A05457, Forwards Results of RCS Seismic Reevaluations & Final Plans for Mods Identified in NRC Safety Evaluation of SEP Topic III-6.All Outstanding Issues Re Generic Ltr 84-04 & SEP Topic III-6 Resolved

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Forwards Results of RCS Seismic Reevaluations & Final Plans for Mods Identified in NRC Safety Evaluation of SEP Topic III-6.All Outstanding Issues Re Generic Ltr 84-04 & SEP Topic III-6 Resolved
ML20215J689
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 05/01/1987
From: Mroczka E, Werner R
CONNECTICUT YANKEE ATOMIC POWER CO.
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TASK-03-06, TASK-3-6, TASK-RR A05457, A5457, GL-84-04, GL-84-4, NUDOCS 8705080203
Download: ML20215J689 (4)


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O YANKEE ATOMIC POWER COMPANY C CONNECTICUT tilfrHONE DEnLIN. CONNECTICUT P O T404 270. HAntFORD CONNECTICUf 06t414270 m wum May 1,1987 Docket No. 50-21)

A0$t437 Ite: 10CF1150, Appendix A U.S. Nuclear Itegulatory Comrnission A t tn Docurnent Control Desk Washington, D.C. 20555 Iteferences (1) J. F. Opeka letter to C. I. Grirnes, Ileactor Coolant Systern Selsrnic Iteevaluation, dated August 25,1936.

Gentfernent fladdain Neck Plant Ileactor Coolant Systern Sciunic lleevaluation and Final Plans for Modifications In Iteference (1), two corninitinents(l) to provide additional information on the lladdain Neck Plant reactor coolant system scistnic evaluation were discussed.

The first committnent was to provide a plan of action to perform a review of existing solunic analysis inethodologies and acceptance criteria in order to determine the need for perforn ng modifications identitled in the NitC's Sa'ety Evaluation of SEP Topic 111-6.(2 This committnent was closed out. The second commitment, to provide the results of the re-evaluations and final plans for modifications to the NitC, was reiterated.

Attachtnent i provides our response. Upon conc irrence with this information and cornpletion of the described inodifications, Connet.ticut Yankee Atomic Power Company (CYgCO) concludes that all outstanding issues with respect to Generic Letter 34-04 and SEP Topic !!!-6, as it pertains to the reactor coolant system, are resolved.

(1) These committnents were made in a J. F. Opeka letter to C. I. Grimes, dated April 11,1984.

(2) See the D. M. Crutchfield letter to W. G. Counsil, dated February 25,1983.

()) D. G. Eisenhut letter to All Operating PWil Licensecs, Construction Perrnit iloiders and Applicants for Construction Permits, " Safety Evaluation of Westinghouse Toplcal lleports Dealing with Ellinination of Postulated Pipe lireaks in PWII Primary Main Loops" (Generic Letter 34-04), dated {

February 1,1934.

I 0705000203 070001 ADOCK 03000213 PDH p PDH l ()\

6 U.S. Nuclear Itegulatory Commission A05457/Page 2 May 1,1987 Should you have any questions with respect to the information provided, feel frec ,

to contact our Licensing representative directly.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY

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E. 3. Mru@.ka F Senior Vice President  !

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y: 11. P. Werner i Vice President l

cci W. T. Itussell, Region 1 Administrator P. M. Akstulewicz, NitC Project Manager,lladdam Neck Plant P. D. Swetland, Resident inspector, lladdam Neck Plant T

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Docket No. 50-213 i A05457 i

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Attachment 1 Iladdam Neck Plant l

Itcactor Coolant System Structural Modifications, Itesults of Analyses and Plans for Mod!!! cations 1

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May,1987  ;

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Reactor Coolant System Structural Modifications,

, Results of Analyses and Plans for Modifications As discussed in Reference (!), CYAPCO planned to perform additional analytical work to better define leadings on the steam generator hold-down bolts and pressurizer truss. These evaluations have been completed. The results conclude that modifications are still required on the pressurizer truss. However, modifications previously specified for the steam generator hold-down bolts are not required; safety factors on the bolts are now within allowable limits.

The analytical techniques utilized in arriving at the aforementioned results consisted solely of performing the square root sum of the squares method, as recommended in NUREG 1061, for combining modal and spatial components at critical stress locations. Multiple support response spectra and Code Case N-411, as identitled in Reference (1), were not employed.

Based on these final results, CYAPCO plans are to perform the following work during the 1937 refueling outage l o The pressurizer truss structure will be reinforced to bring member i stress within allowable ASME Section 111, Subsection NF limits.

( o A spring can (RC-H-17) on the pressurizer surge line will be replaced l with a rigid sway strut! this will bring member stresses on that l support within allowable ASME Section Ill, Subsection NF limits and keep the loading on nearby spring can RC-H-18 within allowable  ;

limits.

o Readings taken in 1936 on spring cans for two of the reactor coolant l pumps (P-171 and P-17-3) Indicate there is adequate travel on each l of the six spring cans to sustain movement during a scismic event.  ;

Therefore, reinforcement of the hangers will be unnecessary.

Iteadings on the hangers of the other two pumps (P-17-2 and P-17-4) l will be taken early in the 1937 refueling outage and calculations performed to determine their potential for " bottoming-out" during a selsmic event. Should adequate travel be available, no modifications will be performed and this issue will be closed in the event too little travel is available, our plans are to readjust the spring cans, rather than reinforce the hangers.

l Iteferences (1) 3. F. Opeka letter to C.1. Grimes, dated August 25,1936.

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