05000244/LER-2077-003, Abnormal Degradation of Steam Generator Tubes

Abnormal Degradation of Steam Generator Tubes

ML18142B235
Person / Time
Site:	Ginna
Issue date:	05/16/1977
From:	White L Rochester Gas & Electric Corp
To:	O'Reilly J NRC/IE, NRC Region 1
References
LER 1977-003-00
Download: ML18142B235 (15)

LER-2077-003, Abnormal Degradation of Steam Generator Tubes

Event date:
Report date:
2442077003R00 - NRC Website

text

U.S, NUCLEAR REGULATORY COMMISSION brfiG FOR.I 195 I2.78)

.<<'IIC DISTRIBUTION ro<<PART 50 DOCKET MATERIAL DOCKET NUMBER EO-24/ 0

'ILE NUMBER INCIDENT REPORT TO: James P. OlReilly FROM RGB Rochester, N, Y.

14649 L. D. White,, Jr.

DATE OF DOCUMENT 05-16-77 DATE RECEIVED 05-19-77 ETTE R 0 ORIGINAL

$RCOPII ONOTORIZED

~No LASS IF I E D PROP INPUT FORM NUMBER OF COPIES RECEIVED

DESCRIPTION

OWL,EQGED ENGLosURF Licensee Event Report (RO-50-244/77-0

)

on 05-02-77 concerning Steam Generator

tubes, 3

tubes in th6 "A" SG"and 1 tube in"'the IIB" SG.,i wA showed defects above the wastage criteria of 0

due to c'oncentrations of residual"phosphhtes...

With attahced, Steam Generator'nspection Fina3.

Report 05-11-77.;..

PLANT NAME:

R. E..Ginna Unit g 1

(

11 pages

)

r

~)

DO BOY H,EMOVE,,

NOZE.

IF PERSONNEL EXPOSURE IS INVOLVED SEND DIRECTLY TO KREGER/J COLLINS BRANCH CHIEF:

W 3 CYS FOR ACTION LIC, ASST':

W

, CYS

- ACRS Q CYS RMEEBXNG' NT.,

, FOR ACTION/INFO RMATION INTERNALD ISTR I BUT)ON

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- NRC 'PDR

=-.I k4,. 2 XPG-SCHROEDER/IPPOLITO HOUSTON NOVAK/CHECK GRIMES A

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- BUTLER Hh UER TEDESCO/MACCARY EISENHUT BAER

- SERO VOLLMER/BUNCH KREGER/J COLLINS LPDR:

TIC:

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EXTERNALDISTRIBUTION CONTROL NUMBER 771410010 NRC FORM 196 I2.78)

4'

t v

ROCHESTER GAS AND ELECTRIC CORPORATION t

Nlg g> ~ ~ yvl vv I P IIAIC o

89 EAS T AVENUE, ROCH ES TER, N.Y. 14649 LEON D. WHITE, JR.

VICC I'RCSIDCNT May 16, 1977 IKMUITM7RING RifIIPg Mr. James P. O'Reilly, Director U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 TCCCI'IIONC ARCA CODC TIG 546.2700 2

c+

~CI7

Subject:

Reportable Occurrence 77-03 (14-day report), Abnormal degradation of steam generator tubes R. E. Ginna Nuclear Power Plant, Unit No.

1 Docket No. 50-244

Dear Mr. O.'Reilly:

In accordance with Technical Specifications, Article 6.9.2a, the attached report of Reportable Occurrence 77-03, 14-day, is hereby submitted.

Two additional copies of this letter and the, attachment are enclosed.

Very truly yours, L. D. White, Jr.

Attachment cc:

Dr. Ernst Volgenau (40)

Mr. William G. McDonald (3) 77i4iOOiO

CONTROL BLOCK:

1 UCENSEE NAME QogNYREG100 7

89 14 15 LER 77-03AT LICENSEE EVENT REPORTS (PLEABE PRINT ALLREQUIRED IIIIFORMATION) 6 UCENSE EVENT LICENSE NUMBER TYPE TYPE 0

0 0

0. 0 0

4 1

1 1

1

~01 25 26 30 31 32 REPORT REPORT CATEGORY TYPE SOURCE DOCKET NUMBER EVENT DATE AEPORT DATE

~01 CON'T~

T L

0 5

0 0 2

4 4

0 5

0 2

7 7

0 5

1 5

7 7

7 8

57 58 59 60 61 68 69 74 75 80 EVENT OESCRIPTION During the planned Eddy Current inspection of the Steam Generator (SG) tubes, 13 7

8 9I tubes in the "A" SG and 1 tube in the' B" SG showed defects above the wastage 7

8 9 to@~]

criteria of 40%

~

These tubes have been plugged.

(Reportable Occurrence 77-03, 7

8 9 Qpg 14-day).

7 8 9 rm W

1 2

0 7

8 9 [ops]

complete Eddy Current program description and report is attached.

7 8 9 7

8 9 PAME SYSTEM

CAUSE

COMPONENT COMPONENT CODE CODE COMPONENT CODE SUPPUEA MANUFACTURER VIOIATION

~CI

~B H T 5

X C

H

~I4 N

7 8 9 10 11 12 17 43 44 47 48 CAUSE OESCRIPTION [oOs]

Localized corrosion of tubes caused b

concentrations of residual hos hates.

A 80 80 80 80 80 80 80 OTHER STATUS OFFSITE CONSEQUENCES 7

89 FACILITY STATUS 55 POWER Q~g

~H

~oo o

NA 7

8 9

10 12 13 FORM OF ACTIVITY COATENT RELEASEO OF RELEASE AMOUNT OF ACTIVITY H

H NA 7

8 9

10 11 PERSONNEL EXPOSURES NUMBEA TYPE

DESCRIPTION

~o~00 0

5 NA 7

89 11 12 13 PERS'ONNEL INJURIES 7

89 11 12 METHOD OF DISCOVERY c

44 45 46 44 45 DISCOVERY DESCRIPTION Eddy current of tubes LOCATION OF RELEASE NA 80 80 80 80 80 Q~g 7

89 LOSS OR DAMAGE TO FACILITY TYPE

DESCRIPTION

[sic]

s 7

89 10 PUBLICITY 7

89 AOOITIONAL FACTORS NA 80 80 80 7

8 S 80 7

89 L. S. Lan 80 E.716/546-2700, ext. 291-250 G>0 882 667

I I

~ ~

Ginna Station Steam Generator Ins ection Hay 11, 1977 Rochester Gas and Electric Corporation performed a planned inspection of Ginna Station steam generators from April 16, 1977 through April 23, 1977 in accordance with the Inservice Inspection Program as part of the annual refueling and maintenance outage.

This inspection consisted of eddy current examinations at 400 KHz and 100 KHz to detect defects from corrosion or cracking, 400 KHz for dent evaluation, 25 KHz for sludge profiling and 3.5 KHz to verify support plate integrity.

The inspection included the following:

L~e A-Inlet Tubes 195 1731 148 125 100 207 60 25 KHz 400 KHz 400 KHz 400 KHz 400 KHz 3.5 KHz 100 KHz Concern Sludge Defect Defect Defect Dent Supports Defect

~Su est 1st 1st 6th 6th 6th U-bend A-Outlet 195 268 100 207 25 KHz 400 KHz 400 KHz 3.5 KHz Sludge Defect Dent Supports 1st 6th 6th 6th

Lese B-Inlet Tubes 195 1252 148 125 70 100 60 25 KHz 400 KHz 400 KHz 400 KHz 400 KHz 3 '

KHz 100 KHz Concern

'Sludge Defect Defect

Defect, Dent Support Defect

~su ore 1st 1st 6th U-bend 6th 6th U-bend B-Outlet 195 25 KHz Sludge 1st.

208 400 KHz Defect 6th 100 400 KHz Dent 6th 100 3.5 KHz Support 6th Results of these examinations are given in Figures 1 through 3 which includes the "A" inlet and the "B" inlet and outlet, respectively.

Tables (1) and (2) of this report are included for comparison of the last five steam generator inspections.

All of the eddy current. indications were within the first few inches of tubing directly above the tube sheet with the ex-ception of one in the U-bend area of the "B" steam generator.

This indication in the "B" steam generator U-bend area was on a periphery tube and appears to be a one-of-a-kind construction-type defect of very small volume, 95% through the tube wall.

The other, indications with the exception of two, are postulated to be due to wastage, based on growth rates.

The excepted two

tubes, in the "A" inlet, due to their large growth rate compared to the mean, may have been caused by concentrations of caustics.

Confirmation of this is not possible since the eddy current examination method cannot differentiate between indications resulting from either wastage or stress corrosion cracking, but.

can accurately measure the maximum defect penetration to within nozmal statistical variation.

The indications are seen only in those areas where indications have been previously noted, with the exception of the indication in the "B" steam generator U-bend area.

There has been no expansion of the indication region to other areas of the steam generators.

This inspection verified that. there are not any other tubes with ID Indication in the periphery tubes associated with the wedge areas where there had been a leak in April 1976 (see Licensee Event Report 76-15).

The dent evaluation program performed on both steam generators revealed no dents above 10 mils with only a few of the tubes in each generator being involved.

Along with the dent evaluation b

program, a support plate integrity examinati'on was performed which did not reveal any problems.

The small number of tubes in the "A" inlet which have experienced deterioration, inconsistent with the average, are believed to have deteriorated from the concentration of residual phosphates in the secondary side sludge deposits.

These concentrations of phosphates are caused by the remaining traces of sludge deposits formed during the period of phosphate control of the secondary system water chemistry before the conversion to AVT chemistry control in November 1974.

Removal of residual phosphates from the secondary side over the past 31 months of operation has been accomplished= by continuous steam generator blowdown and high pressure water lancing.

The high pressure water lancing performed on the secondary side is designed to remove as much of the sludge as possible, which contains the undesirable residual phosphates and/or caustics.

Blowdown samples taken during normal operation indicated only

.small amounts of phosphates

present, although phosphates in the 4

sludge could revert back into PO : and concentrate on the tube surfaces which, depending on the molar ratio, would result in acidic or caustic attack of the tube.

The corrective action taken to ensure the continued reliability of the steam generators includes the following:

a.

All tubes with eddy current indications of wall penetra-tion greater than or equal to 40% were plugged.

b.

A thorough lancing of the secondary side of the tube sheets was performed in both steam generators to remove as much as possible any remaining phosphates and/or caustics con tained within the sludge.

Sludge lancing of both steam generators will be continued in an effort to keep sludge content, to a minimum.

The lancing, coupled with blowdown during startup and normal operation, should considerably reduce the probability of significant tube degradation during the plants'ubsequent operation.

c ~

0

~

In addition, a modification of the plant's secondary condensate system is under construction which is de-signed to insure that the feedwater entering the steam generators will be of the highest purity.

This modifi-cation will add in-line demineralizers to the condensate

system, and is scheduled to be placed in service in

/

August of this year.

Because it, has been established that all but 14 tubes from both steam generators had less than 40$ defect indications; because there were only a few tubes which experienced comparatively rapid degradation; and because sludge lancing should further reduce the probability of phosphate and caustic pockets forming, the steam generators are considered acceptable for uninterrupted use until the planned refueling outage in the Spring of 1978, approximately 1 year from the expected date of return to service of Ginna Station.

An eddy current examination of the steam generators in accordance with the Inservice Inspection Program shall be performed during the 1978 refueling outage.

TABLE (1)

STEAM GENERATOR XNDXCATION POSITION SIZE(%)

FEB.

1974 NOV. 1974 EXAMINATIONDATE MAR. 1975 FEB.

1976 APR.

1977 IIAIl INLET 20 20-24 25-29 30-34 35-39 40-44 45-49

> 50 329 63 50 36 14 24 12 17*

631 59 46 31 25 14 52*

655 109 63 38 27 22*

14*

10*

230 59 47 50 31 19*

8*

12*

730 39 37 23 8

0ll*

2*

Total "A" Inlet Tubes Examined 3260 At Each Inspection at 40014HZ 2174 3192 2003 STEAM GENERATOR INDICATXON POSITION SIZE(S)

~ 't FEB.

1974 EXAMINATIONDATE NOV;.1974 MAR. 1975 FEB.

1976 APR.

1977 IIAll OUTLET 20 20 58 0

278 0

10 0

113 0

Total "A" Outlet Tubes Examined At Each Inspection at, 400KHZ 430 442 3192.

268 tubes were explosively plugged

STEAM GENERATOR INDICATION POSITION SIZE(

)

FHB., 1974 NOV. 1974 EXAMINATION DATE MAR. 1975 FEB.

1976 APR.

1977 IIBII INLET 20 20-24 25-29 30-34

, 35-39 40-44 45-49 50 21 4

2 0

0 0

0 0

490 3

4

~

1 1

0 0

0 411 13 10 9

5 1*

0 10*

764 25 8

9 31*

0 1*

719 12 8

0 1

0 0

Total "B" Inlet Tubes Examined 1098 At Each Inspection at 400 KHZ 675 1931 3247 1525 STEAM GENERATOR INDICATION POSITION SIZE (8)

FEB.

1974 EXAMINATIONDATE NOV. 1974 MAR. 1975 FEB.

1976 APR.

1977 IIB II OUTLET 20 20 0

0 1003 2

90]*

Total "B". Outlet Tubes Examined 516 At Each Inspection at 400 KIIZ 39 442 3247 268 NOTE:

Two tubes in the "B" steam generator were explosively plugged in January 1976 and fifteen in April 76 TUBES WHRH EXPLOSIVELY PLUGGED

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