ML20247B768
| ML20247B768 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 05000000 |
| Issue date: | 07/17/1989 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8907240202 | |
| Download: ML20247B768 (5) | |
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HAL B.TUCiKER 1 Trupesorte X.2-vium purennewt (y04) ay3 4333
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July 17, 1989 6,
' Document Control Desk n@
U. S.. Nuclear Regulatory Commission
. ashingtoc, D. C.
20555 W
Subject:
McGuire Nuclear Station
. Docket Nos. 50-369 and 50-370 Catawba Nuc1 car Station H.
- Docket Nos.-50-413 and 50-414 v
Main Steam Lino Breaks in Ice Condenser Plants
. Attached is Duke Powor's' responce to your letter of May 11, 1989,' sol'. citing comments on Argonne National Laboratory's' report on the subject. It is hoped >
'that with this information, this long standing issue can be resolved.
Very truly.yours,
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Hal B. Tucker PGL/IV/50' Attachments _',4.pages)
I xc: S. D.'.Ebneter, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101'Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
.Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station Mr. P.'K. Van Doorn NRC Resident Inspector McGuire Nuclear Station i
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DUKE POWER COMPANY RESPONSE TO NRC LETTER DATED MAY 11,.1969 1
L McGuire Nuclear Station Catawba Nuclear Station Main Steamline Break Superheat Issue Duke Power has reviewed Argonne National Laboratory's evaluation of the COBRA-NC analysis.
The following response addresses comments raised in the May.11, 1989 letter from NRC along with other issues vital to this subject:
I. Introduction In 1983, the NRC expressed concern that superheated steam, rather than saturated, might be released in a steamline break accident subsequent to tube uncovery. Westinghouse modified their mass and energy release models tc reflect this change and then performed some containment l
' response analyses using the LOTIC-3 code.
Inclusion of superheated steam led to predicted global lower containment temperatures well in excess of the current Equipment Qualification (EQ) limit of 327 F, Westinghouse proceeded to remove some of the conservatism in LOTIC-3 by including the previously neglected ice condenser drain flow. This addition decreased the predicted temperature back below the EQ lbnit. Westinghouse then contracted for an-independent evaluation of the new version of LOTIC-3 i
using COBRA-NC (Reference 1).
The COBRA-NC resultn confirmed global lower containment temperature predictions by LOTIC-3, but'the multi-volume representation of lower containment used in the COBRA-NC analysis predicted peak temperatures in the break vicinity slightly above l
the EQ limit.
NRC has received evaluations of both the ice condenser drain model and COBRA-NC analysis.
The drain model, reviewed by Los Alamos National Laboratory (LANL), wcs found to hare some minor errors. westinghouse corrected these, but there was almost no change in the predicted temperature.
The COBRA-NC analysis was reviewed by Arconne National Laboratory (ANL) using I
the COMM1X (Reference 2) code. COMMIX predictions of global lower containment temperature agreed well with those from COBRA-NC.
It has been concluded that global temperatures do not precent a saperheated steam concern.
However, local temperatures in the vicinity of the braag can be superheated.
The purpose of tois document is to summarize the evaluation performed by Duke Power that addresses localized suparheated steam.
II. COBRA-NC/ COMMIX Ccmparison and Analysis Rasults 1
The COBRA-NC code is the state-of-the-art containment thermal-hydraulic simulation code.
Reference 3 contains a detailed description of COBRA-NC. COBPA-NC itodels a two-phase mixture, three momentum equatioris, two energy equations, and accounts for the transport of entrained liquid droplets.
COBRA-NC has benefitted from substantial validation efforts, L
some of which era documented in Reference 3.
COMMIX is a more limited
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code, in the context of containment simulation capabilities, since it is
a decigned to solve single-phase problems, and, as such, cannot model a j.
L two-phase mixture.
COMMIX has only one mixture momentum and one mixture energy equation.
Therefore, Duke Power asserts that local temperature predictions'from COBRA-NC are significantly'more acen. ate than those of g
COKMIX. Duke will address the impact of localized superheated temperature effects as predicted by the COBRA-NC analysis.
In the COBRA-NC analysis, temperatures above the EQ limit of 327 F are confined to the cell containing the break; adjoining cells have peak l
temperatures loss than this limit.
In the COBRA-NC model, the nodalizatien and selection of the break location intentionally maximized the resulting superheated temperatures.
In the COBRA-NC analysis three different models are used: Model I contains no ice condenser drain flow; Model 2 is identical to Model 1, but includes drain flow; Model 3 is identical to Model 2, but has finer noding in the break vicinity. In Model 3, the break is located in cell 107, a representation of wb2ch is b
I' given in Figure 1.
The region of temperatures in excess of 327 F is limited to distances from the break less than or equal to the maximum cell dimension, which is 13.5 feet. Beyond this distance, temperatures remain below the EQ limit, and therefore the temperatures are acceptable.
1 The approach taken to resolve temperatures in excess of 327 F is to evaluate all applicable EQ equipment within 13.5 feet of any steam line in the lower containment.
III. Results and Ccnclusion To determine what equipment is potentially affected (necessary to bring the plant to hot shutdown following a steam line break in co,ntainment),
appropriate Emergency procedures were reviewed and the essential equipment identified.
If the equipment was located in lower containment, the distance between a particular piece of equipment and the nearest main steamline was determined. Results show the minimum distance to be 24 feet for Catawba and 21 feet for McGuire.
Since all of the equipment necessary to bring the plant to hot shutdown following a steamline break is well beyond the critical distance of 13.5 feet, mitigation of a steamline break accident would be unaffected by peak local temperatures in excess of the EQ limit.
In conclusion, both the COBRA-NC and COMMIX analyses confirm that the new LOTIC-3 version edequately predictc global lower containment temperature, which is an acceptable criterien for determining equ oment qualification
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requirements per NUREG-0588. Additionally, COBRA-NC does predict local
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temperatures near the break that exceed the EQ limit. An evaluation for the McGuire and Catawba Nuclear Stations has determincd that no equipment required for main steam line break mitigation is located close enough to the break location to be affected. Therefore, there are no remaining potential safety concerns associated with superheated steam inside containment for the McGuire and Catawba Nuclear Stations.
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References:
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- 1. ' COBRA-NC Analysisfor a Main'Steamline Break in the Catawba Unit 1
- Ice Condenser Containment, WCAP-10988, Hovember, 1985,-
- 2. ' COMMIX Analysis of a Main Steamline Break in the Catawba Lower
-Containment, Abgonne National-Laboratory, February 1989.
c 3.
COBRA-NC - A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor Components, NUREG-CR-3262, March 1986.
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