Information Notice 1990-49, Stress Corrosion Cracking in PWR Steam Generator Tubes

Stress Corrosion Cracking in PWR Steam Generator Tubes

ML013100207
Person / Time
Site:	Millstone, Palisades, Indian Point, Sequoyah, Seabrook, North Anna, Zion, McGuire, Trojan
Issue date:	08/06/1990
From:	Murphy E, Rossi C Office of Nuclear Reactor Regulation
To:
References
+sunsimjr=200611, -RFPFR, FOIA/PA-2001-0256 IN-90-049
Download: ML013100207 (4)

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON,

D.C.

20555 August 6,

1990

NRC INFORMATION NOTICE NO.

90-49:

STRESS CORROSION CRACKING IN PWR STEAM

GENERATOR TUBES

Addressees

All holders of operating licenses or construction permits for

pressurized-water reactors (PWRs).

Purpose

This information notice is intended to inform licensees of recent problems

involving stress corrosion cracking (SCC)

in

PWR steam generator (,SG) tubes.

In particular, this information notice is intended to alert licensees to

recent findings at Millstone Unit 2 and to recent problems in SCC detection

during inservice inspections.

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 do not constitute NRC requirements; therefore, no

specific action or written response is required.

Description of Circumstances

1. Circumferential Cracking at Millstone Unit 2 In October 1989, the licensee for Millstone Unit 2 conducted a mid-cycle

inspection of the SG tubing using eddy current testing (ECT).

Circumferential SCC had been observed in previous inspections to be

affecting the outer diameter (OD)

surface (that is, the secondary side) of

the tubes at the expansion transition at the top of the tubesheet.

The

mid-cycle inspection followed a previous inspection during the February 1989 refueling outage and was performed, in part, out of concern for the

relatively high rate of SCC growth observed during the previous inspection

and to ensure that SCC did not excessively degrade the integrity of the

tubes.

Just before the mid-cycle outage, leakage from the primary side to

the secondary side was less than 5 gallons per day (gpd).

The plant's

Technical Specifications limit for such leakage is

144 gpd.

The ECT inspections at Millstone 2 were conducted with a rotating pancake

coil (RPC) probe to ensure optimal sensitivity to circumferential cracks.

Tubes found with circumferential crack indications were also inspected by

ultrasonic testing (UT) to obtain additional information regarding the

length and depth of the cracks.

In addition, two tubes with circumferential

crack indications were removed and examined.

9007310195 IN 90-49 August 6,

1990 The ECT/RPC inspections revealed 104 tubes with circumferential cracks at

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the expansion transition.

The macrocracks, as defined by ECT/RPC, consisted

of several discontinuous microcracks that were separated by small ligaments

of sound material.

The discontinuous nature of the array of microcracks was

confirmed by the UT and examination of the removed tube specimens.

As

measured by UT, the macrocracks ranged in circumference from 84 degrees to

329 degrees and ranged in depth up to 100-percent throughwall.

All tubes with crack indications were staked and plugged.

In addition, the

licensee evaluated the residual strength of the cracked tubes to assess

their capability to sustain normal operating and postulated accident

loadings before their removal from service.

This structural evaluation

considered the profiles for each crack obtained from the UT examination.

This evaluation revealed one cracked tube which failed to meet the ASME

Code,Section III, NB-3225 and Appendix F stress limits for postulated

accident conditions.

(Regulatory Guide 1.121,

"Bases for Plugging Degraded

PWR Steam Generator Tubes," states that margins should be consistent with

the stress limits in Section III of the code.)

Based on these findings, the

staff concludes that the integrity of the subject tube was not ensured under

postulated accident conditions.

The staff has recently identified service induced, circumfefential SCC,

such

as at Millstone Unit 2, to be a source of significant degradation to tubes

in

PWR steam generators.

Such cracking is particularly noteworthy because

.it

is generally not detectable with conventional bobbin probes used

routinely for inservice inspection.

Such cracking is generally only

detectable through the use of specialized probes, such as the RPC probe.

Most circumferential cracking has been observed at tube expansion

transitions at or near the top of the tubesheet.

In addition to Millstone

Unit 2, circumferential cracking at the expansion transition has recently

been identified at one other Combustion Engineering

(CE) plant (Maine

Yankee),

at three plants with Westinghouse Model 51 steam generators (North

Anna Unit 1, Trojan Unit 1, and Sequoyah Unit 1), and at one plant with

Westinghouse Model D steam generators (McGuire Unit 1).

Tubes in the

affected CE and Westinghouse Model 51 steam generators were explosively

expanded against the tubesheet.

Tubes in the McGuire Model D steam

generators were expanded against the tubesheet by mechanical rolling.

In addition to being found at the expansion transition location, widespread

circumferential SCC has been observed at drilled-hole support plate

locations at Palisades

(CE steam generators).

Isolated instances of

circumferential SCC have been reported at the uppermost support plate of a

pre-replacement Westinghouse Model 44 steam generator of Indian Point Unit 3 and at a row 1 U-bend of a Model 51 steam generator at Zion Unit 1. The

circumferential SCC at Palisades and Indian Point Unit 3 appears to be

associated with significant denting at the support plates.

IN 90-49 August 6,

1990 2.

Axial Cracking at Support Plates

Licensees have reported secondary side-initiated, axial SCC at several

plants with Westinghouse Model 51 and Model D steam generators at support

plate intersections exhibiting little

or no denting.

Recent difficulties

experienced in the detection of such cracks were the subject of Westinghouse

Customer Information Letter GEN-LTR-90-006,

"Steam Generator Tube Outer

Diameter Stress Corrosion Cracking at Tube Support Elevation-Eddy Current

Detection Issue," which was issued on or about February 8, 1990, to all

utilities

with Westinghouse steam generators.

Westinghouse reported that

metallographic examinations of tubes removed from the field have revealed

the presence of OD-initiated SCC at tube support plate (TSP) intersections

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that were not reported by personnel using a bobbin probe to perform field

eddy current tests.

For example, these examinations revealed one tube

containing axial cracks within two 30-degree-wide bands on opposite sides of

the tube, with the deepest crack penetrating to 62-percent throughwall.

The

field eddy current interpretation of the signal for this TSP location was

"Irno

detectable degradation"

(NDD) using the plant voltage threshold

criteria.

The EPRI "Steam Generator Examination Guidelines, Revision 2" contains

Figure C-55 showing the qualitative relationship between bobbin probe signal

amplitude and crack depth determined metallographically.

The staff believes:

that the amplitude threshold criteria used at the plant in the

above-mentioned example were taken from the actual data used to develop

Figure C-55.

The recent evidence cited by Westinghouse suggests that this

data is not conservative for all plants.

Industry meetings attended by representatives from a number of vendors

providing SG inspection services, EPRI,

and the Westinghouse Owners Group

have been conducted to examine various proposals to detect and measure SCC

1 at TSP locations.

The minutes of the EPRI Guidelines Revision Committee

meeting on February 13,

1990, note that general principles still

apply, pending development of updated guidance, for the detection and measurement

of SCC at support plates.

Section 4.6.1 of the EPRI guidelines states that

.as a general rule, an "extremely conservative position" should be -adopted

for the resolution of distorted indications or undefined signals not covered

by existing analysis guidelines.

Specifically, tubes with these types of

indications should be recommended for plugging unless other supporting data

exists (tube pulls or NDE diagnostic data) that justifies their retention as

active tubes.

Discussion:

The reliable detection and sizing of SCC during inservice inspections pose a

significant challenge to current ECT technology and practice in view of the

low signal-to-noise ratios frequently exhibited by such cracks.

Experience

IN 90-49 August 6, 1990 indicates that SCC is frequently not detected until it

has penetrated beyond

40-percent throughwall.

Fortunately, the vast majority of SCC flaws consist

of short axial or circumferential crack segments.

The staff believes that

such flaws can be detected before they grow sufficiently large to degrade

the structural margins of the tube to below the Regulatory Guide 1.121 criteria.

Tubes are generally inspected once per refueling cycle.

Depending on the

rate of crack growth and the number of tubes involved, this frequency may or

may not be sufficient to ensure that all cracks are detected before they

become sufficiently large to degrade structural margins to less than the

Regulatory Guide 1.121 criteria.

A structural assessment of the crack

geometries found during an inspection, such as performed at Millstone Unit

2, provides a means for assessing whether the inspection frequency is

sufficient to ensure adequate structural margins for all tubes between

inspections.

The staff believes that the effectiveness of eddy current testing for

detecting and sizing SCC can be enhanced through improved criteria for the

qualification and performance demonstration of the eddy current data

acquisition equipment (including probes), data analysis procedures, and data

analysts.

The staff and the industry, including the EPRI Steam Generator

Reliability Project, are evaluating this issue.

In the meantime, field

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experience indicates that careful attention to the potential for SCC at the

tubesheet region, at tube expansion transitions, at support plates, and at

tube U-bends is important.

The recommendations in Section 4.6.1 of the EPRI

steam generator examination guidelines indicate the importance of being

alert to distorted and undefined signals at these locations and employing

diagnostic measures as appropriate to establish the cause of these signals

and to validate the plant data analysis procedures and criteria.

Finally, the findings at Millstone Unit 2 illustrate that cracks will not

necessarily cause leakage approaching the Technical Specifications limit

before the structural margin in the affected tube drops below the Regulatory

Guide 1.121 criteria for postulated accident conditions.

This point is

further illustrated by the steam generator tube rupture (SGTR)

event that

occurred at McGuire Unit 1 on March 7,

1989, as a result of axially oriented

SCC.

Leakage before the SGTR event was about 15 gpd, which was small

compared to the plant's Technical Specifications limit of 500 gpd.

The

McGuire event occurred under normal operating conditions.

Thus, the McGuire

event was preceded by a period during which the subject tube was vulnerable

to rupture if

challenged by a postulated accident.

IN 90-49 August !6,

1990 This information notice requires no specific action or written response.

If

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

the technical contact listed below or the appropriate NRR project manager.

Charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contact:

E.

Murphy, NRR

(301)

492-0710

Attachment:

List of Recently Issued NRC Information Notices

ENDEND

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