ML19325E160

Forwards Addl Info Re Five Section D Items in SER Accepting Steam Generator Tube Rupture Analysis Methodology Documented in WCAP-10698,per NRC Request.Revised FSAR Table 15.6.3-1 Will Be Included in Future FSAR Update

ML19325E160
Person / Time
Site:	Catawba
Issue date:	10/23/1989
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8911020126
Download: ML19325E160 (3)

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(U.ES? Nuclear' Regulatory Commission

Attention: Document Control Desk ll

-Washington, D. C.-'20555 .i e

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i , T. f Subject : Catawba Nuclear Station, Units 1 and 2 'l

" Docket Nos.'50-413 and 50-414 3

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, Steam Generator Tube Rupture Analysis 'l e

t Gentlemen 1  :

On'Harch:30, 1987 the NRC staff issued a Safety' Evaluation Report (SER)L ,

accepting the Steam Generator Tube Rupture (SGTR) analysis methodology i q- . documented in WCAP-10698,;SGTR Analysis Methodology to Determine the Margin to l

". Steam Generator. Overfill.'Section D, Enclosure 1, of the NRC's SER required additional plant specific input from each utility referencing WCAP-10698. My ~i December 7,'1987 letter to the NRC addressed the five iters. required by SER ,

.Section D,.for Catawba Nuclear Station. Please find attached additional j information regarding the five Section D items for Catawba.'This information j

'is being. requested per conversations with the NRC' staff. ,;

L My August 24,',1988 letter transmitted SGTR analysis and a Technical i <~' Specification amendment request in response to License Conditions 16 (Unit 1)  :

Jand.10 (Unit.2). Revised FSAR Table 15.6.3-1 was attached to my August 24, l 1988 letter. Two typographical errors have been found in the revised FSAR 1 Table 15.6.3-1 that relate to the SGTR sequence of events. Cooldown completion .

u should have been at 64.2 minutes instead.of 64.8 minutes. Depressurization l

should'have been at 66.2 minutes instead of 66.3 minutes. As indicated,in the attachment, simulator training response times are consistent with the FSAR -

Y analysis. The revised FSAR Table 15.6.3-1 will be included in a future FSAR

,, update.

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H. B Tucker 1, '  !

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JGT/5/SGTR 1 '

y xc: Mr. S. D. Ebneter l Regional Administrator, Region II l> U. S. Nuclear Regulatory Commission 101 Marietta St., NW, Suite 2900 0 <

Atlanta, Georgia 30323 Mr. W. T. Orders N

.NRC Resident inspector j L

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CATAWBA NUCLEAR STATION.

STEAM GENERATOR TUBE RUPTURE ANALYSIS

. UPDATE TO DECEMBER 7, 1987 SUBMITTAL ITEM 1 TRAINING PROGRAMS Additional information to that in Item 1 of my December 7, 1987 submittal was provided to'the NRC per my August 8, 1988' letter. Operators have been retrained'in all five Action Items in the August 8, 1988 submittal. Simulator training response times are consistent with safety analysis assumptions.

ITEM 2 COMPUTER CODE On October 19, 1988 the NRC approved the computer code RETRAN02/ MOD 004. This is the computer code used for the plant specific dose analysis in Item 2. J Additional information to that in Item 2 of my Dtcember 7,.1987 submittal was

provided per my August 24, 1988 letter to the NRC .

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ITEM 2 INITIAL CONDITIONS In the SER issued on March 30, 1987 accepting the SGTR analysis methodology documented in WCAP-10698, the NRC staff concluded that "the single failure analysis and sensitivity studies in (WCAP-10698) Supplement 1, have identified the worst'aingle f ailure and the analysis assuiaptions which are conse vative with respect to offsite doses." The SER also indicates that this conclusion  !

will stand unless the review of WCAP-106989 causes it to be challenged or some reason is'found on a plant specific basis for it not being applicable. No such reason was found for Latawba Nuclear Station. The worst single failure identif.ied in WCAP-10698 Supplement 1 as a failed-open steam line PORV on the ruptured steam generator. This worst single failure was used for the Catawba-Nuclear Station offsite dose analysiv.

Break location is not a sensitive parameter for the offsite dose analysis unicas the steam generator tubes are uncovered. No credit is taken for '

partitioning in the secondary liquid if the steam generator tubes are uncovered. This is equivalent to assuming a break at the top of the tube bundle.

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ITEM 3 STRUCTURAL ANALYSIS OF MAIN STEAM LINES The Catawba Nuclear Station main steam piping was qualified for structural adequacy including the postulation of a steam generator tube rupture (SGTR).

I The SGTR was considered in the analysis by including a static loading case

l. with each main steam isolation valve. All supports - with the exception of l' snubbers - were assumed to be active during this loading case. Spring supports were assumed to be un-pinned (travel stops removed) as during normal operating L conditions.

Qualification of the piping was performed using the above loading case in l equation 9 of the ASME code, 1974 edition through summer 74 addenda. The SGTR case is considered a faulted loading with respect to support loads. This particular case is enveloped by other faulted loadings and does not control

support design.

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• CATAWBA NUCLEAR STATION ]

6 STEAM. GENERATOR TUBE RUPTURE ANALYSIS

? ' UPDATE T0' DECEMBER 7,1987 UUBMITTAL l

1 TEM 5 RELAT10NSilIP TO REFERENCE PLA(

As stated'in the December 7, 1987 response,the Catawba steam generators have ll smaller tube' diameters and larger initial steam volumes than,the reference

'. plant. As stated on page 4-3 of WCAP-10698, the reference' plant inside; tube i diameter is 0.775", while the Catawba Nuclear Station inside. tubo diameter is 0.664" (see Catawba FSAR. Table 5.4.2-1).-As indicated in WCAP-10698 Table 4.1-1 the' reference plant steam volume is 3658 cubic it. The Catawba Nuclear

' Station Unit 1 steam volume is 3848 cubic ft. and the Unit 2 steam volume is 3883 cubic it. Following the December 7,1987 submittal, Catawba Unit 2 has

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lowered the full power normal operating water level in the steam generators.

This Catawba Unit 2 modification results in an even larger margin to overfill.

The break location was modeled conservatively. For the system  ;

f' thermal-hydraulic analysis, the break was assumed to occur at the top of the '

tubesheet on the outlet plenum side. This assumption is consistent with the guidance in WCAP-10698, page 3-6.

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' ITEM 5 SINGLE FAILURE

-As stated in the December 7, 1987 submittal, the single failure evaluation in WCAP-10698 either is applicable to or bounds the Catawba Nuclear Station singic failure, depending on the specific failure. Because certain failures are bounding on Catawba, rather than being directly applicabic to it, the numerica1'results in WCAP-10698 for relative severity of individual singlo failures would not be directly applicable to Catawba. It can only be determined that the worst single failure at Catawba would be no worse than the +

numerical results given in the WCAP. A plant specific analysis would be  :

required to further determine which failure is worst and to quantify its ,

severity. A plant specific analysis is contrary to the purpose of the owner's l' group approach of bounding all possible plants via a generic analysis.

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