ML20247P270

Forwards Conclusions from Util Final Rept of Engineering Evaluation of Bent Rock Bolt in Anchorage Sys Tying Torus Support Structure to Basemat Conducted in Response to NRC Insp.Review Not Requested

ML20247P270
Person / Time
Site:	Hatch
Issue date:	06/01/1989
From:	Crocker L Office of Nuclear Reactor Regulation
To:	Bagchi G, Cheng C, Jocelyn Craig Office of Nuclear Reactor Regulation
References
TAC-71384, NUDOCS 8906060142
Download: ML20247P270 (6)

Text

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MEMORANDUM FOR: Goutam Bagchi, Chief Structural and Geosciences Branch Division of Engineering and Systems Technology C. Y. Cheng, Chief Materials Engineering Branch Division of Engineering and Systems Technology J. W. Craig, Chief Plant Systems Branch Division of Engineering and Systems Technology FROM: Lawrence P. Crocker, Project Manager Project Directorate II-3 Division of Reactor Projects I/II

SUBJECT:

BENT TORUS ROCKBOLTS AT HATCH UNIT 1 (TAC N0. 71384)

During an inspection of the modifications made to the Hatch plant in response to the Mark I Containment Long-Term Program, the inspector observed in Hatch Unit I a bent rock bolt in the anchorage system tying the torus support structure to the basemat (Inspection Report 50-371,366/88-03). Subsequent licensee examination revealed that about 190 of the Unit i rock bolts were bent, with deflections ranging from 1/8" to 1" (Inspection Report 50-321, 366/88-38). Based upon this finding, the licensee initiated an engineering examination to determine the impact of the bent bolts on plant safety.

A meeting between the NRC staff and licensee representatives was held in Rockville on December 1,1968, to enable the licensee to present to the staff '

the preliminary results of the engineering examination. At the time, Hatch Unit I was within days of being ready to start up from a refueling outage.

Based upon information presented during the meeting, the staff consensus was that there were no safety concerns that posed an obvious impediment to Unit I restart or operation. However, the staff requested that the licensee complete the engineering evaluation and submit a final report to the staff. The December 1 meeting is documented in a meeting sumary dated December 8,1988.

Subsequent to the m6eting, the staff forwarded several questions to the licensee to be addressed in the final report (letter dated December 14,1988).

By letter A ted March 23, 1989, the licensee forwarded the final report of the engineering evaluation of the anchor bolts. The evaluation includes an analysis of the root cause of the bent bolts, an evaluation of load-carrying j capacity of the bolts in their present condition, an analysis of conservatism i in the postulated uplift loads on the torus, a comparison of the calculated l bolt loads versus the bolt load capacity, and an analysis of the effects of l shrinkage-induced stresses on attachments to the torus. The report concludes j that the bolt deformation is attributable primarily to shrinkage of the torus i due to welding operations conducted after the bolts were installed, that the i

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Multiple Addressees .

capacity of the deformed bolts is adequate to withstand postulated uplift forces, and that the residual stresses on the torus pose no problem to the attachments.to the torus. In effect, the final evaluation confirms the tentative information that was presented at the December 8,1988 meeting. A copy of the conclusions from the final report is enclosed.

In view of the earlier staff consensus that the deformed bolts posed no significant safety concerns, and the corroborating nature of the licensee's final evaluation report, there appears to be no need to commit significant staff resources to a formal review of the March 23, 1989 submittal. Accordingly, I am not requesting a review. However, a copy of the final report is attached for your information and for such review as you may deem necessary or desirable. I also have several spare copies of the report should you need an additional copy. Review efforts, if any, should be charged against TAC No.

71384 Unless I hear to the contrary from you before June 15, 1989, I plan to simply acknowledge receipt of the report to the licensee and close out any further effort on this matter.

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Lawrence P. Crocker, Project Manager Project Directorate II-3 Division of Reactor Projects I/II

Enclosure:

As stated

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Multiple Addressees l' capacity of the deformed bolts is adequate to withstand postulated uplift forces, and that the residual stresses on the torus pose no problem to the attachments to the torus. In effect, the final evaluation confirms the tentative information that was presented at the December 8, 1988 meeting. A copy of the conclusions from the final report is enclosed.

In view of the earlier staff consensus that the deformed bolts posed no significant safety concerns, and the corroborating nature of the licensee's final evaluation report, there appears to be no need to commit significant staff resources to a formal review of the March 23, 1989 submittal. Accordingly, I am not requesting a review. However, a copy of the final report is attached for your information and for such review as you may deem necessary or desirable. I also have several spare copies of the report should you need an additional copy. Review efforts, if any, should be charged against TAC No.

71384.

Unless I hear to the contrary from you before June 15, 1989, I plan to simply acknowledge receipt of the report to the licensce-and close out any further effort on this matter.

MJ Lawrence P. Crocker, Project Manager Project Directorate 11-3 Division of Reactor Projects I/II

Enclosure:

As stated DISTRIBUTION 1 Docket File - LCrocker, w/ encl. CSellers, w/o enci.

NRC & Local PDRs JKudrick, w/o enc 1. KManoly, w/o encl.

PDII-3 Reading DJeng, w/o enc 1. JBlake, w/o encl.

DMatthews w/o encls. HAshar, w/o enc 1. RChou,w/ enc 1.

MRood, w/o enc 1. CTan, w/o enc 1. BBorchardt, w/o enc 1.

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. Enelnura

' In February 1988, during an NRC inspection of the modifications made to the Plant Hatch containment systems as i

a result NRC inspector of the Mark I short-term identified a Unit and long-term programs, an 1 torus rock bolt which was bent inwards (toward the reactor).

The Plant Hatch Unit 1 torus mitre joint rock bolts have experienced could not be assumed sufficient angular deflection such that they design basis torus uplift loads. However, studies have been to withstand the extremely conservative performed which show that more realistic torus uplift loads can be accommodated during normal operations, anticipated transient conditions, or LOCA conditions and that no other containment components or structures are affected by this condition.

weld inducedThe root cause shrinkage of the of the rock bolt perimeter deformations of the torus. This is resulted from the sequencing of the modifications performed under the Mark I short-term and long-term modifications.

The Mark I containment structure consists of the drywell, vents (connecting the drywell to the torus), and the steam suppression chamber (torus). The torus is supported by 64 columns, with inner and outer columns located at each mitre joint (where the bays are connected) and two similar columns at each midbay. Each column base plate is positioned on a self lubricating bearing plate to allow the torus support system to move radially to accommodate torus temperature fluctuations.

The Mark I short-term and long-term programs involved performing a plant unique analysis to address the hydrodynamic design basis accident loads the containment would experience during a (and other blowdown events) and making the necessary postulated loads.

structural modifications to accommodate these One of the initial modifications made was the joint installation, columns. in 1976, of rock bolts to anchor the mitre Specifically, the design function of the bolts is to restrain the vertical motion of the columns and therefore the torus during the postulated uplift loading.

These design basis uplift loads result from hydrodynamic phenomena associated with a LOCA and other blowdown phenomena.

The rock bolts do not provide any radial, axial, or compressive support to the torus / support column structure.

about 5 million Thepounds weightofofwater) the torus structure (including and the design of the containment structures ensure the realistic uplift loads will be accommodated during normal operations, anticipated transient conditions, or LOCA conditions. In addition, the design basis containment load definition in the Unit 1 PUAR contains many conservatism that can justifiably be removed due to advances in methodology. Portions of this methodology p e.

have already been approved by the NRC at other U. S. licensed plants with Mark I containments. Therefore, the design basis  ! !1 uplift loads can be justifiably reduced to within the J demonstrated structural capacity of the rock bolts.

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Under the short-term program, structural modifications were

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performed in response.to an expedited assessment of the adequacy l

ofEthe containment to maintain its integrity under the loads.

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' postulated to occur during a LOCA. The installation of the nitre joint rock bol'ts was part of the short-term program modifications. At this point, the structural modifications which would result from the long-term program could not be~ foreseen. In

.the long-tera program, the detailed. containment load definition was established and the necessary structural modifications were made in 1983. A calculated radial displacement of'the torus.of approximately 0.5 inches resulted from weld induced shrinkage due

'to the installation of T-stiffeners, saddle supports, and conical

web plates. This radial displacement has been analytically shown to constitute a displacement approximately equal to the measured bolt deformaticas.

In July 1988, a detailed inspection was completed by the architect engineer's (A/E's) personnel of the 384-rock bolts on Unit 1.

Radial inward deflection of the rock bolts was determined to be widespread, with the worst cases being the nitre joint rock' bolts. The angular deviations on the midbay supports are small and can be attributed to construction installation. The worst case deflection on the mitre joint rock bolts is about 1 inch over an. approximate 6 inch distance. The average outside-column displacement is about 1/2 inch and the average inside column displacement is about 1/4 inch.

A team, composed of corporate and-A/E engineering personnel and consultants experienced in the Mark I program modifications, was formed to evaluate the deflection of the Unit 1 rock bolts. In.

early August, ultrasonic testing of 48 of the more severely bent bolts provided initial confirmation of bolt integrity. On August 15, 1988, a qualitative operability assessment was completed which was based on the ultrasonic testing results, the small probability of a Loss of Coolant Accident (LOCA) occurring which would impose the bounding vertical uplift loads, a preliminary evaluation of the bolts' structural capacity, and the significant conservatism in the Plant Hatch containment load definition.

In late August, some additional, more refined, ultrasonic testing L was performed which again confirmed the integrity of the rock bolts. Efforts were initiated to quantify the structural capacity of the deflected rock bolts and reduce the conservatism in the containment load definition. Other items potentially impacted by the contraction of the torus were reviewed and evaluated. Th6 NRC was briefed on the status of these efforts as well-as the root cause determination at a meeting on December 1, 1988. At the time, NRC staff indicated their concurrence with the operability l assessment and the root cause determination.

The purpose of this report is to document the methodology and process used by Georgia Power Company to evaluate the capacity of deformed anchor bolts in the Unit 1 suppression

, cha'aber supports. The report clearly shows.that the bolt deformations were caused by weld induced shrinkage during the E. I. Hatch Unit 1 Mark I Containment Long Term Program. The report also demonstrates that.in-situ bolt capacity is sufficient to resist justifiable containment uplift loads and that no other attachments have been adversely affected by the weld induced shrinkage of the suppression chamber.

. Based on the information contained in this report, it is concluded that the veld induced shrinkage of the torus and the deflected rock bolts have no adverse impact on nuclear safety.

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