ML19345A525

Approves Proposed Change 83 to Tech Specs,Authorizing Replacement of One 5-1/4 Inch Diameter Reactor Vessel Studs

ML19345A525
Person / Time
Site:	Yankee Rowe
Issue date:	04/26/1968
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Minnick L YANKEE ATOMIC ELECTRIC CO.
References
NUDOCS 8011240116
Download: ML19345A525 (3)

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j APR 2 61968 In Reply Refer To:

Docket No. 50-29 Yankee Atomic glectric Company 441 Stuart Street Bostoo, Massachusetts 02116 Attention:

Mr. L. 3. Minnick Change No. 84 Vice President License No. DFg-3 Gentlemen p

By letter dated April 17, 1968, and supplemental telegram dated April 25, 1968, Yankee Atomic 31ectric Company requested approval of Proposed Change No, 84 to Technical Specifications of Facility License No. DFg-3 for the Yankee lhaclear Power Station. The proposed

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change would authorise replacement of one of the 52 existing 51/4 I

inch diameter reactor vessel stude, made out of ' ~'-SA-193, Grade 316 stsel, with a stud having a diameter of 4 3/4 inches, made est of ASTM-A-320 Grade L43 steel. A threaded sleeve will be inserted in the flange to accept the 4 3/4 inch diameter stud.

Char calculations indicate that the proposed 4 3/4 inch stud has sufficient strength to satisfy the ASTM Code requirements conseralag bolt-load for the reactor operating conditions without exceeding the allevable stress intensities. The Young's modulus of elastisity varies only slightly between different steels. As long as all reactor vessel studs respond to the same tensioning force, the proposed change does not alter the pressat design or its effectiveness. The licenses has stated that the single stud with the smaller diameter l

will be tensioned to the same load as is employed on the remaining l

51 stede. Ma agree with the licensee's conclusion that the dimen-sions of the Laserted sleeve will cat appreciably weaken the ligament i

between the adjacent stud holes.

Our fatises analysis of the proposed 4 3/4 inch stud indicates a reduction la the number of alloweble stress cycles; however, titt 8011240/lb

2-reduction does not alter its effectiveness over the life of the reactor. The Technical Specifications require that the licensee inspect all the studs of the reactor vessel head at every reactor refueling to verify the integrity.

Consequently, we have concluded that the proposed change does not present significant hasards considerations not described or implicit in the hazards summary report, and that there is reasonable assurance that the health and safety of the public will not be endangered by operation of the facility with the proposed change.

In view of the foregoing and pursuant to Section 50.59, 10 CFR Part 50, Change No. 84 is hereby authorised as proposed.

Sincerely yours, Peter A. Morris, Director Division of Reactor Licensing JI3i.Ub CTI ON:

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%rr W April 26, 1968 In Reply Refer To:

Docket No. 50-29 Yankee Atomic Electric Company 441 Stuart Street Boston, Massachusette 02116 Attention:

Mr. L. E. Minnick Change No. 84 Vice President License No. DPR-3 Gentlemen:

By letter dated April 17, 1968, and supplemental telegram dated April 25, 1968, Yankee Atomic Electric Company requested approval of Proposed Change No. 84 to Technical Specifications of Facility License No. DPR-3 for the Yankee Nuclear Power Station.

The proposed change would authorize replacement of one of the 52 existing 5 1/4 inch diameter reactor vessel studs, made out of SA-193, Grade B16 steel, with a stud having a diameter of 4 3/4 inches, made out of ASTM-A-320 Grade L43 steel. A threaded sleeve will be inserted in the flange to accept the 4 3/4 inch diameter stud.

Our calculations indicate that the proposed 4 3/4 inch stud has sufficient strength to satisfy the ASTM Code requirements concerning bolt-load for the reactor operating conditions without exceeding the allowable stress intensities.

The Young's modulus of elasticity varies only slightly between different steels. As long as all reactor vessel studs respond to the same tensioning force, the proposed change does not alter the present design or its effectiveness.

The licensee has sr.ated that the single stud with the smaller diameter will be tensioned to the same load as is employed on the remaining 51 studs. We agree with the licensee's conclusion that the dimen-l sions of the inserted sleeve will not appreciebly weaken the ligament between the adjacent stud-holes.

Our fatigue analysis of the propoaed 4 3/4 inch stud indicates a j

reduction in the number of allowable stress cycles; however, this l

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1 reduction does not alter its effectiveness over the life of the reactor. The Technical Specifications require that the licensee inspe:t all the studs of the reactor vessel head at every reactor refueling to verify the integrity.

Consequently, we have concluded that the proposed change does not present significant hazards considerations not described or implicit in the hazards summary report, and that there is reasonable assurance that the health and safety of the public will not be endangered by operation of the faci).ity with the proposed change.

In view of the foregoing and pursuant to Section 50.59, 10 CFR Part 50, Change No. 84 is hereby authorized as proposed.

Sincerely yours, Peter A.

rris, Director Division of Reactor Licensing i

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