ML20151U939

Forwards Upper Drywell Analysis to Support Util Position That Change to Tech Spec 5.2A Not Required.Drywell Proven Structurally Adequate by ASME Code Methodology W/Alowable Stress Derived from CMTR

ML20151U939
Person / Time
Site:	Oyster Creek
Issue date:	04/25/1988
From:	Wilson R GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20151U944	List: ML20151U939 ML20151U990
References
5000-88-1544, NUDOCS 8805020214
Download: ML20151U939 (2)

Text

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,, r GPU Nuclear Corporatica s

3 gg One Upper Pond Road Parsippany, New Jersey 07054 201-316-7000 TELEX 136 482 Writer's D; rect Dial Number:

April 25, 1988 5000-88-1544 U.S. Nuclear Regulatory Comission Attention : Document Control Desk Washington, D.C. 20555 Gentlemen :

Subject : Oyster Creek Nuclear Generating Station Docket No. 50-219 License No. DPR-16 Oyster Creek Drywell Containment On November 13, 1987, the GPU Nuclear Staff met with NRR and Region 1 representatives to review the data and assessments related to UT measurements at elevations 50'-2" and 87'-5" of the Oyster Creek Drywell . These measurements were initiated by GPUN during an outage of opportunity to confirm the condition of the drywell above the sand entrenchm?nt region. i To support the Oyster Creek restart, GPUN prepared an analysis that assumed unifom thinning to the upper regions of the drywell over the entire surface of the drywell vessel . UT measurements made on the drywell had demonstrated that this was not the case, and the assumption of unifom thinning was conservative. Using this assumption, GPUN had concluded that the drywell was structurally adequate based on the analytical approach described in Attachment 5 to the safety evaluation (SE No. 000243-002). This analysis was identical in method to earlier drywell evaluations which were reported to NRC except that allowable stress used in the upper cylinder analysis was derived from the lowest value af certified mill test report (CMTR) prcperties for the plate material actNily used in constructing portions of the drywell.

The purpose of the safety evaluation was to address the structural adequacy for a vessel inservice condition. Even though the t.taff agreed that it would be safe to restart and operate Oyster Creek, GPUN agreed to explore alternate analysis methods to demonstrate ASME code compliance for a new vessel. GPUN contracted CBal for this effort. Briefly, the CB&I ef fort was an axisymmetric analysis of a cylinder with a uniform thickness of 0.59 inches, and subject to an internal pressure of 62 psig. The cylindrical model included the effects of the stiffeners which are located at three elevations (80'-6-1/4",

84'-11-3/4", and 88'-8-1/2") on the Oyster Creek Drywell as well as the stiffening effect of the knuckle transition below elevation 71 ' 3/4" .

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P DCD GPU Nuclear Corporabon is a subsidiary of General Pubhc Utit,es Corporation

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s The results from the CB&1 code analysis have concluded that three of four regions adjacent to the stiffeners meet the code criteria. However, a 40" reginn of the cylinder model near elevation 75'0" exceeds the code allowable values for general primary membrane stress by approximately 8.5%. Since one area of this hypothetical model does not meet the code, we cannot conclude that full cade compliance for a design condition is demonstrated by this model.

This model has inherent conservatisms which must be considered in the overall drywell evaluation. First, the model does not duplicate the actual local areas of degradation. For example, a single thickness (0.59) may facilitate the analysis, but may overpredict the true stress state. Second, the ASME Code allowable stresses are intended for new construction where an underrun of 0.010" below specified metal thickness is permitted. This underrun which is equivalent to a 1.6% overstress is not factored into the analysis. Third, the ASME Code allowable stresses are based on specifled minimum mechanical properties versus actual material properties. Fourth, the Code includes a factor of safety over the design stresses which is calculated by dividing the minimum tensile strength of the material by the specified allowable stess, which numerically is 3.64. By using the calculated general primary membrane stress, the factor of safety for this model is approximately 3.37.

Based on these facts and good engineering judgement, we are concluding that the drywell was proven structurally adequate by ASME code methodology with the allowable stress derived from the CMTR's. Since we are analyzing an inservice condition and not a design condition, we do not believe a change to Technical Specification 5.2A is required.

l The CB&I report is enclosed for your inspection. If you have any questions or comments on this subject, Please contact Mr. M.W. Laggart (201) 316-7968.

y t ,uly yours, (Q

.F. Wil on Vice President Technical Functions RFW/DJ/jbw 6494f l cc: Mr. William T. Russell, Administrator l Region I i

U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 NRC Resident Inspector Oyster Creek Nuclear Generating St? tion Forked River, NJ 08731 l

Mr. Alex Dromerick, Jr.

l U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l

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