The results showed that the unblocked steam cooling model still had approximately the same degree of margin as the uncorrected model (References 12 & 13). Therefore, we find that the corrections made to T00DEE2 are acceptable, and require no additional changes to the steam cooling model.
. Table 1 - Maine Yankee Cycle 4 and 5 LOCA Comparison Double Ended Guillotine Break, CD"I Cycle 4 Cycle 5 1CE Anal.)
TYAEC Anal.)
16.5 16.5 Peak Linear Power at t = 0 (Kw/ft) 2125 2059 PCT (*F) 250 137 Time to PCT (sec.)
13.6 2.3 Local M/W Reaction (%)
Differences in PCT and time to PCT are due to use of top-skewed axial axial power distribution in Cycle 4 analysis and to cosine Note:
pcwer distribution for Cycle 5 analyses.
. Staff Position The ECCS Evaluation Model proposed for t'.e Maine Yankee LOCA analysis is essentially identical to that used and accepted for a previous reload The changes to licensing application for the Yankee-Rowe plant.
be used for Maine Yankee analyses consists of the updated hot wall delay calculation in the refill interval which has been accepted for use by As the hot wall delay calculation is Exxon, and the T00DEE2 corrections.
performed manually, no computer program change is involved requiring computational verification.
As a result of the findings summarized above, it is concluded that the YAEC Evaluation M0 del described in this topical report and in references 1 to 6, and 11 to 13 will be acceptable for PWR reload licensing applications on a generic basis upon completion of the documentation requirements specifie by paragraph c, part II of 10 CFR 50, Appendix K concerning computer listings Submittal of these listings is required prior for accepted evaluation models.
to, or upon, submittal of the next YAEC reload application.
The YAEC EM as identified in this report consists of the following codes:
RELAP4 EM (YAEC-058) for blowdown and hot channel analysis 1.
T00DEE2 (YAEC-03) for hot channel analysis during refill and reflood 2.
3.
RELAP4-FLOOD for reflood analysis.
The current YAEC versions of these codes as described in Referenc and 11 to 13, and in this topical report constitute the accepted YAEC-EM for reload licensing applications.
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. References 1.
XN-75-41, Volumes 1 to 3 and all Revisions, Supplements, and Appendices.
ENC-WREM Based Generic PWR-ECCS-Evaluation Model, July 25, 1975.
2.
P.T. Antonopoulos, A. Husain, " Method for Calculating End-of-Bypass Time for Yankee Rowe Loss-of-Coolant Accident Analysis". YAEC-1125, (March 1977).
3.
J. Consolatti, A.S. Hanson, A. Husain, J.C. Turnage, " Method for Calculating Low-Flow Film-Boiling Coefficients for Yankee-WREM-Based Generic PWR ECCS Evaluation Model", YAEC-1131, (June,1977).
4.
XN-76-44, " Revised Nucleate Boiling Lockout for ENC-WREM-Based ECCS Evaluation Models", September,1976.
5.
J. Consolatti, " Core Flood Rate Stabilization for Yankee-WREM-Based Generic PWR ECCS Evaluation Model", YAEC-1133, (July,1977).
6.
XN-76-27:
Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update.
ENC-WREM-II, July 1976.
7.
Block, J. A., and Crowley, C.J., " Hot Wall Experiments in a Simulated Multiloop PWR Geometry", CREARE-TN-202, February 1975.
8.
Block, J. A., et.al., " Analysis of ECC Delivery", CREARE-TN-231, April 1976.
9.
XN-75-58: Loss of Coolant Accident Analysis for Exxon Nuclear and Gulf Reload Fuel for Yankee Rowe, October 30, 1975.
10.
XN-75-41 Volume II, Appendix D, " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model, 3 Loop Westinghouse Large Break Exemple Problem (Using September 26, 1975 Model), October 2,1975.
8-References (Continued)
D.E. Vandenburgh, Letter to U.S. NRC, B3.2.1, WMY 79-1, Jan. 8, 1979.
11.
D.E. Vandenburgh, Letter to U.S. NRC, WMY 79-3, Jan.15,1979.
12.
13.
D.E. Vandenburgh, Letter to U.S. NRC, dated Jan. 16, 1979.
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