ML20205G200

Safety Evaluation Supporting Results of near-term Insp of Thermal Shield & Thermal Shield Support Sys,Contained in Util 860425 Final Core Support Barrel Insp Rept (post-Cycle 6)

ML20205G200
Person / Time
Site:	Saint Lucie
Issue date:	08/08/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205G181	List: ML20205G175 ML20205G200
References
NUDOCS 8608190427
Download: ML20205G200 (5)

Text

%, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION t y WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FLORIDA POWER AND LIGHT COMPANY .

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ST. LUCIE UNIT NO. 1 00CKET NO. 50-335

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THERMAL SHIELD REC 0VERY PROGRAM Introduction During a refueling outage which commenced in February 1983, the thermal shield and the thermal shield support system at St. Lucie 1 were found to be damaged. The damage consisted of cracking of the thermal shield to the extent that some pieces had broken off and some positioning pins and positioning lockbars were lost. Associated with the thermal shield damage was cracking of the core support barrel _ (CSB) in the . area of the thermal shield support lugs. The extent of the cracking varied from very exten-sive at several lug areas to no cracking at all in one lug area. Subse-quent evaluation led to a decision to remove the thermal shield.

The staff published a Safety Evaluation dated March 14, 1984 that evalu-ated the licensee's report " Final Ir.tegrity and Stability of Internals -

Conclusions and Findings" (Reference 1). Because.of the uncertainty whether the visual inspection has located all of the cracks in the core support barrel, and also to verify that additional degradation has not occurred, the staff recommended in the Safety Evaluation that an l inspection of the core support barrel be performed at the next refueling i outage and that future inspections be scheduled based upon the results of

this inspection. Furthermore, the inspection system should incorporate any advances that are made in the state-of-the-art to the maximum extent possible.

In a letter dated April 25, 1986, the licensee, Florida Power and Light Company (FP&L), submitted a report entitled, " Final Core Support Barrel Inspection Report Post Cycle Six," (Reference 2) describing the inspec-tion of the CSB repaired areas. The staff evaluation of the licensee's inspection results is presented below.

Description of Repairs In order to define the objectives of the post Cycle 6 inspection, the post Cycle 5 repairs will be described. The repair methods utilized extensive machining to remove part of the throughwall cracks and all of the non-throughwall cracking. Those throughwall cracks remaining were stop drilled at the termination points with various sized crack arrester holes which were closed with expandable plugs of 3", 5" or 8" diameter.

At five (5) of the lug areas, cracking was so extensive that a patch fastened with expandable plugs was used to cover the machined area. The 8608190427 DR 860808 ADOCK 05000335 PDR -

, purpose of both the plugs and patches was to prevent bypass coolant flow through the core support barrel. The inspection methods used to evaluate the damage consisted of: (1) visual (underwater TV), (2) eddy current testing, and (3) ultrasonic testing. [

Summary of Post Cycle 6 Inspection Section 10, " Monitoring and Inspection Programs" (Reference 1), describes the licensee's short-term and long-term commitments with regard to the monitoring and inspection of the reactor coolant system. The licensee submitted Reference 2 to document the short-term inspection described in paragraph 10.5.1 of Reference 1. During the refueling outage following Cycle 6, the licensee performed a visual examination of the repaired areas to look for evidence of looseness, motion, or wear on the plugs and patches as well as indications of new or continued crack growth in the base metal. The residual flange deflection of each of the installed plugs was checked by using the same remote inspection technique used after the initial plug installation; the status was documented and compared to pre-determined screening criteria.

The visual inspection of the CSB was performed with a remotely controlled underwater TV camera system. Eddy current examination techniques were l available as backup to evaluate any indications which were identified and considered unresolvable by visual means. The examination included both video scanning for the actual inspection of the CSB and still photography of the repaired areas to provide a permanent visual record. All repaired themal shield support lug areas on the CSB were examined. The minimum area of examination extended 3 inches beyond each support lug repair.

l 1 All previous crack locations were examined for a minimum of 3 inches on l either side. Each plug and patch was visually inspected for looseness as would be indicated by wear at the flange, gaps indicating loss of preload, or changes in plug orientation (identified by a match mark and the identification number stamped in the plug flange).

The licensee's results and conclusions are sumarized as follows:

A. All visual inspections were recorded and detailed by qualified inspectors on data sheets showing the configurations of each lug area inspected. The data sheets and video recording were then evaluated by the licensee for final acceptability. The results showed no crack extension and no evidence of plug or patch move-ment. The CSB condition was found to be unchanged from the post Cycle 5 baseline visual inspection of the repaired areas. There-fore, the licensee concluded that the core support barrel is acceptable for continued service based on this visual examination.

B. All flange deflection measurements taken during the post Cyle 6 inspection were above the screening criteria for maintaining the plug preload to the end of the plant's design life. The licensee T

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V also concluded that the plugs have sufficient preload in their present installed condition to continue to provide their design function and are acceptable for continued service. ..

t Safety Evaluation

• The staff has completed the review of the licensee's report entitled

" Final Core Support Barrel Inspection Report Post Cycle Six," submitted in a letter dated April 25, 1986. The licensee used a state-of-the-art underwater TV system to perform the visual examination of the repaired areas. The inspection tool is nomally used for reactor vessel examina-tions and thus provided a stable working platfonn with a remote posi-tioner. Oblique and variable intensity light sources were used to enhance the visual examination capabilities. The visual system was qualified and suspected indications of at least 5 mils in width can be resolved with the TV camera. Still photographs were taken with a con-ventional 35 mm camera mounted in a dry box.

The plugs and patches were intended to remain in place in the CSB for the design life of the plant. A fundamental design objective for the installed plugs was that they remain tight in the CSB during plant operation. The licensee considered loss of plug preload due to the following mechanisms: (1) stress relaxation from irradiation, (2) thennal cycling and (3) vibration. The results of the licensee's mechanical inspections indicate that all plugs have retained sufficient preload to remain tight.

The staff agrees with the licensee that post Cycle 6 inspection of St.

Lucie 1 CSB showed that no degradation had occurred to the repaired areas i of the CSB for the 18-month period encompassing Cycle 6 operation, (May 16, i 1984 to October 20,1985) based on the licensee's visual inspection which l revealed no crack extension or any plug and/or patch motion. The visual l

examination results were corroborated by measurements of the plug flange deflection which showed that, although some preload loss had occurred, the

, residual flange deflection would maintain sufficient preload for the plugs i

to provide their design function. The staff concludes that the licensee l has completed the commitment to perform a short-term inspection described in paragraph 10.5.1 of Reference 1.

The staff has also determined that the next examination of the repaired areas of the CSB can be performed during the next scheduled 10-year inservice inspection of the reactor vessel in accordance with the long-term inspection described in paragraph 10.5.2 of Reference 1. The licensee has performed the required reactor vessel examination for the first 10-year inspection interval and the next scheduled reactor vessel examination could occur in 1998. Therefore, the staff recommends that the licensee perform a detailed evaluation to determine the feasibility of an in-place visual examination of any accessible repaired area of the l

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V CSB to the extent that this examination can be accomplished without defueling the entire reactor core. A conservative schedule for performing this limited sample examination to confirm the integrity of the repairs, provided access is available, would be a refueling outage within a,4-6 year period from the last inspection. (

Conclusion The staff concludes that the licensee has completed the commitment to perform a short-term inspection described in paragraph 10.5.1 of Reference 1. The staff also concludes that the next cxamination of the repaired areas of the CSB can be performed during the next scheduled 10 year inservice inspection of the reactor vessel in accordance with the long-term insnection described in paragraph 10.5.2 of Reference 1. Finally, the staff recommends that the licensee preform a detailed evaluation to determine the feasibility of an in place visual examination of any accessible repaired area of the CSB to the extent that this examination can be accomplished without defueling the entire reactor core.

Principle Contributor:

M. R. Hum ~

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References 0

1. U.S. Nuclear Regulatory Connission, St. Lucie Plant, Unit No. 1, Docket No. 50-335, Safety Evaluation Related to Thermal Shield Recovery Program, Letter from J. R. Miller to J. W. Williams, Jr., dated March J4, 1984.

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2. Florida Power and Light Company, St. Lucie Plant, Unit No. 1,* Docket No.

50-335, Final Core Support Barrel Inspection Report (Post - Cycle 6),

Letter from C. O. Woody to A. C. Thadani, dated April 25, 1986.

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