ML20044A352

Submits Info Re Sensitivity Study Performed on Number of Fuel Axial Intervals,Per Topical Rept, American Electric Power Reactor Core Thermal-Hydraulic Analysis Using Cobra III-C/MIT-2 Computer Code.

ML20044A352
Person / Time
Site:	Cook
Issue date:	06/22/1990
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:1081A, NUDOCS 9006280367
Download: ML20044A352 (5)

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DNDQANA MfCNf0AN POWER AEP:NRC: 1081A Donald C.-Cook Nuclear Plant Units 1 and 2 Docket Nos, 50 315 and 50 316 License Nos. DPR 58 and DPR-74 REQUEST FOR REVIEW OF AEP REACTOR CORE THERMAL-HYDRAULIC ANALYSIS USING THE COBRA IIIC/MIT 2 COMPUTER CODE U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Attention: T. E. Murley June 22, 1990

Dear Dr. Murley:

This letter references a telephone conversation with your staff on May 31, 1990 on the topical report entitled, "AEP Reactor Core Thermal-Hydraulic Analysis Using the COBRA III-C/MIT-2 Computer Code," submitted on January 30, 1989 for review and approval.

During this conversation, the reviewer of this report asked us to provide the sensitivity study performed on-the number of fuel axial intervals used in developing our methodology. Details of the sensitivity; study.. performed on this parameter and the results obtained are given below:

The number of axial intervals used in our methodology is 42 (Table 2 1 of the topical report). In a similar analysis, the Virginia Elcetric and Power Company (VEPCO), however, used 156 intervals-(Reference 2 of our topical report). Therefore, a sensitivity-study was performed to examine the effect of this parameter on the minimum departure of nucleate boiling ratio (MDNBR). In this study, the intervals used were 42, 60 and 156.

In order to perform this study, it was necessary to make a minor modification in the source code. The maximum problem size of the COBRA IIIC/MIT-2 computer code is limited to 80,000 words. This size was not sufficient to accommodate the dynamic storage requirements for the runs with larger intervals. Therefore, as recommended in the code users' manual (Reference 5, pages 63 and i

64), the dimensions of the DATA array given in the MAIN program and the-value of KMAX given in the CORE subroutine were increased from 80K to 240K.

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e o Dr. T. E. Murley AEP:NRC: 1081A The results of this study are depicted in the attached table. Two steady state and two transient DNB cases analyzed were taken from l Section 6 of the topical report.

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The first column of the table lists the plants / analyses studied. The second column shows_the j MDNBR values of these analyses available in the existing-licensing documents (Table 7 1 of topical report) and included for comparison purposes. The last three columns show the MDNBR values obtained for 42, 60 and 156 mesh spacings.

The results of this study reveal that, for steady state analyses, the value of MDNBR increases slightly with axial intervals. For  ;

transients this effect is reversed. It is noted, however, that the j MDNBR values obtained using 156 nodes are within +2.6% and -1.5% of '

the values for 42 nodes for the steady state and transient I analyses, respectively. This variation is minor, at;d the MDNBR ' f values for 42 axial nodes are in excellent agreement with those of l existing analyses (Column 2 of the table). The computer run time for the 156 node model, however, increased many-fold to achieve 4 this minor improvement in MDNER, ,

We would also like to bring to your attention that the author of i the COBRA IIIC/MIT-2 computer code found an axial mesh spacing 4 between 2 inches and 1 foot to be usually adequate for most j problems (Reference 5, page 64). This translates into 78 and 13 l inte rvals , respectively, for the fuel length of 156 inches for- the Cook Nuclear Plant.

It-is concluded from this sensitivity study that the nodalization.  ;

scheme _ of 42 axial intervals used in developing our methodology for I thermal-hydraulic analysis of the Cook Nuclear Plant is sufficient ]

enough to yield acceptable results.

This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and I completeness prior to signature by the undersigned.

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Sincerely, q

M. P. Alexich Vice President '

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Dr. T. E. Murley AEP:NRC:1081A-cc: D.11. Williams , Jr.

A. A. Blind G. Charnoff J. P. Padgett NFEM Section Chief A. B. Davis - Region III N. R. C. Resident Inspector

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e ATTACHMENT 1 TO AEP:NRC:1081A REQUEST FOR REVIEW OF AEP LEACTOR CORE THERMAL-HYDRAULIC ANALYSIS USING THE COBR.k IIIC/MIT-2 COMPUTER CODE

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MINIMUM DNBR Utility / Vendor Axial Intervals Axial Intervals Axial Intervals

'42 60 156 Steady State Analyses

1. Surry Plant 1.94-(VEPCO) 1.90 1.91 1.95

2. Cook Plant (Unit 1 Cycle 1) 1.97 (FSAR)' 1.93 1.94 1.98 Transient Analyses for Cook Plant (Unit 1. Cycle 1)

1. Excessive 1.oad increase 1.56 (FSAR) 1.57 1.56 1.55

2. Complete loss of Reactor Coolant Flow 1.38 1.36~

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1.40 (FSAR) 1.38