ML20023C691

Discusses Implications of Indications on Upper Core Barrel Bolts Discovered During Ultrasonic Exams.Replacement of Seven Affected Bolts Not Necessary.Restart of Facility May Proceed W/O Undue Risk to Public

ML20023C691
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/11/1983
From:	John Marshall ARKANSAS POWER & LIGHT CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
1CAN058302, 1CAN58302, NUDOCS 8305170551
Download: ML20023C691 (4)

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ARKANSAS POWER & LIGHT COMPANY POST OFFICE box 551 LITTLE ROCK, ARKANSAS 72203 (501) 371-4000 May 11, 1983 1CAN058302 Director of Nuclear Reactor Regulation ATTN: Mr. J. F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Inspection of Upper Core Barrel Bolts Gentlemen:

As you are aware recent inspections of reactor internals bolting at Sacramento Municipal Utility District's Rancho Seco plant and Florida Power Corporation's Crystal River 3 plant have revealed unexpected indications on a number of bolts during ultrasonic (UT) examinations. The purpose of this letter is to discuss the implications of this issue relative to ANO-1 and to provide a current status of our efforts in response to this iss 2.

In addition this letter provides a discussion of the basis for AP&L's decision regarding the restart of ANO-1 following the current outage.

Problems with reactor internals bolting at 88W plants were first observed during visual inspections at Duke Power Company's Oconee 1 plant in July of 1981. During the inspection a number of lower thermal shield bolts were observed to be missing or broken. These lower thermal shield bolt failures were discussed in detail in a letter dated August 11, 1981, from Mr. J. H.

Taylor of B&W to Mr. Victor Stello of the NRC. As discussed in this letter, failure of the upper or lower thermal shield bolts was deemed not to constitute a safety hazard. As a result of these failures, and subsequent failures of lower thermal shield bolts observed during inspections at Oconee 2 and 3, the lower thermal shield bolts at AN0-1 were inspected and subsequently replaced during the fifth refueling outage which began in November of 1982.

Based on the information available at the time, including UT and laboratory examinations of bolts from the Oconee units, the failures were I felt to be restricted to the lower thermal shield bolts and UT inspections OgI g

of other internals bolting was felt to be unnecessary at ANO-1. However, subsequent to the reinstallation of the reactor internals and reactor 8305170551 830511 PDR ADOCK 05000313 G PDR MEME4E A MtOOLE SOUTH UTILITIES SYSTEM

dr. J. F. Stolz May 11, 1983 vessel head at ANO-1, but prior to restart, a nbmber of UT indications were observed in the upper core barrel bolts at Rancho Seco and later in the

>pper and lower core barrel bolts at Crystal River 3. Specifically, at Rancho Seco 19 of 120 upper core barrel bolts exhibited UT indications and at Crystal River 3, 51 of 120 upper core barrel bolts and 4 of 108 lower core barrel bolts showed UT indications. With the exception of the lower thermal shield bolts and surveillance specimen holder tube bolts, no other UT indications were observed in the remaining internals bolting inspections at these units. Surveillance specimen holder tubes are not installed at ANO-1. The configuration of the reactor internals for a B&W plant is shown in attachment 1 to this letter.

The results of the inspections at Rancho Seco and trystal River 3, prompted AP&L's decision to delay the restart of ANO-1 following its refueling outage in order to remov( the reactor vessel head and perform an inspection of the upper core barrel bolts. The UT inspection of the AN0-1 upper core barrel bolts was completed on May 3, 1983. Only 7 of the 120 bolts inspected showed UT indications. These indications were near the head to shank transition, consistent with those observed at Rancho Seco and Crystal River

3. The locations of the affected bolts appeared to be randomly distributed.

Based on the results of this inspection, evaluations conducted by B&W, and review by the Plant Safety Committee and Safety Review Committee, AP&L plans to proceed with the restart of AN0-1 and subsequent power operation without further inspections or repairs. The information discussed above was presented to the NRC staff in a meeting with the B&W Owners Group Task Force on May 6, 1983. During this meeting information was also presented relative to the reactor internals design basis, field inspection methodology, laboratory examinations of sample bolts, safety implications and future Owners Group plans to address this issue. The detailed information presented in the May 6,1983, meeting will be submitted in the form of a B&W Owners Group report within two weeks. Although this letter is not intended to duplicate the detail of the planned Owners Group submittal, following is a summary of the key items which form the basis for AP&L's decision to proceed with the restart of AN0-1, as presented to the NRC staff on May 6,1983.

An evaluation of the loads applied to the upper core barrel bolts indicates I that only 8 symmetrically located bolts are required to support the core i during normal operations. For faulted conditions (safe shutdown earthquake plus LOCA loads) 45 symmetrically located bolts are required. Since a total of 113 upper core barrel bolts were inspected with no UT indications, a very large margin presently exists in the load carrying capability of the upper core barrel joint. The observed failure rate at AN0-1 (7 failed bolts in 5.5 reactor years at temperature) would not result in a significant degradation of this margin during the next cycle of operation. Even if the failure rate during the next cycle of operation is conservatively postulated to be twice that observed at Crystal River 3 (51 failed bolts in 3.9 reactor years at temperature) a sufficient number of bolts would remain at the end of the cycle to support faulted conditions. Based on this margin AP&L does ,

not consider replacement of the 7 affected bolts to be necessary at this time. Complete failure of these 7 bolts would not result in a significant decrease in the load carrying capability of the joint and, since the bolt heads are held in place by welded clips, no loose parts would result from such failures. Replacement of these bolts is a complex operation possibly involving the removal of bolt shanks which have separated from the bolt

Mr. J. F. Stolz May 11, 1983 heads. Such operations would have required the fabrication of special tooling as well as replacement bolts and resulted in an extention of the current outage until the end of June 1983. Since the replacement of these bolts would not have significantly increased the available margin, such a delay was not warranted.

For the lower core barrel joint even fewer bolts (only 19 of 108) are required to support the core under faulted conditions. Based on the greater margin availiable, the favorable results of the upper core barrel bolts inspection at ANO-1, and the results of lower core barrel bolt inspections at other B&W plants, an inspection of the lower core barrel bolts is not warranted at this time. Such an inspection would require defueling of the reactor and removal of the reactor internals in order to allow access to the lower core barrel bolts.

As discussed above, due to the extremely large margins in the number of required bolts and the results of the inspection of the upper core barrel bolts, we feel there is adequate assurance that AN0-1 can be operated during the next cycle without failure of the core barrel joints. However, an evaluation of the consequences of such a failure has been conducted. As presented during the May 6 meeting, in the event of failure of a core barrel joint, the core will be supported by the core support lugs (12) welded to the reactor vessel wall. The evaluation concluded the support lugs are adequate for both dynamic and long term cyclic loading. The core drop would be limited to approximately 0.5 inches. Such movement would not result in control rod disengagement and, although core bypass flow could increase by approximately 5% in the case of an upper joint failure, core cooling would be maintained. In addition, the core guide blocks and the engagement of the bolt shanks would prevent any significant core tilt and control rod insertion would not be prevented.

Based on the considerations discussed above we conclude that restart of AN0-1 may proceed without undue risk to the health and safety of the public. Although this issue does not constitute an immediate safety concern, AP&L plans to actively pursue further evaluation of this issue and to participate in the ongoing efforts of the B&W Owners Group Task Force. In addition, as a precautionary measure, AP&L is proceeding with the installation of neutron noise analysis equipment for use during the next cycle of operation. This equipment will provide the capability to detect any significant movement of the core barrel during operation.

The need for and extent of future inspections and/or repairs will be based on the results of the Task Force investigations and upcoming inspections at other facilities. We will continue to keep the NRC staff informed of developments on this issue.

Very truly yours, John R. Marshall Manager Licensing JRM/DRH/rd

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