ML20024A229
| ML20024A229 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/01/1983 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20024A226 | List: |
| References | |
| NUDOCS 8306160164 | |
| Download: ML20024A229 (4) | |
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UNITED STATES' e
!?)m)'..j NUCLEAR REGULATORY COMMISSION 1
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WASHINGTO N, D. C. 20555 SAFETY EVALUATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION
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DOCKET NO. 50-285 The Fort Calhoun Station was returned to service on April 7,1983 following a refueling outage which included the.10-Year Inservice Inspection of the reactor vessel and internal components.
Part of this inservice inspection was a visual examination of the accessible portions of the core support barrel and thermal shield.
The ISI plan included a visual inspection of all accessible sections of the core support barrel per ASME,Section XI, Table IWB-2500, Paragraph 8-N-3, an inspection of points of attachment (core support barrel flange and snubbers) of the core support barrel to the reactor vessel, and an i spection of the core support barrel and thermal shield to determine the general condition alter nearly 10 years of service.
This inspection' was performed by the District's ISI contractor, Southwest Research Institute, with assistance provided by OPPD engineers and QC technicians.
As discussed in the following paragraphs, this inspection clearly demon-strated that the reactor vessel, the core support barrel, and the thermal shield are in excellent condition.
There is no evidence which indicates that the type of problem which occurred at St. Lucie 1 would occur at Fcrt Calhoun.
Additionally, the reactor vessel internals at Fort Calhoun were subjected to an extensive pre-critical vibration monitoring program.
No abnormal or unanticipated motions were detected during this program.
Tne operating history of the reactor coolant system for the.past ten pars has been excellent.
No abnormalities have ever been detected carir.; refueling cperations.
No loose parts frc, the primary system have ever been noted in either the reactor vessel or the steam gaaerators, i
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A preliminary evaluation was performed by Combustion Engineering soon ll after evidence became available that the thermal shield at St. Lucie Unit 1 was in a degraded condition.
The intent of the evaluation was to make an early determination of the safety significance of the inci-dent.
C-E's findings indicated that there was no loss of any safety function (to the extent that a major reduction existed in the ' degree of protection provided to the health and safety of the public) and, tnerefore, no substantial safety hazard was found to exist.
These conclusions were based on the following considerations:
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2 Damage to the thermal shield and supp;rt system appeared to be extensive, however, there was no evidence indicating that the St. Lucie-1 thermal shield was in imminent danger of failing.
Even if the thermal shield had dropped fro,m the core thermal shield support. lugs, a preliminary assessment indicated that the core support barrel snubbers would have prevented further downward movement of the shield.
The. thermal shield, in this position, would not have had a significant impact on core flow patterns.
Should it be postulated that the thermal shield falls beyond the core support barrel snubbers, there exists an additional support location, the core stops.
The core stops 'were designed to hold the core support barrel including thermal shield, the reactor core and vessel internals in the event of a failure of the core support barrel.
Examination of the St. Lucie thermal shield revealed both large and small pieces missing from the shield.
Preliminary investi-gations indicate that the largest pieces would not pass through the flow skirt (flow baffles) and therefore, could be expected to reside outside this component.
Pieces small enough to enter the flow skirt orifices would
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either settle inside the flow skirt or may be propelled upward by coolant flow.
Pieces small. enough to enter the reactor core might cause limited fuel damage due to fretting.
Technical specifications en adionuclide activity are, however. sufficient to allcw acequate monitoring and protection as a result of any possible fuel damage.
It is not expected that the missing pieces of thermal shield, regardless of their final location, would prevent the core from being maintained in a coolable array.
This preliminary assessment of safety significance also applies to the thermal shield at the Fort Calhoun Station.
There is no evidence, at this time, which indicates that there is a significant probabi.lity of thermal shield support failures at the Fort Calhoun Station in either the near-term or the long-term.
The Licensee has initiated a plant-specific analysis of the effects of the complete loss of the thermal shield supports and a drop of the shield.
If the rasults of this analysis differ significantly from the preliminary assessment discussed above, the results will be forwarded to the Commission by June 15, 1983.
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Tne Operating histcry of tne Fort aincun Station and the resulis of the recent visu,al inspection of the reactor vessel internals justify continued operation of tne station.
This is supported further by the preliminary assessment of the unlikely event of thermal shield support failure.
Due to the concerns expressed by CE ADP Info Bulletin 82-12, regarding the thermal shield positioning pin situation at Maine Yankee, emphasis was placed on the inspection of the accessible portions of the thermal shield positioning pins, the locking collars and the lock welds.
More than half of these positioning pins would have been examined in the course of the originally pla.nned inspection, but it was decided to devote extra effort in this area to examine all of the pins where possible.
The thermal shield positioning pins at Fort Calhoun are 4-7/16" in length and 2-1/4" nominal diameter in the thread region.
The thread pitch is 12 per inch.
The pins were installed through the thermal shield and bear against hard-faced surfaces on the exterior of the core support barrel.
The pins were torqued to 250 foot pounds; a threaded locking collar was then installed around each pin and torqued to.50 foot pounds.
Following this operation, the locking collars were fillet welded to both the pins and the thermal shield over their entire circumference.
The core suppo.rt barrel and upper guide structures were stored in a lower portion of the refueling cavity during the reactor vessel ISI.
Although close clearances between the cavity structures and the core-support barrel prevented a thorough inspection of all surfaces, snubbers, positioning pins, etc.; sufficient inspections were per-formed to determine tnat the core support barrel and thermal shield are in excellent condition.
The detailed results of the Fort Calhoun thermal shield and core support barrel visual inspection are as follows:
Tne neacs of all up;er and lower pcsitiening pins wer.e inspec ec anc all pins, locking collars and lock welds were verified to be intact.
No evidence of motion or cracking was noted.
Three of the six snubber assemblies on the core support barrel were
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visually inspected.
No evidence of wear or abnormal motion between y'
the core support barrel and the reactor vessel was noted.
All six l'
of the reactor vessel portions of the snubbers were examined with no
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deterioration found.
Four of the eight shield support lugs were examined with no evidence of deterioration noted.
The inspection of the lugs, support pins, fillet welds, and shims showed no abnormalities.
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_4 More than 150 of the core support barrel upper to lower shell weld was examined,with no indications found.
More than 180 of the core support barrel to flange weld was examined.
l Again, no indications were noted.
1 One outlet (hot leg) nozzle was examined.
There were no indications of wear between this nozzle and the corresponding nozzle in the reactor vessel.
The accessible portion of the points of attachment of the core support barrel flange to the reactor. vessel were examined.
Again, no indi-cations of abnormal motion were found.
The visual inspection of the reactor vessel showed no indica'tions of abnormal motion of the core support barrel at the flange support lugs or at the outlet (hot leg) nozzles.
As a result of this visual inspection of the core support barrel and thermal shield, the ISI contractor concluded that these reactor vessel components are in excellent condition.
There is no evidence which indicates that the types of problems which have occurred at Maine Yankee and St. Lucie I would occur at Fort Calhoun.
The Fort Calhoun reactor vessel has been reassembled, refueled and returned to service.
i The next inspection of these components is scheduled at 20 calendar
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years of service, in 1993.
Based on a review of the above information, the NRC staff conclud s tnat continued operation of Fort Calhoun is justified because of the detailed knowledge of the conditions of the thermal shield support system and internals gained in the recent 10 year inservice inspection, conducted with the internals removed; on the knowledge that core flow cassation wculd r.ct occur even if the thermal shield were to drop cecause motion would be arrested by either the barrel snubbers or core stops; and on the evidence that thermal shield support system distress might start early but progresses gradually over a period of years rather than suddenly or as a result of a single plant event, thus allowing periodic assessment.
The condition of both the Vibration and Loose Parts Monitoring System and the Excore System are such that little or no dependence should be placed upon them for monitoring the thermal shield.
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