ML20214N328

Forwards Proprietary Westinghouse Repts Re Analyses of RCS Component Evaluations & Component Supports,Including Rev 2 to Seismic Reevaluation of Haddam Neck Plant RCS Component Supports. Encls Withheld

ML20214N328
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/26/1986
From:	Opeka J CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
Shared Package
ML19292F881	List: ML20214N328
References
A05457, A5457, GL-84-04, GL-84-4, NUDOCS 8609160244
Download: ML20214N328 (5)

Text

CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N. CONNECTICUT P o BOX 270 HARTFORD. CONNECTICUT 06141-0270 TELEPHONE 203-665-5000 August 26,1936 Docket No. 50-213 A05457 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U. S. Nuclear Regulatory Commission Washington, D. C. 20555 References (1) 3. F. Opeka letter to C. I. Grimes, dated April 11, 1986.

Gentlemen:

Haddam Neck Plant Reactor Coolant System Seismic Reevaluation in Reference (1) two commitments were made to provide additional information on the Haddam Neck Plant reactor coolant system seismic evaluation.

Summarizing the background briefly, Reference (1) provided a response to the ,

question of:

"How does the Haddam Neck plant propose to satisfy the two conditions identified in Generic Letter 34-04(l) with the Systemmatic Evaluation Program (SEP) seismic spectra, including a description of improvements planned to specific supports?"

The response to the first condition, which requires one to confirm that the results of the seismic analysis show that the maximum bending moments do not exceed 42,000 in-kips for the highest stressed vessel nozzle / pipe junction, confirmed this is the case provided no support failures occur. However, it was noted that the SEP seismic analysis, that was performed to provide input to the nozzle loads and confirm that WCAP-9553 was applicable, showed that certain code allowable loading values were exceeded for some reactor coolant system components. These specifics were revealed af ter confirmation from Westinghouse that the leak-before-break philosophy, as verified by WCAP-9558, (1) D. G. Eisenhut letter to All Operating PWR Licensees, Construction Permit Holders and Applicants for Construction Permits, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops" (Generic Letter 34-04), dated ()

February 1,1934.

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2-vas applicable to the Haddam Neck Plant. Recommended modificatjo turn identified in the NRC's Safety Evaluation of the SEP, Topic III-6t2)ns were in In Reference (1), Connecticut Yankee Atomic ' Power Company (CYAPCO) pointed out that due to recent developments in industry knowledge of seismic analysis methodologies, a review of the existing analysis methodologies and acceptance criteria and a reassessment of the need for performing these modifications was planned. A plan of action, including proposed analytical techniques for the evaluation was committed to be submitted. The response to the commitment is provided in Attachment 1.

The second and final commitment made in Reference (1) was to transmit Westinghouse reactor coolant system seismic evaluation reports that were done as part of the original SEP seismic evaluation. Attachment 11 provides a summary of analyses performed to date by Westinghouse on the Haddam Neck plant reactor coolant system. Included are data and results on the following items:

o Reactor coolant loop piping, o Reactor vessel, reactor vessel internals, and control rod drive mechanism, o Pressurizer and pressurizer surge line, o Steam generator, o Reactor coolant pump, o Reactor coolant system piping and component supports.

Although the above data provide for a thorough seismic structural qualification of the system, we feel that further refinement of the analyses, as described in Attachment 1, is in order. These refinements, based upon current methodologies, should more realistically define which modifications are required.

We trust the above information satisfies the commitments made. Should you have any further questions, feel free to contact us.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 3.F.\op6ka b0h U Senior Vice President (2) D. M. Crutchfield letter to W. G. Counsil, dated February 25,1983.

Attachn.ent i Haddam Neck Plant Proposed Analytical Techniques and Plan of Action for Reactor Coolant System Modification Reevaluation August,1936

X Commitment:

Provide a plan of action and include proposed analytical techniques to be used to re-evaluate the need to implement recommended modifications identified in NRC's Safety Evaluation of the SEP, Topic III-6.

Response

Presently, we have compared calculated loads to conservatively established code values and shown acceptable margins with a few exceptions. One of the most critical locations is where the bolts connect the steam generator to its support skirt. These loads exceed acceptance criteria established by the ASME Code.

Also, the pressurizer lateral truss loads exceed appropriate buckling acceptance criteria.

Based upon these results, our plans are to perform additional analytical work to avoid what we believe to be unnecessary costs for bolt replacement work and truss modification. Replacement costs for the steam generator support bolts alone are in the area of 1.2 million dollars. Both modifications would require significant occupational exposure. In addition, due .to the stringen; material requirements for these support bolts, larger diameter bolts would have to be installed. Damage to the steam generator support blocks from the cutting and tapping operation is possible, and if it were to occur, replacement costs would increase dramatically, as would overall radiation exposure on this portion of the job.

In order to provide more realistic loadings on both the steam generator hold-down bolts and the pressurizer lateral truss, our plans are to use the multiple support response spectra method, using the grouping technique which has been favorably reviewed in NUREG/CR3911 by Brookhaven National Laboratory and endorsed by the NRC Staff. For the present application, the multiple support response spectra rnethod will be used with the absolute sum combination between groups. For modal and spatial components, the square-root sum of the squares (SRSS) method will be used as recommended in NUREG 1061.

Since this work is being conducted based upon three earthquake input components, Code Case N411 piping damping values will be used where appropriate.

This additional work should be completed shortly, whereupon we will rep,gsess our need for the modifications. As committed in the April 11, 1936 letterm results of the re-evaluations and final plans for modifications will be reported to the NRC in January of 1987.

(1) The 3. F. Opeka letter to C.1. Grimes, dated April 11, 1986.

Attachment II Haddam Neck Plant Analysis Summary of the SEP Seismic Evaluation of Reactor Coolant System Components August,1986