View source for Information Notice 1992-80, Operation with Steam Generator Tubes Seriously Degraded

{{Adams
| number = ML031190743
| issue date = 12/07/1992
| title = Operation with Steam Generator Tubes Seriously Degraded
| author name = Grimes B
| author affiliation = NRC/NRR
| addressee name = 
| addressee affiliation = 
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| license number = 
| contact person = 
| document report number = IN-92-080, NUDOCS 9212010198
| document type = NRC Information Notice
| page count = 6
}}
{{#Wiki_filter:UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 December 7, 1992 NRC INFORMATION NOTICE 92-80:    OPERATION WITH STEAM GENERATOR TUBES SERIOUSLY

DEGRADED

==Addressees==
All holders of operating licenses or construction permits for pressurized

water reactors (PWRs).

==Purpose==
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information tube

notice to inform licensees of recent findings from steam generator (SG) The

inspections and investigations at Arkansas Nuclear One Unit 2 (ANO-2).

Arkansas Power and Light Company, the licensee for ANO-2, found three       tubes to

be degraded to the point where they no longer retained    adequate  structural

margins to sustain the full range of normal operating, transient, and

postulated accident conditions without rupture. It is expected that

recipients will review the information for applicability to their facilities

and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements;
therefore, no specific action or written response is required.

==Description of Circumstances==
In 1978, the NRC licensed ANO-2 for operation. ANO-2 is a two-loop PWR

designed by Combustion Engineering, Incorporated (CE). On March 9, 1992, the

licensee shut down ANO-2 upon detecting a primary-to-secondary leak of

0.95 liters per minute [0.25 gallons per minute]; half of the technical

specification limit. The licensee conducted an eddy current inspection of the

the

SG tubes using a motorized rotating pancake coil (MRPC) probe and found    leg

source of the leak to be a circumferential  crack in a  tube at  the  hot

expansion transition location, which is near the top of the tubesheet. The

licensee reviewed the eddy current test data from the previous refueling coil

outage inspection in 1991 and found that this tube had exhibited a bobbin

indication at that time. Two independent data analysts had missed this the

indication. The licensee found six other bobbin coil indications that

data analysts had also missed. The licensee reports that, if these

indications had been correctly analyzed, the licensee would have evaluated

them further. It is common industry practice to perform supplemental MRPC

inspections (and sometimes pulled tube examinations) to better characterize

low amplitude, ambiguous, or distorted bobbin coil indications. In 1991, the

licensee did not perform MRPC inspections at the expansion transition

locations of these steam generators.

9212010198      loner                Pa          7ctE-C      tOo

br-I;e      - // C_-

.1.

IN 92-80
                                                                            December 7, 1992 Because of the finding of the circumferential

100 percent MRPC inspection of the expansion crack, the licensee conducted a

leg side of both steam generators and              transition locations on the hot

a   2 0-percent

expansion transition locations on the                       MRPC inspection of the

cold

The licensee found indications (generally       leg   side   of one steam generator.

of 488 tubes. The licensee found no indicationscircumferential)        on the hot leg side

side. Tubes with MRPC indications were                    in  the   tubes  on the cold leg

however, this probe did not detect most    also     inspected     with  a  bobbin probe;
   sleeved 448 of the tubes with indications  of    the   MRPC   indications.     The licensee

remaining 40 tubes. The licensee pulled        and   plugged     and  stabilized  the

tubes with circumferential indications       a   number    of  tubes,   including  three

to measure the extent of damage and to in the expansion transition location, determine the degradation mechanism.

On August 7, 1992, the licensee reported

from the examination of the pulled tube to the NRC the preliminary results

segments. On three of the tubes, the

examination found circumferentially oriented

cracking (IGSCC) beginning on the outer            intergranular stress corrosion

extended 360 degrees around each of the    diameter       of the tubes. The cracks

between 88 to 94 percent of the tube wall  tubes      and  had average depths ranging

that all three tubes failed to satisfy         thickness.        The licensee reported

the

Regulatory Guide 1.121, "Bases for Plugging      structural     margin    criteria of NRC

Tubes." Thus, at the time of the March            Degraded PWR Steam Generator

inadequate structural margins to ensure 1992 shutdown, these tubes retained

normal operating, transient, and postulatedtheir integrity for the full range of

accident conditions.

Discussion

The licensee attributed the missed bobbin

to (1) a lack of training for the eddy          indications from the 1991 inspection

tube damage mechanisms specific to ANO-2  current      data analysts in guidelines for

performance demonstration test of the        and    similar     sites; (2) the lack of a

data

and (3) the inherent difficulties in analyzing   analysts      using   actual site data;
transition locations caused by interference            signals     at  the expansion

transition geometry and deposits on the            from   the   tubesheet,    the expansion

have been missed if the entire tubesheet   SG    tubes.      The   indications   may also

distortions that could indicate flaws        entry signal was not screened for

at the expansion transition locations.

Examining the entire tubesheet entry signal

location of the expansion transition in           is important because the precise

and thus the location of the cracks, variesrelation       to the top of the tubesheet, among tubes.

===Although the above may have been contributing===
that the failure to prevent the excessive             factors, the NRC staff believes

ANO-2 resulted primarily from using test       loss    of tube structural margin at

locations susceptible to circumferential     probes      inappropriate for tube

cracking.

other CE plants, the tube expansion transition                 Based on experience at

generators are known to be susceptible                 locations      in CE steam

to   circumferential

probes are generally insensitive to circumferential                    IGSCC. Bobbin

IGSCC unless the cracks

K>                               '-J     IN92-80
                                                             December 7, 1992 are large enough to have a significant axial component or crack opening.

Before this condition could occur, the structural margins of the tube would

already be substantially reduced.

Circumferential cracks can be detected satisfactorily only by specialized

probes such as the MRPC. This has has been discussed extensively in industry

literature and reported by the NRC. In Information Notice 90-49, "Stress

Corrosion Cracking In PWR Steam Generator Tubes," the NRC alerted licensees to

problems with such cracks at Millstone Unit 2 and at Maine Yankee. During the

ANO-2 SG tube inspections in 1991, the licensee did not perform MRPC inspec- tions of the expansion transition locations.

Effective inspection programs to ensure early detection of SG tube degradation

can be achieved through the use of data acquisition equipment and procedures

and data analysis procedures appropriate for the detection of potential flaw

types, including circumferential IGSCC. Training and performance

demonstration testing of the data analysts for all potential flaw types are

important parts of such programs. Further, by reviewing relevant industry

experience, insights may be gained into the types of flaws that affect a

particular SG as a function of the design or fabrication process. For

example:
      Industry reports show that circumferential IGSCC indications have

been observed at the tube expansion transition locations of SGs at

four CE plants (including ANO-2) and at several Westinghouse

plants. Tubes in the SGs at the four CE plants were explosively

expanded against the tubesheet using the CE "explansion" process.

Tubes in seven of the affected Westinghouse plants with Model 51 SGs were explosively expanded against the tubesheet using the

Westinghouse WEXTEX process. Tubes in two of the affected

Westinghouse plants with Model D SGs were expanded using a

mechanical rolling process over the full depth of the tubesheet.

Tubes in another affected Westinghouse plant with Model 51 SGs

were expanded over only the lower 2-1/2 inches of the tubesheet

thickness (partial depth expansion) using a mechanical rolling

process.

Widespread circumferential IGSCC has also been observed at tube

support plate intersections in Westinghouse Model 51 SGs at the

North Anna Power Station Unit 1 and in the SGs that were replaced

at the Palisades Nuclear Power Station. Axial stress caused by

denting is believed to have caused the cracks in these steam

generators. Denting at the top of the tubesheet crevices in

Westinghouse Model 51 SGs at the Donald C. Cook Plant (Cook)
      Unit I appears to have caused circumferential SCC at these

locations. The tubes in the SGs at Cook were partial depth

expanded against the tubesheet. Isolated instances of

circumferential cracks have also been observed at the upper "egg- crate" supports and U-bends in the CE SGs that were replaced at

<_<IN 92-80
                                                              December 7, 1992 the Millstone Nuclear Power Station Unit 2 and at a row 1 U-bend

in the Westinghouse Model 51 SGs at the Zion Nuclear Power Station

Unit 1.

This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

the te rnical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:   E. Murphy, NRR

(301) 504-2710
Attachment:   List of Recently Issued Information Notices

S      (2y(c~c

,  I -
                                                         KIttachment

IN 92-80
                                                           December 7, 1992 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information                                   Date of

Notice No.             Subject                Issuance     Issued to

92-79,         Non-Power Reactor              12/01/92     All holders of OLs or CPs

Emergency Event Response                    for test and research

reactors.

92-78          Piston to Cylinder             11/30/92     All holders of OLs or CPs

Liner Tin Smearing on                       for nuclear power reactors.

Cooper-Bessemer KSV

Diesel Engines

92-77          Questionable Selection         11/17/92     All holders of OLs or CPs

and Review to Deter-                        for nuclear power reactors.

mine Suitability of

Electropneumatic Relays

for Certain Applications

92-76          Issuance of Supple-            11/13/92     All holders of OLs or CPs

ment 1 to NUREG-1358,                       for nuclear power reactors.

"Lessons Learned from

the Special Inspection

Program for Emergency

Operating Procedures

(Conducted October 1988  -
               September 1991)"
92-75          Unplanned Intakes of           11/12/92     All holders of OLs or CPs

Airborne Radioactive                        for nuclear power reactors.

Material by Individuals

at Nuclear Power Plants

92-74          Power Oscillations at          11/10/92     All holders of OLs or CPs

Washington Nuclear                          for nuclear power reactors.

Power Unit 2
92-61,         Loss of High Head              11/06/92     All holders of OLs or CPs

Supp. I        Safety Injection                            for nuclear power reactors.

92-73          Removal of A Fuel              11-04/92     All holders of OLs or CPs

Element from A Re-                          for nuclear power reactors.

search Reactor Core

While Critical

OL = Operating License

CP = Construction Permit

I

U'             -
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on recycled

paper

Federal Recycling Program}}

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