ML20005G306

LER 89-020-00:on 890921,abnormal Degradation of Steel Containment Vessels Observed.Caused by Corrosion by Standing Water in Annulus Areas.Steel Containment Vessels Will Be Recoated During Next Refueling outage.W/900109 Ltr

ML20005G306
Person / Time
Site:	Catawba
Issue date:	01/09/1990
From:	Glover R, Owen T DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-020-02, LER-89-20-2, NUDOCS 9001180408
Download: ML20005G306 (6)

Text

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• Ilaar Iburr Cornpany (203) M13000 ^-

. Calouba Nuclear Station PO tha 256 Civver, S C 29710 DUKE POWER January 8, 1990  !

Document Control Desk-U. S. Nuclear Regulatory Commission .

Washington, D. C. 20555  :

Subject:

Catawba Nuclear Station Docket No. 50-413 LER 413/89-20 Gentlemen:

Attached is Licensee Event Report 413/89-20 submitted as a Courtes'y Report concerning both abnormal degradation of the steel containment vessel due to corrosion by standing water in~the annulus areas.

This event was considered to be of no sianificance with respect to.the health and safety of the public.

I Very truly yours, j l w

. Owen Ton j

1 Stat on Manager keb:COLRTESY.LER xc: Mr. S. D. Ebneter American Nuclear' Insurers Regional Administrator, Region II c/o Dottie.Sherman, ANI Library I U. S. Nuclear Regulator commission The Exchange, Suite 245 l 101 Marietta Street, NW, Suite 2900 270 Farnington Avenue Atlanta, GA 30323 Farmington, CT 06032 _]

-l M & M Nuclear Consultants Mr. K. Jabbour-1221 Avenues of the Americas U. S. Nuclear Regulatory Commission New York, NY 10020 Office of Nuclear Reactor Regulation Washington, D. C. 20555:

fr INPO Records Center Suite 1500 Mr. W. T. Orders 1100 circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station l[{ i

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'"" Abnormal Degradation of Steel Containment Vessels Due To Corrosion By Standine Water In The Annulus Areas EVENT DATE 198 LER NUMBER ter REPORT DATE 47s OTHER F ACILITIES INVOLVED lel MONTH DAY vtAR vtAR y y '^6 , [*ff,'$ MONTH DAY vgAR f ACit Tv NaMis DOCKET NUMBER!si CNS. Unit 2 015 l 0 l 0 l 014 l 1l 4 q1 89 q1 0l9 0l 9 8l9 0 l 2l 0 0l0 9l 0 MNS, Unit 1 o islo 1 0 1 0 : 31n 19 THIS REPORT IS SUBMtTTED PUR5UANT TO THE REOVIREMENTS OF 10 CFR 9 (Cneca pne er more of the forsawmpf (11)

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NAME TELEPHONE NUMBER R.M. Glover, Compliance Manager 8 l ol 3 81311 l- 131213 I 6 COMPLETE ONE LINE FOR E ACM COMPONENT F AILURE DESCRl8ED IN THl$ REPORT (131 CAvst 8vsf EM COMPONENT "# 'C "fo$,1 u CAUSE Sv8 TEM COMPONENT "A AU^ "T^ '

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% ES (19 vos comotore EXPECTED SUD4rtS$10N OA TE) NO l l l 1.087RACT ftsmit to 1400 speces. i e eparesperet erreen s.ap.e space typewestren imess (16l On September 21, 1989, a preliminary visual inspection of the Catawba Units 1 and 2 Steel Containment Vessel (SCV) exterior surfaces was performed between azimuths 0 degrees and 360 degrees at elevation 552 feet + 0 inches. Units 1 and 2 were in Mode 1, Power Operation, at 100% and 98% power respectively, when the inspection was conducted. The observed corrosion was caused by standing water in the Annulus Areas. The most significant corrosion occurred in areas where boric acid deposits were also founo. The root cause of this event is assigned as a design oversight because the slope of the Annulus floors was not sufficient to cause flow to the installed floor drains. A contributing cause of Unanticipated Environmental Interaction is also assigned because of the chemical reaction between the boric acid and SCV. Design Engineering personnel evaluated the extent of corrosion by comparison to the conditions previously observed at McGuire Nuclear Station. The corrosion observed at the Catawba SCVs bases is not as advanced as at McGuire and does not warrant a separate operability review. The SCV will be repaired and recoated, as applicable, during the next two refueling outages for each Unit. This is a Courtesy LER.

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BACKGROUND

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' f' The general purpose of the Reactor Building and the Containment Vessel i [EIIS:VSL] is to provide a barrier confining potential releases of radioactivity i resulting from design basis accidents. This is accomplished by maintaining leak y tightness within specified bounds. As a design feature, the Reactor Building ,

and Containment Vessel are provided primarily for the protection of public. l s

health and safety. The free standing steel containment has an outer reinforced  ;

- concrete Reactor Building and an annular space which is maintained at a-  ;

lower-than-atmospheric pressure following a loss of Coolant Accident (LOCA). j These structures form a double barrier to prevent the escape of fission products  :

should a LOCA occur.

j The Steel Containment Vessel (SCV) is an ASME Class MC free-standing welded

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steel structure consisting-of a vertical cylinder with a hemispherical dome and

' a flat circular base. The Containment shell is anchored to the Reactor Building I foundation by 182 anchor bolts equally spaced around the perimeter of the cylinder base. The flat base of the Containment vessel is a 1/4 inch liner plate encased in concrete which functions only as leak-tight membrane and is not designed to serve as a structural element.

The vertichi cylinder and. hemispherical dome are constructed from carbon steel )

plate, ASME specification SA516 GR60. The lower approximate 14 feet of the .

vertical cylinder is 1 inch nominal thickness. The upper cylinder is 3/4 inch nominal thickness and the dome is 11/16 nominal thickness.

The Code of Federal Regulations,. Title 10, part 50 (10CFR50), Appendix J,Section V, Subsection A, requires that a general-inspection of the Containment ,

vessel be performed prior to each Type A Integrated Leak-Rate Test (ILRT). This inspection is intended to identify any evidence of structural deterioration and i interferences which may affect either the Containment structural integrity or i leak tightness.

Technical Specification 4.6.1.6 states in part, that the structural integrity of the Containment vessel shall be determined during the shutdown for each Type A ,

Containment leakage rate test by visual inspection of-the exposed accessible interior and exterior surfaces of the vessel.

The Annulus area concrete floor was designed to circumferentially slope toward four equally spaced drains located at azimuths 45, 135, 225, and 315 degrees.

The Annulus area, including the outside of the SCV, is coated to Service Leve?.

II requirements. Service Level 11 coating is meant for surface areas outside the primary Containment facilities that are subject to radiation or contamination. Service Level 11 coating is not intended for immersion conditions. ,

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0 l0 0l3 oF 0l5 stxi w me. ume e o. eau. a **m,w unc re= mw on During the previous structural inspections conducted prior to Integrated Leak Rate Testing, the SCV corrosion was not identified because the SCV/ concrete floor interface is inaccessible due to HVAC duct.

EVENT DESCRIPTION The corrosion of the McGuire Nuclear Station Unit 2 Steel Containment vessel ,

discovered during the structural inspection prior to an Integrated Leak Rate l Test (ILRT) prompted, by procedure, an inspection of the Catawba Units 1 and 2 SCV for similar type degradation (see McGuire LER 369/89-020). On September 21, 1989, Design Engineering and Catawba Station personnel performed a preliminary visual inspection of Units 1 and 2 SCV exterior surfaces between azimuths 0 degrees and 360 degrees at elevation 552 feet + 0 inches. Units 1 and 2 were in Mode 1, Power Operation, at 100% and 98% power, respectively, when the inspection was performed. The observed corresion was caused by standing water in the Annulus Area with the most significant corrosion occurring in areas where boric acid deposits resulted from leaking instrumentation connections.

CONCLUSION The root cause of this event is assigned a design oversight due to the unaaticipated interaction between the SCV and leaking instrumentation lines. The slope of the Annulus floor was not sufficient to cause flow to the installed floor drains. The design is considered adequate to handle large water spills.

Design Engineering personnel evaluated the extent of corrosion by comparison to the conditions previously observed at McGuire Nuclear Station. The McGuire inspections found corrosion with an average depth of 0.1 inch with pits of up to 0.125 inch. The corrosion observed at the Catawba SCV base is not as advanced as at McGuire and does not warrant a separate operability review. The need to continue weekly or bi-weekly Annulus inspections, as well as the need for periodic coating inspections in the Annulus, will be evaluated.

The pooling conditions allowed time for the water to infiltrate the coating to the SCV through pin hole surface imperfections that are allowable under Service Level II requirements. The water could then wick its way between the coating and SCV. Although no accumulation of water had been identified in previously documented pre-ILRT inspections, pooling in the Annulus area is evident from previous observance of standing water and from multiple water level lines on the SCV and other Annulus area components. The number of occasions or the length of time that water was standing in the Annulus areas could not be determined.

Inspection accessibility will be improved by removing or relocating HVAC ducts.

Areas not already inspected due to inaccessibility will be inspected.

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2 i A contributing cause of Unanticipated Environmental Interaction because of the i chemical reaction between the low carbon steel and boric acid is assigned, f Corrosion is more prevalent in areas where boric acid deposits were present.

These boric acid deposits were from leaking instrumentation line connections.

J Immersion of the SCV base in pooled, borated water was not anticipated when

, Service Level II coating was prescribed. Concrete floors and the lower portions of the SCVs will be recoated or topcoated and sealed for immersion conditions.

The need for other SCV protection barriers, or repairs to the Annulus floor slope, will be evaluated after other corrective actions are implemented and

} their effectiveness is assessed.

. This event is not Nuclear Plant Reliability Data System (PPRDS) reportable.

I LER 414/89-017 describes a previous incident in which the Turbine Driven-Auxiliary Feedwater Pump [EIIS:P] was rendered inoperable due to control valve

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V] stem corrosion; this incident, however, did not involve a design

oversight / unanticipated system interaction. Other drain problems have been associated with blockage, debris,.etc. Corrosion / degradation of.the SCV is not considered a recurring problem at Catawba.

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= There were no personnel injuries, radiation overexposures, or releases of radioactive materials as a result of this event. This report is submitted as a Courtesy LER.

y CORRECTIVE ACTION SUBSEQUENT

1) Operations began performing weekly surveillances of both Unit 1 and Unit 2 Annulus areas to identify and remove any standing water or boric acid deposits. No major problems were identified from these inspections. Bi-weekly inspections began in December,1989.

PLANNED l

= 1) Those areas of the Unit 1 and Unit 2 Annulus where significant coating failures were observed will be inspected and repaired, as required, in their upcoming outages. In addition, the interior face of the SCV will be inspected.

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2) Nuclear Station Modifications will be written to relocate or remove the HVAC duct along both Unit 1 and Unit 2 Annulus floors to make those areas accessible for visual examination.

3) Remaining Unit 1 and Unit 2 Annulus Areas will be fully inspected and repaired, as required, following duct relocation / removal. In addition, the interior face of the SCV will be inspected.

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$ 4) Tne SCVs in both Unit I and Unit 2 will be weld repaired as necessary 3 per applicable acceptance criteria. Concrete floors will be recoated 1 'or topcoated and sealed for immersion conditions. The SCVs will be 7

recoated or topcoated and sealed for immersion conditions up to 9 inches above the floor base.

p 5) Installation of SCV protection barriers (water dam and diverter) or repair to the floor slope will be considered after the effectiveness of the above corrective actions can be evaluated.

H 6) The need to continue weekly or bi-weekly inspections of the Annulus b area and the need for periodic coating inspections in that area will p be evaluated.

SAFETY ANALYSIS a

j The design of the Catawba Containment is based on analysis in accordance with p ASME Section III, NE-3200. Supplemental calculations using code-formula i

i equations (NE-3300) are alio contained in the Catawba SCV Stress Report. These show the minimum required ICV wall thickness at the base to be 0.6875 inches as e

i governed by external pressare (NE-3133). Re-analysis would give a lower value than that determined by formula but is not warranted at this time. Thus, the

' Catawba SCVs can be considered operable if the minimum wall thickness at the base is greater than 0.6875 inches.

The preliminary visual inspection of the. Catawba SCVs revealed that the corrosion present is not nearly as advanced as that observed at McGuire. The A time frame for corrective actions at McGuire and Catawba are similar (17 months 4

and 24 months, respectively).' Based on the calculated worst case corrosion rate at McGuire (38.5 mils per year), the amount of corrosion that will occur during the delayed time for implementation of corrective actions will not result in.

further significant loss of metal in the 1 inch nominal shell plate at the base of the Catawba SCVs.

As discussed above, the SCVs were not rendered inoperable as a result of the corrosion and subsequent loss of metal, and is capable of fulfilling its intended safety functions. The health and safety of the public were not affected by this event.

MC FOMM MSA 'U.S. GP0s 1988 520 58'J/ M M