ML20059H080
| ML20059H080 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/26/1993 |
| From: | Grimsley D NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Blackwell M AFFILIATION NOT ASSIGNED |
| References | |
| FOIA-93-195, RTR-NUREG-0844, RTR-NUREG-844 NUDOCS 9401260241 | |
| Download: ML20059H080 (4) | |
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FOIA 19s 4.
AESPONSE TYPE j
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RESPONSE TO FREEDOM OF x I n~^t l I r^ailat i
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INFORMATION ACT (FOIA) REQUEST o^26
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IQY 2 6 E DOCKE T NUMBL Rt$3 utspp/matWe/
HEOUE51CR Ms. Margo Blackwell PART 1.-AGENCY RECORDS RELE ASED OR NOT LOCATED /See checAed bo>es/
No agency ret ords subject to the request have been located No additional agency records >utyect to the request have been located.
Requested ruords are asadable through another public d;stribution program. See Comments section.
Agency records subject to the request th6t are identified in Append.v f es)
A are already avadable for pubhc inspection and copying at the X
NGC PubHc Document Room,2120 L Street N.W., Washington, DC.
Agency records subject to the request that are identified in Appendm(es) 8 are being made avadable for pubbc inspection and copying X
at the NRC Pubhc Document Room. 2120 L Street, N.W., Washington, DC, in a folder under this FOI A number.
The nonproprietary version of the proposaHs) that you agreed to accept in a telephone conversation with a rnember of my staff is now being made available for pubhc inspection and copying at the N RC Pubhc Document Room,2120 L Street, N W., Washington, DC, in a folder under this FOI A number.
Agency records subject to t e request that are identified in Appendix (es) may be inspected and copied at the NRC Local Pubhc Document h
Room identified in the Comments section.
Enclosed is information on how you may obtain access to and the charrs for copying records located at the NRC Pubhc Document Room,2120 L Street, X
N.W. Washington, DC C
y Agency records subject to the rrquest are enclosed.
Records sub;nt to the request have been referred to another Federal arncybes) for reoew and direct response to you, Fees y
You wdl be bSed by the NRC for fees totahng $ 20.40 You wM rece:ve a refund from the NRC in the amount of $
In view of N RC's response to thrs request, no further action is being taa en on appeal letter dated
, No.
PART 11 A-INFORMATION WITHHELD F ROM PUBLIC DISCLOSURE Certain information in the reavested records is being withheld from pubhc disclosure pursuant to the e =emptions described in and for the reasons ststed in Part II, B, C, and D. Arw reteased portions of the documents for which only part of the record is being *sthheld are being rnade availab!e for pubhc inspection and copying in the NRC Pubhc Document Room,2120 L Street, N W., Washington, DC in a fcider under this F Ol A number.
COMM E NTS
- NRC staff has sta'ted that records relating to the subject of your request, " safety of steam generators,* could pertain to thousands of records and would involve a great deal of staf f ef f ort resulting in a large amount of money to process your request.
The subject of your request has been addressed by records that are routinely made part of the public record.
For your information, the NRC routinely makes copies of records exchanged between the NRC and its nuclear power plant licensees, and final major pertinent records regarding the licensed nuclear power plants, available for inspection free of charge and copying at nominal charges in docket files at the NRC's Public Document Room (PDR) in Washington, DC, and at the Local Public Document Rooms (LPDR's) located near the sites of the plants.
The LPDR for the Trojan plant is located at the Branford Price Miller Library, Portland State University, Portland, OR, (503) 725-4735.
SgAt UR E, DlH E CT O, MSIONg RE EDOM Of M ORMAllON AND FUBUCATION5 SE RVICES LW
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BLACKWE93-195 PDR
@ T "t * $ M D R R A $ JB Irt
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.... The records identified on enclosed Appendices A and B are subject to your request.
Copies of these records are enclosed.
Since you are entitled to receive 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of search and 100 pages of records at no charge, you are only being charged $20. 40 f or the duplication of 102 pages of these records.
In addition to these records, a-report entitled, NUREG-0844 - 'NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity,' may answer some of your concerns as this report addresses steam generator tube rupture issues.
You may obtain a copy of this report by contacting the address below:
The National Technical Information Service Springfield, VA 22161 Telephone:
(703) 487-4650 Also identified on Appendix A,
is a printout listing Safety Evaluation Reports and Systematic Assessment of Licensee Performance Reports maintained in the docket file for the Trojan plant.
After your review of this printout, you may obtain copies of any of these records you feel may be of interest to you by contacting the PDR directly.
I have enclosed inf ormation to assist you in ordering records from the PDR.
This completes NRC's action on your FOIA request.
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Re:
FOIA-93-195 APPENDIX A DOCUMENTS MAINTAINED AT THE PDR HUMBER DATE DESCRIPTION 1.
12/16/92 SECY-92-412, " Trojan Steam Generator Issues," Acc. No. 9212290328.
Printout Listing of SER and SALP reports 2.
for Trojan docket.
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Re:
FOIA-93-195 APPENDIX B
RECORDS MAINTAINED IN THE PDR UNDER THE ABOVE REQUEST NUMBER HUMBER DATE DESCRIf(ION 1.
11/92 Report entitled, " Examination of Trojan Steam Generator Tubes, Vol. 3:
Rockwell, Auger, and XPS Analyses,"-
(66 pgs.).
2.
1/18/93 Report entitled, " Safety Review of Trojan Plant Restart:
Steam Generator Deterioration and Interim Plugging
- Criteria, (18 pgs.).
60- @
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Keywords:
EPRI TR-101427 Nuclear stearn generators Volume 3 Intergranular corrosion Project S413-04 Electric Power intergranular stress corrosion cracking Final Report Research Institute inconel alloys November 1992 s
Examination of Trojan Steam Generator Tubes
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Volume 3: Rockwell, Auger, and XPS j
Analyses i
I Prepared by i
I ROCKWELL INTERNATIONAL, Thousand Oaks, California k7 h
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M ATERI A L l
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(continued on back cover)
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REPORT S
U M
'M A
R Y
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Examination of Trojan Steam Generator Tubes Volumes 1-3 I
Examination of 10 tubes removed from Trojan steam generators characterized the depth and type of defects associated with eddy-l!
current signals originating at the tube support plate (TSP) locations.
The TSP indications were associated with 49-89% through-wall l
secondary-side intergranular degradation confined within the support.
Burst pressures of these degraded TSP locations exceeded regulatory l
requirements.
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BACKGROUND The Trojan nuclear power plant is a Westinghouse-designed INTEREST CATEGORIES four loop PWR that began commercial operation in 1976. Secondary-side-initiated pitting attack of steam generator tubing was identified in 1986. Plant changes-i such as discontinued condensate polisher operation and increased steam genera-Steam generat0r reliability tor blowdown-reduced the ingress of copper, chloride, and sulfur impurities Nuclear plant hfe extension influencing the acid-driven pitting degradation. In addition, blowdown water was
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Nuclear plant Operations treated with a deep-bed demineralizer and returned to the system; secondary l.
and maintenance water pH was raised from 8.5 to 9.2. Secondary side axial-crack-like eddy-current lh indications at TSP intersections were first detected in 1988. The number of indica-KEYWORDS tions increased. In 1991, a total of 10 tubes. including 24 TSP intersections, were removed for destructive examination.
Nuclear steam generators
,,l' intergranular corr 0Sion OBJECTIVES To characterize inside diameter (ID)-initiated defects at the tube-Intergranular Stress sheet top; to define the relationship between defects and tube properties; to char-i Corr 0Sion Cracking acterize outside diameter (OD) degradation and determine the burst pressures of defects at TSP locations; to evaluate the effectiveness of eddy-current field test inconel alloys techniques; and to characterize deposits.
I l
APPROACH The project team characterized the pulled tubes by visual and dimensional examination, double-and single wall radiography, burst testing, frac-tography, metallography, X ray photoelectron spectroscopy (XPS), and Auger analysis.
RESULTS Investigations revealed the following:
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. No ID-initiated defects were observed in the tube samples examined.
. Burst pressures of degraded TSP regions exceeded the NRC requirement of -
three times the normal operating primary-to-secondary water pressure differential.
Eddy-current testing provided a relatively good indication of burst capability but did not correlate to the depth of degradation.
i
. Metallography within the TSP region showed OD axialintergranular cracks ranging between 49-98% through the wall with patches of intergranular attack.
Up to 0.23 mm of black, hard, tenacious deposits covered the free surface of the tubes. On the basis of secondary water chemistry and nickel / chromium ratios at crack tips, corrosion was likely due to the concentration of caustic species in TSP crev;ces.
EPRI TR 101427s Vols.1-3 Electric Power Research Institute
Volume 1 contrins the results of d structiva ex;mination. Volum3 2 includ:s the cppendixes for th3 destructive examination. Vclume 3 Y
describes the results of the Auger and XPS analyses.
t EPRI PERSPECTIVE Though the conclusion that a caustic environment was responsible for degradation is not without uncertainties,it received i
suppon from MULTEO (EPRI report NP-5561-CCML) and hideout return j
studies that also suggest alkaline-forming tendencies. The alkaline-forming tendency likely originated after the 1987-1988 change in plant operation to remove acid-forming species responsible for pitting attack. The degrada-tion occurred during the 1988-1991 period, despite boric acid addition l
begun in August 1969. It is not clear when initiation occurred-whether during the transitional period from acid forming to alkaline-forming -
l chemistry or during the alkaline period of operation only. Both chemistries have been shown to cause attack in laboratory tests. The fact that detect-b.
able cracks appeared during boric acid operation suggests initiation under acid conditions because boric acid is believed to more effectivery n
inhibit initiation than growth under alkaline chemistries. If cations such as l
sodium are not balanced by anions such as chloride, the product will be an alkaline-producing sodium hydroxide. The need to maintain this bal-ance by molar ratio control criteria is the subject of the Trojan plant's cur,
'l rent water chemistry program and a revision to EPRI's Secondary Water Chemistry Guidelines (report NP-6239).
I PROJECTS I
RPS413-02, RPS413-04 Project Manager: Allan R. McIlree Nuclear Power Division f
Contractors: ABB Combustion Engineering; Rockwellinternational For further information on EPRI research programs, call l-EPRI Technical Information Specialists (415) 855-2411.
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Examination of Trojan Steam Generator Tubes a
i Volume 3: Rockwell, Auger, and XPS Analyses I
TR 101427, Volume 3 a
Research Project S413-04 Final Report, November 1992 9.
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Prepared by ROCKWELL INTERNATIONAL Science Center f
1049 Camino Dos Rios Thousand Oaks, Cakfornia 91358 1
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f Prepared for Portland General Electric Trojan Nuclear Plant 71760 Columbia River Highway i
Rainier, Oregon 97048
.l Electric Power Research Institute 3
3412 Hillview Avenue j
Palo Alto, California 94304 f-EPRI Project Manager A. R. McIlree.
l-Steam Generator Reliabihty Program Nuclear Power Division
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DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS REPORT WAS PREPARED BY THE ORGAN 2ATION(S) NAMED BELOW AS AN ACCOUNT OF WDRK SPONSORED OR COSPONSORED
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a Price: $10,000.00 (Available only as a 3. volume set.)
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f Electnc Power Research Institute arid EPRI are registered service rnarks of Electric Power Research institute, Inc.
Copyright ? 1992 Electric Power Research Institute. inc. All rights resewed
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EPRILicenced M:t:rl:1 k,
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ABSTRACT l
Corrosion products were examined on tubing from tube / tube support plate intersections from steam generators C and D of the Trojan 1
Nuclear Station.
Tubes R29C70, R30C64, R16C74, R20C66 and R12C70 were removed in 1991 after eddy current indications of IGSCC.
The in-depth composition profiles of the films on the OD and fracture faces of R29C70, R30C64, R16C74, and R12C70 had Cr levels less l
than that in the alloy when compared with Fe and Ni.
Laboratcry data and thermodynamic considerations, while somewhat limited in scope, suggest that these low Cr levels correlate with a highly alkaline or caustic environment.
Tube R20C66, which was plugged in 1990 but removed in 1991, had Cr levels in the corrosien products which indicate a neutral or alkaline crevice; however, the data from R20C66 had considerably more scatter than that from the other tubes.
Films of corrosion products were also examined on Tube R25C58, which was removed in 1986.
This tube had a film 1
chemistry which suggested the crevice pH was less than that experienced by the tubes removed in 1991.
No cracks were present in R25058.
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CONTENTS j
Section page r
1 INTRODUCTION 1-1 2
METHODOLOGY 2-1 3
RESULTS 3-1 4
DISCUSSION 4-1 5
CO!;CLUSIO!:S 5-1 I
6 REEERENCES 6-1 h-6 I
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EPRILic:nsed Material 6-l' i
d tj ILLUSTRATIONS l
Ficure f.A22 2-1 Optical Photomicrographs of Tube R25C58 Showing I
(a) Cross Sectional View and (b) View of OD (4X) 2-2 j
3-1 SEM Photomicrograph of the Area of Tube R25C58
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Analyzed by AES 3-3 3-2 Auger In-depth Composition Profile of the Film on the OD of Tube R25C58 at the Tube / tube Support Plate Intersection Showing the Alloy Metals Normalized to 100%
3-3 3-3 As-received Section of Tube R29C70 Showing the Points A-L Where an Analysis of the Surface was 3
Made by AES 3-4 3-4 SEM Photomicrograph of the Crack Which was Analyzed in Tube R29C70 3-6 3-5 SEM Photomicrograph of the Fracture Face of the Crack Shown in Figure 3-4 (Tube R29C70) with the A: as Analyzed Indicated 3-6 3-6 SEM Photomicrograph of the Center of Region A Near the Crack Mouth in Figure 3-5 (Tube R29C70) 3-7 i
3-7 SEM Photomicrograph of the Center of Region B in Figure 3-5 (Tube R29C70) 3-7 3-8 SEM Photemicrograph of the Center of Region C in Figures 3-5 (Tube R29C70) 3-8 3-9 SEM Photomicrograph of the Center of Region D Near the Crack Tip in Figure 3-5 (Tube R29C70) 3-8 i
l 3-10 Optical Photomicrograph of As-Received Section of Tube R30C64.
The Arrows Indicate The Location of the Crack Analyzed 3-13 3-11 SEM Photomicrograph Showing the Fracture Face of the Crack Analyzed in Tube R30C64, Where Regions
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A, B,
C are the Individual Grains Analyzed 3-14 i'
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EPRILicensed M:t:ri:I Ficure Page 3-12 SEM Photomicrographs Showing the Grains Analyzed
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on the Fracture Face of Tube R30C64 in (a) Regions A and B (b) Region C 3-15 3-13 Optical Photomicrograph (6X) of the Section of Tubing Form Tube R16C74 Received Showing the Crack Examined and the Initial Cut Made to Extract a
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Section for Analysis 3-17 i
3-14 SEM Photomicrograph of the Section of Tube R16C74 Received Showing the Extensive IGA
?-2c 3-15 SEM Photomicrograph of the Fracture Face of R16C74 Examined Showing the Grains Analyzed 3-19 3-16 SEM Photomicrograph of a Typical Region Near the OD on the Fracture Face of R16C74 Examined 3-19 3-17 SEM Photomicrograph (10X) of As-Received Section of Tubing R12C70 showing the OD surface 3-22 3-18 SEM Photomicrograph of As-received Section of Tubing R12C70 Showing the Fracture Face Analyzed 3-23 3-19 SEM Photomicrograph Showing the Grains Profiled on the Fracture Face of R12C70 3-23 3-20 SEM Photomicrograph Showing Typical Region of the Fracture Face of R12C70 in High Magnification 3-24 3-21 Optical Photomicrograph (6X) of As-Received Section of Tubing from R20C66.
The Cut in the Section was Made After Receipt from ABB/ Combustion Engineering 3-26 3-22 SEM Photomicrograph of the CD of Tube R20C66 Showing the Shallow IGA 3-27 e
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rigure page i
3-23 SEM Photomicrograph Showing the Grains Analyzed on the Fracture Face of a Crack in Tube R20C66, Where A is near the Crack Mouth, D is Near the Crack Tip and Grains B and C are Between the
-l Crack Mouth and Crack Tip 3-30 3-24 SEM Photomicrograph of As-received Section from Tube R12C8 3-35 3-25 SEM Photomicrograph of the Fracture Face Analyzed from Tube R12C8 3-36 4-1 Potential-pH Diagram for the re-water System (at
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288 C) with Dissolved Species Activities of 10-3 [3]
4-2 4-2 Potential-pH Diagram for the Cr-water System (at 288 C) with Dissolved Species Activities of 10-3 [3]
4_3 4-3 Potential-pH Diagram for the Ni-water System (at 288'C) with Dissolved Species Activities of 10-3 [3) 4_4 i
4-4 Auger In-Depth Composition Profile for Alloy 600 Which was Exposed to 50% NaOH at 320 C 4-5 4-5 Auger In-Depth Composition Profile for Alloy 600 Which Had Been Exposed to 50% NaOH at 320*C While Polarized 150 mV Above the Corrosion Potential, E
4-5 corr 4-6 Change in Fe, Cr, and Ni Composition of the Surface Film on Alloy 600 After Exposure to H SO /Na2 SO /NaOH Solutions of Varying pH's 2
4 4
( 2 8 0'C ) [2]
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~l TABLES Table Page 2-1 List of Tubes and Tube Support Plate (TSP) Inter-section Where the Tubing Examined Was Removed 2-1 3-1 Auger In-Depth Composition Profile from a Region i
i on the OD of Tube R25C58 3-2 3-2 Auger Analysis of Points Indicated as E-S on the Surface of Tube R29C70 3-4 3-3 Auger In-Depth Composition Profile from the Region Centered at Point D on the Surface of Tube R29C7C (Figure 3-3) 3-5 3-4 Auger In-Depth Composition Profile of Region A (Crack Mouth) in Figure 3-5 (Tube R29070) 3-9 3-5 Auger In-Depth Composition Profile of Region B on the Fracture Face in Figure 3-5 (Tube R29C70) 3-10 3-6 Auger In-Depth Composition Profile of Region C on the Fracture Face in Figure 3-6 (Tube R2 9C7 0) 3-11 3-7 Auger In-Depth Composition Profile of Region D (Crack Tip) in Figure 3-7 (Tube R29C70) 3-12 3-8 Auger In-Depth Composition Profile on a Region on the CD of Tube R30C64 3-14 3-9 Auger In-Depth Composition Profile of Region A (Crack Mouth) on Fracture Face in Tube R30C64 Shown in Figure 3-12(a) 3-16 3-10 Auger In-Depth Composition Profile of Region B (Mid-Point on Fracture Face in Tube R30C64) in Figure 3-12 (a) 3-16 3-11 Auger In-Depth Composition Profile of Region C j
(Crack Tip) on Fracture Face in Tube R30C64 Shown in Figure 3-12(b) 3-16
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3-12 Auger In-Depth Profile of a Region on the OD of Tube R16074 3-18 j
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EPRIUc:nsed M:t:rlal
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Tab 2e F.a.gr 3-13 Auger In-Depth Composition Profile of Region A i
(Crack Mouth) on Fracture Face of Tube R16Cl4 Shown in Figure 3-15 3-20 3-14 Auger In-Depth Composition Profile of Region B (Mid-Point on Fracture Face of Tube R16C74) in Figure 3-15 3-20 3-15 Auger In-Depth Combustion Profile of Region C (Crack Tip) on Fracture Face of Tube R16C74 Shown in Figure 3-15 3-21 3-16 Auger In-Depth Composition Profile of a Region on the OD of Tube R12C70 3-22 3-17 Auger In-Depth Composition Profile of Region A (Crack Mouth) on Fracture Face of Tube R12C70 Shown in Figure 3-19 3-25 3-18 Auger In-Depth Composition Profile of Region B (Crack Tip) on Fracture Face of Tube R12C70 Shown in Figure 3-19 3-25 3-19 Auger In-Depth Composition Profile of the film on the OD of Tube R20C66 Between IGA 3-27 3-20 Auger In Depth Composition Profile of a Region Inside a Crack Caused by IGA (Tube R20C66) 3-28 3-21 Auger In-Depth Composition Profile of Grain A (Crack Mouth) on Fracture Face of Tube R20C66 in Figure 3-23 3-31 3-22 Auger In-Depth Composition Profile of Grain B in Figure 3-23 (Tube R20C66) 3-32 3-23 Auger In-Depth Composition Profile of Grain C in Figure 3-23 (Tube R20C66) 3-33 3-24 Auger In-Depth Composition Profile of Grain D (Crack Tip) on Fracture Face of Tube R20C66 in Figure 3-23 3-34 h
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EPRILic;nred M;t;ri:1 Table Pace
'4 3-25 Auger In-Depth Composition Profile of the OD of
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Tube R12C8 3-35 3-26 Auger In-Depth Composition Profile of the Fracture Face of Tube R12C8 3-36 r
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EPRILic::nsed M:terLI t
1-I Section 1 s
INTRODUCTION i
i The chemistry of the film, or corrosion products, on the surface of a metal or alloy is determined by the environment to which a j
metal or alloy was exposed. As the pH, temperature, composition and/or oxidizing conditions of an aqueous environment change, l
different reaction products become thermodynamica11y stable.
Laboratory results obtained from an EPRI funded program at Rockwell [1] and subsequent Japanese work (2] demonstrated that the concentration of alloy metals in the corresion products on alloy 600 change in a predictable manner with pH.
Previous work has used the composition of the corrosion products on the surface of alloy 600 tubing to identify the local crevice environrent in a steam generator which may have caused intergranular stress corrosion cracking (IGSCC) [1.A].
This report presents results of an analysis of the corrosion products on sections of alloy 600 tubing from the tube / tube support plate crevices in the Trojan Nuclear Plant.
This evaluation has been used to hypothesize the crevice chemistry.
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HETHODOLOGY The sections of tubing analyzed were received from ABB/ Combustion i
Engineering.
All sections contained a tube / tube support plate intersection zone.
Table 2-1 identifies the tubes analyzed.
This table also indicates the tube / tube support plate intersection, the steam generator, the year in which the tube was plugged, and when the tube was removed.
Tube R25C58 was received in a metallographic mount.
Figure 2-1 shows a cross sectional and OD view af ter this tube was broken f rom the mount.
The as-received condition of the other samples were sections of tubing approximately 1 cm x 1 cm.
Each of the small sections received had one or more axial cracks which were clearly visible under a
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low power microscope.
All the 1 cm x 1 cm tube secticns received, except that from tube R12C70, had been cut from tubes after they had been burst tested.
A burst test was not performed on tube R12C70 prior to sectioning.
Table 2-1 LIST OF TUBES AND TUBE SUPPORT PLATE (TSP)
INTERSECTION WHERE THE TUBING EXAMINED WAS REMOVED TUBE TSP S/G YEAR PLUGGED REh0/ED R25C58 1
C Not Plugged 1986 R29C70 1
C 1991 5/91 R30C64 1
C 1991 9/91-10/91 R16C74 2
D 1991 11/91 R12C70 2
C 1991 11/91 R20066 1
D 1990 11/91
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R12C8 1
D 1989 5/91 1,
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Figure 2-1.
Optical Photomicrographs of Tube R25C58 l
Showing (a)
Cross Sectional view and (b) View of OD l
i.
.a l
Two surface analysis techniques were used to determine the l
fracture face and OD chemistries, Auger electron spectroscopy f
i (AES) and X-ray photoelectron spectroscopy (XPS).
AES was used to l
obtain in-depth, elemental composition profiles of the films by i
l sequentially sputtering by argon ion bombardment and performing J
]
Auger analysis.
A Perkin-Elmer PHI Model 590 Scanning Auger j
Microprobe was used for the Auger examination.
The analysis was
]l i
performed using an electron beam excitation voltage of 3 Kv, i
l charging was observed at higher voltages in some cases, and a 6 eV l
l modulation voltage.
The OD was analyzed with the electron beam l
rastered 30 pm x 30 m.
Individual gra, ins were analyzed on the I
fracture face in all sections of tubing, except that from tube 1
l R29C70, with a stationary electron beam defocused to a diameter of approximately 10 m.
The electron beam was rastered 100 pm x 100 pm during the analysis of the fracture face of R29C70 i
because the large density of deposits made the fracture face I
composition nonuniform from grain to grain.
Sputtering was i
2-2
EPRILic:n: dM;t:ri:I performed with a 5 KV argon ion beam rastered over an area 2.7 mm x 4 mm, which gave a sputtering rate of approximately 5A/s.
I i
X-ray photoelectron spectroscopy (XPS) was used for compound identification without sputtering since ion bombardment has been j
found to change the valence state and chemical state oc surface species in some cases.
A Perkin-Elmer, PHI Model 548 XPS spectrometer was used for this analysis.
The excitation source for the photoelectrons was the MgKa line.
The diameter of the j
analysis area was several hundred micrometers.
I b
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1 Section 3 RESULTS i
i AES Results
"""'"9;-
Table 3-1 shows the Auger in-depth profile of-tha Since this sample had been removed from a metallographic mount, care was taken to select sections of the OD for analysis where the t
appearance of the oxide and inspection of the sections of mount indicated that the scale was undisturbed during removal.
This section of tubing was examined optically and in the SEM for cracks.
None were found.
Figure 3-1 is an SEM photomicrograph of the area analyzed.
An area 30 m x 30 pm was profiled in the center of Figure 3-1.
Table 3-1 is the in-depth composition profile obtained from this analysis area.
The table is formatted F
such that the last column tc the right gives the depth sputtered, the next three columns to the left give the relative atomic i
percent of the alloy metals normalized so that they add to 1001, and the remaining columns give the atomic percents of all the detected species, which originated from the secondary water.
The oxide was sputtered to a depth of 4 pm.
Zine was present in the oxide at relatively high concentrations, particularly in the first 5000A.
Carbon was high near the surface, but quickly decreased to low levels during sputtering after removing a few hundred Angstroms of film.
Examining the relative amounts of Cr, Fe, and Ni, the Fe concentration is approximately three times that of the sum of Cr and Ni for the first 600A.
Iron then decreases in concentration with depth with a corresponding increase in N1.
This is more clearly indicated in Figure 3-2, which shows graphically the relative changes in the alloy metals.
The concentration of chromium in the oxide is approximately that in 1
1-3-1
EPRILic:n:ed M:t:ri:I A
the alloy for the first 1000A in depth, after which there is a layer approximately 1 pm in width where it is enriched.
Table 3-1 AUGER IN-DEPTH COMPOSITION PROFILE FROM A REGION ON THE OD OF TUBE R25C58 Atomic Percents Normalized Atomic Percents O
Cr Fe Ni C
C3 Zn Cr Fe Ni A
7 41.1 3.8 15.7 2.8 17.3 0.3 19.0 17.0 70.5 12.5 0
,{
44.5 5.1 20.8 2.9 8.0 0.2 18.4 17.8 72.3 10.0 100 44.7 5.8 22.8 2.7 6.5 0.2 17.1 18.6 72.7 8.7 250 46.3 6.3 23.2 3.7 3.6 0.0 16.9 18.8 69.9 11.3 400 47.6 6.7 25.0 3.2 2.5 0.0 14.9 19.2 71.5 9.3 600
.3 46.4 8.3 26.6 4.9 1.2 0.0 12.6 20.9 66.8 12.3 1600 45.8 11.5 25.4 6.9 0.6 0.2 9.8 26.2 58.1 15.7 3200-42.1 12.9 22.5 14.1 0.6 0.0 7.8 26.1 45.4 28.5 6400 32.0 13.2 18.2 30.4 0.7 0.0 5.5 21.3 29.5 49.2 12800 21.0 13.2 13.8 47.9 1.1 0.0 3.1 17.6 18.4 64.0 25600 24.7 11.9 13.5 44.5 1.6 0.0 3.7 17.0 19.3 63.6 35000 19.8 13.4 15.7 50.3 0.7 0.0 0.0 16.9 19.8 63.3 40000 Figure 3-3 shows the as-received section of tube R29C70.
e.n Auger analysis was performed before sputtering at the points indicated by letters A through L.
The results from the points (beam size approximately 1 pm in diameter) which did not charge are given in Table 3-2.
No Cr was observed at five of the nine points, two points had Cr levels, relative to Fe and Ni, considerably below that in the alloy.
At two points, Ni was the only alloy metal detected; whereas, all other points were substantially enriched in Fe relative to that in the alloy.
Of the irnpurities found Si, F and Zn were in the highest concentrations.
The elect ron beam was rastered to give an analysis area of approximately 30 pm x 30 pm, centered at area D, to obtain the in-depth profile in Table 3-3.
Examining the relative amounts of Fe, Cr, and Ni, the concentration of Cr was approximately the same level as in the alloy, and Fe was three to four times higher in concentration than 4
3-2
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SEM Photomicrograph of the Arda of Tube i.
R25C58 Analyzed by AES 1
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1 10 100 1000 10000 100000 i
l DEPTH SPtJTTERED(A) l Figure 3-2.
Auger In-depth Composition Profile of the Film on the OD of Tube R25C58 at the Tube / tube Support Plate Inte rs e ct ion Showing the Alloy Metals Normalized to 100%
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Figure 3-3.
As-received Section of Tube R29C70 Showing l
the Points A-L Where an Analysis of the Surface was Made 1
by AES l
Table 3-2 AUGER ANALYSIS OF POINTS INDICATED AS E-S i
ON THE SURFACE OF TUBE R29C70
[
j AtorrWe Percents Normalized Atomic Percents l
O Cr Fe Ni Si Ca C
0; Zn F
Cr Fe Ni LOCATION l
61.6 0.0 0.0 3.6 11.6 2.2 13.6 0.0 0.0 7.4 0.0 0.0 100.0 A
i 36.5 0.0 3.4 11.0 3.2 0,0 37.9 0.0 0.0 8.0 0.0 23.8 76.2 B
l 50.8 0.0 0.0 4.3 7.2 1.2 26.9 0.0 3.3 6.2 0.0 0.0 100.0 C
j 32 7 J.0 2.6 2.0 19.5 0.8 31.6 2.3 2.2 6.4 0.0 56.9 43,1 D
2 3..'
9 16.1 20.5 2.2 1.0 17.2 1.9 5.0 0.0 9.9 32.5 57.6 E
3 5.T
.0 0.0 2.3 6.1 0.0 40.7 1.8 6.5 7.3 0.0 0.0 100.0 F
4 7.f
,5 4.7 4.9 15.5 3.2 12.5 2.1 3.0 3.7 20.6 39.0 40.4 G
3 3.f.
'F 18.8 15.1 7.5 1.8 13.4 0.0 4.9 0.0 14.2 47.6 38.2 H
40.4 10.6 7.7 10.5 6.6 11.9 2.0 2.5 4.0 17.4 48.1 34.6 1
47.6 4.6 7.3 13.1 11.7 0.9 12.5 2.3 0.0 0.0 18.4 29.3 52.3 J
J 57.1 12.3 0.0 11.9 0.0 3.3 3.9 4.0 0.0 7.5 50.8 0.0 49.2 K
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e Table 3-3
} ~l AUGER IN-DEPTH COMPOSITION PROFILE FROM THE REGION CENTERED AT POINT D ON THE SURFACE OF TUBE R29C70 (FIGURE 3-3) i Atomic Percents Normalized Atomic Percents i
O Cr Fe Ni Cu Si Oa C
Cr Fe Ni A
47.5 4.6 7.3 13.1 2.3 11.7 1.0 12.5 18.4 29.3 52.3 0
39.2 4.5 7.2 21.1 5.5 8.3 1.2 13.1 13.8 21.9 64.4 800 34.3 6.7 9.1 26.8 4.1 6.9 1.7 10.4 15.7 21.4 62.9 1000 32.6 8.0 9.1 29.3 3.0 5.6 1.9 10.5 17.3 19.5 63.1 1200 30.7 8.8 9.3 31.3 3.5 4.7 2.2 9.4 17.8 18.9 63.3 1400 29.2 8.9 10.1 33.3 3.2 4.1 2.4 8.8 17.0 19.4 63.6 1600 27.7 10.6 9.6 35.1 3.1 3.7 2.2 7.9 19.2 17.4 63.4 2000 26.5 11.2 10.1 38.2 2.5 3.0 2.0 6.4 18.8 17.0 64.2 2600 26.7 12.5 10.0 39.8 2.2 2.1 1.6 5.0 20.0 16.1 63.9 3400 27.3 13.4 9.9 40.5 2.3 1.4 1.5 3.7 21.1 15.5 63.5 4600 26.2 14.3 10.0 42.7 2.2 0.6 1.3 2.7 21.3 14.9 63.7 6000 i
26.0 14.0 9.0 45.5 1.5 0.6 1.2 2.2 20.4 13.2 66.4 9000 25.2 14.1 8.8 47.6 1.1 0.3 0.8 2.1 20.0 12.5 67.5 15000 24.2 14.7 9.7 47.4 1.7 0.2 0.5 1.6 20.5 13.5 66.1 17000 that found in Alloy 600.
Both Cu and Si were in high concentrations.
The fracture face of a totally intergranular crack approximately 60% through-wall, shown in Figure 3-4, was examined.
Four areas designated A, B,
C, and D were analyzed as indicated in the SEM photomicrograph in Figure 3-5.
The beam was rastered so that areas 100 pm x 100 pm were analyzed.
This approach differs from all other fracture face examinations irt which individual grains were examined.
Figures 3-6 through 3-9 are SEM photomicrographs of the center of the analysis areas.
There are numerous deposits on the grains, the density of which decreases progressing from the crack mouth to the crack tip.
The Auger in-depth profiles are i.
given in Tables 3-4 through 3-7.
Silicon was found in concentrations as high as 8.0 a/o at the crack mouth (area A) and 3-5 l
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Figure 3-4.
SEM Photomicrograph of the Crack Which Was l,
Analyzed in Tube R29C70 l
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SEM Photomicrograph of the Fracture Face of i
the Crack Shown in Figure 3-4 (Tube R29C70) with the Areas Analyzed Indicated I
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SEM Photomicrograph of the Center of Region A 4,
Near the Crack Mouth in Figure 3-5 (Tube R29C70) 4* a i
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Figure 3-7.
SEM Photomicrograph of the Center of Region B in Figure 3-5 (Tube R29C70) l l
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SEM Photomicrograph of the Center of Region C j
in Figure 3-5 (Tube R29C70) l' l
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Figure 3-9.
SEM Photomicrograph of the Center of Region D Near the Crack Tip in Figure 3-5 (Tube R29C70)
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.I Table 3-4 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION A J
(CRACK ' HOUTH)
IN FIGURE 3-5 (TUBE R29C70) 1 Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Si C
F Cr Fe Ni A
l 31.4 1.8 0.0 21.8 5.6 15.7 2.2 7.8 0.0 92.2 0
33.8 1.9 0.0 23.1 6.4 10.7 1.7 7.5 0.0 92.5 50 33.3 1.7 0.0 24.8 7.1 8.7 1.5 6.4 0.0 93.6 100 31.5 1.9 3.2 28.4 7.0 6.9 0.0 5.7 9.5 84.8 200-30.7 2.0 3.2 30.3 7.2 5.8 0.0 5.6 9.0 85.4 300 31.2 2.0 3.3 30.7 7.2 5.1 0.0 5.7 9.1 85.3 400 l
30.1 2.0 2.7 31.4 8.0 4.3 0.0 5.7 7.4 86.9 500 30.0 2.3 3.4 33.1 6.7 4.4 0.0 5.9 8.7 -85.4 700 27.4 3.6 3.4 39.4 6.1 3.0 0.0 7.7 7.4 84.9 1650 25.1 4.9 3.7 43.2 5.6 2.7 0.0 9.5 7.1 83.5 3000 23.8 5.7 4.4 45.9 5.4 2.2 0.0 10.2 7.8 82.0 5000 21.9 6.6 5.1 49.7 5.2 1.8 0.0 10.8-8.3 80.9 7500 20.6 7.6 6.9 49.9 4.9 1.5 0.0 11.8 10.7 77.5 10000 16.8 9.2 7.0 56.9 3.5 1.6 0.0 12.6 9.5 77.8 20000 13.7 10.4 7.7 60.4 2.3 1.9 0.0 13.3 9.8 76.9 30000 12.5 10.4 8.3 62.0 1.6 2.7 0.0 12.9 10.3 76.8 40000 11.3 10.9 8.4 63.8 1.1 2.6 0.0 13.1 10.1 76.8 50000 10.0 11.2 8.1 66.4 0.5 2.1 0.0 13.1 9.5 77.4 60000 l
j 3-9
EPRILicen:ed M:t:ri:I
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Table 3-5 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION B ON THE FRACTURE FACE IN FIGURE 3-5 (TUBE R29C70)
Atomic Percents Normalized Atomic j
Percents O
Cr R
Ni F
Si C
Cr R
Ni A
36.8 3.2 0.0 30.0 4.1 1.3 24.6 9.6 0.0 90.4 0
39.3 4.0 0.0 34.3 2.5 1.9 18.0 10.4 0.0 89.6 50 40.0 3.7 0.0 37.0 2.1 2.4 14.7 9.0 0.0 91.0 100 31.5 3.4 0.0 51.8 1.2 2.1 10.0 6.2 0.0 93.8 200 38.4 4.2 1.9 41,0 1.6 2.5 10.4 8.9 4.1 87.0 300 36.7 4.2 3.9 42.9 1.2 2.4 8.7 8.3 7.6 84.1 400 36.4 4.1 4.1 42.7 1.6 3.1 8.2 8.0 8.0 84.0 500 34.3 6.2 4.1 45.7 1.6 2.6 5.6 11.1 7.3 81.6 700 29.7 5.9 4.3 52.1 0.5 2.4 5.2 9.4 6.8 83.7 1650 24.8 9.2 5.0 54.0' O.0 2.3 4.6 13.5 7.3 79.1 3000 23.3 10.3 7,4 53.1 0.0 2.5 3.4 14.6 10.5 75.0 5000 20.1 9.5 6.0 59.1 0.0 2.3 3.1 12.7 8.0 79.3 7500
~
17.9 12.1 7.0 57.7 0.0 2.2 3.i 15.8 9.1 75.1 1000 14.6 14.3 7.2 59.8 0.0 1.6 2.5 17.5 8.9 73.6 20000 11.9 11.2 7.7 64.7 0.0 0.8 3.7 13.4 9.2 77.5 30000 10.8 11.4 7.6 64.7 0.0 0.3 5.3 13.6 9.0 77.4 4C000 10.0 11.3 8.3 66.1 0.0 0.2 4.2 13.2 9.6 77.2 50000 9.4 10.7 8.4 67.4 0.0 0.1 3.9 12.4 9.7 77.9 60000 9
3-10 l
EPRILicen:ed Mat:ri:1 Ii.
Table 3-6 AUGER IN-DEPTH. COMPOSITION PROFILE OF REGION C 4'
ON THE FRACTURE FACE IN FIGURE 3-6 (TUBE R29C70) 2 Atomic Percents Normalized Atomic I
Percents O
Cr Fe Ni F
Si C
Cr Fe Ni A
k 32.0 3.1 0.0 34.3 4.6 0.0 26.1 8.2 0.0 91.8 0
33.0 3.4 0.0 38.5 3.7 0.0 21.4 8.1 0.0 91.9 50 33.5 4.0 0.0 40.8 2.6 0.0 19.1 9.0 0.0 91.0 100 1
33.3 4.2 0.0 44.3 2.3 0.0 16.0 8.6 0.0 91.4 200 33.4 4.3 0.0 47.6 1.9 0.0 12.7 8.4 0.0 91.6 300 32.5 4.5 0.0 50.0 1.9 0.0 11.1 8.3 0.0 91.7 400 30.5 4.6 3.6 48.9 1.8 0.0 10.6 8.0 6.4 85.6 500 29.8 4.7 3.6 52.1 0.8 0.0 9.0 7.8 6.0 86.2 700 23.5 6.6 4.5 58.6 0.5 0.0 6.3 9.4 6.5 84.1 1650 18.9 8.9 6.6 60.8 0.2 0.0 4.5 11.7 8.6 79.7 3000 16.8 9.7 6.8 62.0 0.0 0.0 3.9 12.2 8.5 79.3 6000 14.4 10.6 6.9 64.9 0.0 0.0 3.1 12.9 8.4 78.7 9000 13.4 11.0 7.8 64.2 0.0 0.0 3.6 13.3 9.4 77.3 10000 10.0 11.4 7.8 67.3 0.0 0.0 3.5 13.2 9.0 77.8 20000 l.
8.3 10.6 8.0 68.7 0.0 0.0 4.3 12.2 9.1 78.7 30000 7.7 10.5 8.0 69.3 0.0 0.0 4.4 11.9 9.2 78.9 40000 7.5 10.0 8.4 69.8 0.0 0.0 4.4 11.3 9.5 79.2 50000 7.3 10.4 9.1 69.0 0.0 0.0 4.2 11.7 10.3 78.0 60000 e
e 3-11
EPRILicensed MaterlOI l
Table 3-7 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION D (CRACK TIP)
IN TIGURE 3-7 (TUBE R29C70)
Atomic Percents Normalized Atomic Percents O
Cr R
Ni F
Si C
Cr Ni A
28.0 3.7 0.0 42.0 3.6 0.0 22.8 8.0 0.0 92.0 0
26.3 4.7 5.2 51.3 1.6 0.0 11.0 7.7 8.4 83.9 50 19.1 5.3 6.2 58.8 1.1 0.0 9.6 7.5 8.9 83.7 100 7
l 17.6 5.1 7.0 63.9 0.0 0.0 6.4 6.8 9.2 84.1 200 16.1 5.5 7.1 64.5 0.0 0.0 6.8 7.1 9.2 83.7 300 15.5 5.9 6.8 65.9 0.0 0.0 5.9 7.5 8.7 83.9 400 15.1 5.6 7.5 64.7 0.0 0.0 7.1 7.2 9.7 83.1 500 q
13.4 6.6 8.0 67.1 0.0 0.0 5.0 8.1 9.8 82.2 700 13.5 7.5 8.7 65.9 0.0 0.0 4.4 9.1 10.6 80.3 1650 9.4 9.2 8.7 69.9 0.0 0.0 2.8 10.5 9.9 79.6 3000 i
8.7 9.8 9.1 68.6 0.0 0.0 3.8 11.2 10.4 78.4 5000 6.5 10.0 8.9 71.8 0.0 0.0 2.7 11.1 9.8 79.1 7500 7.7 9.5 8.8 70.2 0.0 0.0 3.9 10.7 10.0 79.3 10000 5.2 9.3 10.2 72.4 0.0 0.0 2.9 10.1 11.1 78.8 20000 5.2 9.7 9.0 72.4 0.0 0.0 3.6 10.6 9.9 795 30000 3.6 10.1 10.1 73.6 0.0 0.0 2.5 10.8 10.7 78.5 40000 3.6 9.8 9.7 74.7 0.0 0.0 2.1 10.4 10.3 79.3 50000 4.5 9.8 9.7 73.7 0.0 0.0 2.3 10.5 10.4 79.1 60000 3
at approximately 2 a/o in area B.
No Si was found in the other two areas analyzed, which were closest to the crack tip.
Fluorine was present up to almost 5 a/a in the outer several hundred Angstroms of the film in all four areas analyzed on the fracture face.
Comparing the relative amounts of the alloy metals, all l
areas examined had much lower Cr levels than that in the alloy.
Iron was not present on the surface and in relative concentrations less than that in the alloy for several hundred Angstroms in the
_2 film.
Figure 3-10 shows the as-received condition of tube R30C64 with the crack which was opened for analysis indicated by arrows.
The
~
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i I j
e
-f b
l l 't I1 t
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l Figure 3-10.
Optical Photomicrograph of As-received Section of Tube R30C64.
The Arrows Indicate The Location of the Crack Analyzed i
4 in-depth composition analysis of the OD of this tube section is
- 1 shown in Table 3-8.
The contaminating species in the oxide having i
the highest concentrations are Zn and Si neither of which decrease
]I in concentration to approximately 2 m in depth.
Comparing the relative amounts of alloy metals, the oxide is enriched in Fe and ii had no Cr until 3200A was sputtered.
The highest relative level pf Cr detected, in the 5 pm sputtered, was slightly more than half L
that in the alloy.
Figure 3-11 is an SEM photomicrograph of the j
totally intergranular fracture face profiled.
The grains profiled are shown in Figure 3-12.
Three grains were analyzed: near the crack mouth (A), approximately mid-way between the crack mouth and 1[
crack tip (B), and near the crack tip (C).
There are a few highly j
4 i
t dispersed deposits on the grains immediately adjacent to the OD i
3-13 1
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j Table 3-8 AUGER IN-DEPTH COMPOSITION PROFILE ON A REGION ON THE OD OF TUBE R30C64 3
Atomic Percents Normalized Atomic l
Percents O
Cr Fe Ni Zn Si G
C Cr Fe Ni A
l 45.5 0.0 7.7 1.8 11.7 11.2 0.9 21.1 0.0 81.4 18.6 0
48.3 0.0 8.8 2.3 9.5 20.1 1.3 9.6 0.0 79.2 20.8 100 49.4 0.0 9.8 2.6 10.3 19.0 1.5 7.6 0.0 79.2 20.8 200 l
l 52.7 0.0 13.0 3.0 11.7 9.6 1.8 8.2 0.0 81.2 18.8 400 51.5 0.0 10.2 2.4 9.2 17.4 2.3 7.1 0.0 81.2 18.8 800
)
i 1
45.5 0.0 19.6 3.9 8.6 15.0 2.2 5.1 0.0 83.4 16.6 1600 45.5 0.0 21.8 3.9 7.2 14.9 2.7 4.0 0.0 85.0 15.0 3200 i
I 43.3 2.6 24.3 4.5 6.5 13.0 2.5 3.2 8.3 77.3 14.4 6400 i
43.8 3.5 24.1 6.7 5.4 11.9 2.3 2.4 10.2 70.3 19.5 12800 I
42.4 3.2 22.4 11.3 4.6 11.8 1.9 2.4 8.8 60.7 30.5 25000 t
40.8 5.3 20.1 19.7 2.7 8.5 1.6 1.4 11.7 44.6 43.6 50000 l
l.
f i
l
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Q,fk }' f.,' '
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l YN
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~
l Figure 3-11.
SEM Photomicrograph Showing the Fracture j
Face of the Crack Analyzed in Tube R30C64, Where Regions l
l -
A, B,
C are the Individual Grains Analyzed i
i a
j 3-14 i
1 l
i
EPRILicensed Material p
1 i
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j Figure 3-12.
SEM Photomicrographs Showing the Grains Analyzed on the Fracture Face of Tube R30C64 in i
(a) Regions A and B (b)
Region C 1
l surface.
Otherwise, the fracture face is almost free of deposits.
l l.
The profiles obtained for the grains are given in Tables 3-9 l
through 3-11.
All three areas examined were almost identical in i
composition.
Low amounts of Cu, and Ti were detected.
The llt relative concentration of both Fe and Cr were slightly above those in the alloy.
i l
The as-received condition of R16C74 is shown in Figure 3-13.
The If l
cut through the center of the specimen was made after the sample was received.
There were numerous shallow cracks on the surface j.
(Figure 3-14).
The Auger in-depth profile from an area on the OD
.t ll between cracks is shown in Table 3-12.
Aside from C, the lj impurities in the oxide in the highest concentrations are Si and i I S.
Copper and calcium are present at approximately I a/o.
The first 1600A of the oxide had no Cr.
Af ter sputtering to a depth l
of 2.5 pm Cr enriched to concentrations, relative to Fe and Ni, l
between 20 and 26 a/o.
Iron was considerably enriched relative to i
the percentage in the alloy.
l-3-15 I
I i
(
EPRILicen:ed M:t:ri:1 o
Table 3-9 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION A (CRACK MOUTH)
ON FRACTURE FACE IN TUBE R30C64 SHOWN IN FIGURE 3-12 ( A)
Atomic Percents Normalized Atomic Percents O
Cr Fe Ni CU S
C Ti Cr Fe Ni A
~
i 31.3 12.6 6.4 39.2 0.0 0.0 8.8 1.7 21.7 11.0 67.4 0
29.6 11.8 6.5 35.8 0.7 0.2 14.2 1.3 21.8 12.1 66.2 100 1
29.1 12.0 6.4 38.0 1.3 0.2 12.0 1.1 21.3 11.3 67.4 200 j
28.6 11.4 6.7 40.6 0.6 0.2 11.0 0.9 19.4 11.4 69.2 400 3
23.0 12.4 7.1 48.0 0.0 0.1 8.6 0.7 18.4 10.5 71.1 1000 l
13.0 14.0 7.4 54.9 0.0 0.1 9.9 0.8 18.3 9.7 72.0 2000 8.7 14.7 7.2 53.6 0.0 0.0 15.3 0.6 19.5 9.5 71.0 4000 Table 3-10 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION B (MID-POINT ON FRACTURE FACE IN TUBE R30C64)
IN FIGURE 3-12 (A)
Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Cu S
C Ti Cr Fe Ni A
~
34.6 13.9 4.8 32.2 0.9 0.8 11.7 1.1 27.4 9.3 63.3 0
41.0 14.5 4.1 33.1 0.0 0.5 5.7 1.1 28.0 7.9 64.1 100 1
33.6 13.1 5.6 40.4 1.1 0.5 4.9 0.8 22.1 9.5 68.4 200 35.3 10.5 6.8 42.6 0.0 0.3 3.8 0.6 17.5 11.4 71.1 400 18.7 14.7 6.4 49.8 0.4 0.4 8.4 1.2 20.7 9.1 70.2 1000 16.5 17.2 7.1 52.1 1.0 0.3 4.9 0.8 22.5 9.3 68.1 2000 12.6 17.1 7.6 56.9 0.0 0.0 4.8 1.1 20.9 9.3 69.8 4000 Table 3-11 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION C (CRACK TIP)
ON FRACTURE FACE IN TUBE R30C64 SHOWN IN FIGURE 3-12 (B)
Atomic Percents Normalized Atomic Percents O
Cr Fe Ni S
C Ti Cr Fe Ni A
a 38.4 9.0 5.4 33.4 1.5 11.6 0.6 18.8 11.3 69.9 0
24.2 8.7 8.0 53.0 0.2 5.9 0.0 12.5 11.4 76.0 100 32.2 9.2 6.7 46.9 0.4 4.2 0.5 1'4.610.774.7 200 32.6 9.2 7.5 46.4 0.2 3.7 0.4 14.5 11.9 73.6 400 20.3 11.3 6.9 51.8 0.2 8.6 0.9 16.2 9.8 74.0 1000 11.9 16.0 8.3 60.8 0.0 2.5 0.5 18.8 9.8 71.5 2000 14.4 15.1 7.4 58.9 0.0 3.7 0.5 18.5 9.1 72.4 4000
~
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y-1i lt Figure 3-13.
Optical Photomicrograph of the Section of i
Tubing Form Tube R16C74 Received Showing the Crack l'
Examined and the Initial Cut Made to Extract a Section for l!
Analysis 1
l The fracture face of a 400 pm deep crack in R16C74 analyzed is f'
shown in Figure 3-15.
Three grains as shown were analyzed:
A, lt near the crack mouth, B,
near the center, and C near the crack 1
tip.
A higher magnification SEM photomicrograph is shown in i
1 Figure 3-16.
The ruptured oxide film on the OD in the vicinity j
of the crack mouth is clearly visible in the photomicrograph.
All j
of the grains including those near the crack mouth are featureless and contain no deposits.
Tables 3-13 through 3-15 are the in-l depth composition profiles of regions A, B,
and C respectively, j j Major contaminants in the oxide are S, Cl, C,
Cu, 2n, and F.
These changed very little in concentration with depth for f*
approximately 1.3 pm, 3000A, and 1000A of sputtering in regions A, j.
B, and C respectively.
These species then decreased slowly with ll 1
1 3-17 1
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Figure 3-14.
SEM Photomicrograph of the Section of Tube l
R16C74 Received Showing the Extensive IGA Table 3-12
]
AUGER IN-DEPTH PROFILE OF A REGION ON THE
.R I
OD OF TUBE R16C74 i
Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Si S
C Ch Cu Cr Fe Ni A
16.6 0.0 7.6 6.9 2.8 3.9 60.6 0.0 1.7 0.0 52.4 47.6 0
1 i
20.8 0.0 7.7 12.3 5.3 4.7 47.1 0.6 1.5 0.0 38.5 61.5 200 35.8 0.0 5.3 13.3 8.4 3.9 31.3 0.7 1.3 0.0 28.7 71.3 400 l
l 29.9 0.0 8.9 18.3 13.5 3.4 23.5 1.2 1.4 0.0 32.8 67.2 800 j
32.2 0.0 9.5 20.5 15.1 2.7 17.3 1.4 1.3 0.0 31.8 68.2 1500 32.6 0.0 12.1 16.0 12.5 2.4 22.5 0.8 1.2 0.0 43.0 57.0 1600 3
r 32.3 3.6 13.1 16.3 11.9 2.3 18.5 0.8 1.2 11.0 39.6 49.4 3200 33.3 3.7 15.9 16.8 12.7 1.6 13.9 1.0 1.2 10.2 43.7 46.2 6400 33.8 7.1 17.1 17.4 10.8 1.1 10.3 1.0 1.3 17.0 41.2 41.8 12800 36.3 9.5 17.1 19.6 9.7 0.4 6.5 1.0 0.0 20.6 37.0 42.3 25600 m
I 41.2 11.6 17.8 23.2 0.0 0.0 5.2 1.0 0.0 22.0 33.8 44.1 36455 i
41.7 13.0 17.0 25.0 0.0 0.0 2.5 0.8 0.0 23.6 30.9 45.5 51200 40.6 12.7 17.2 25.1 0.0 0.0 3.5 0.9 0.0 23.2 31.2 45.6 70630 i
j 40.7 13.2 16.3 27.0 0.0 0.0 2.0 0.8 0.0 23.3 28.8 47.8 80000 un 38.0 14.2 13.5 28.8 3.1 0.0 1.8 0.5 0.0 25.1 23.8 51.0 100000 38.5 15.1 14.5 29.0 1.6 0.0 0.8 0.6 0.0 25.8 24.7 49.5 240000 m
i i
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Figure 3-15.
SEM Photomicrograph of the Fracture Face of R16C74 Examined Showing the Grains Analyzed l-
!!)
i i
i i
t l [
lt E':
w I
~
t I{t l
J.
?@rp' f ,
Figure 3-16.
SEM Photomicrograph of a Typical Region Near the OD on the Fracture Face of R16C14 Examined i
+
3-19 i
l l<
l I
EPRI Lic:nsed M:t: rial Table 3-13 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION A (CRACK MOUTH)
ON FRACTURE FACE OF TUBE R16C74 SHOWN IN FIGURE 3-15 Atomic Percents Normaltred Atomic Percents O
Cr Fe Ni St S
Cl C
Q; Zn F
Cr Fe Ni A
24.5 0.0 0.0 20.2 0.6 5.3 0.7 34.4 7.6 3.4 3.4 0.0 0.0 100.0 0
31.2 0.0 0.0 23.1 0.7 3.2 1.5 27.2 6.4 2.2 4.5 0.0 0.0 100.0 50 33.0 0.0 0.0 25.9 1.1 3.1 1.0 26.0 5.8 0.0 4.1 0.0 0.0 100.0 100 24.3 0.0 0.0 27.6 0.9 4.2 1.0 27.8 8.3 2.1 3.8 0.0 0.0 100.0 200 32.9 0.0 0.0 27.2 1.4 3.4 0.5 23.4 6.2 1.7 3.3 0.0 0.0 100.0 400 32.2 0.0 0.0 26.2 1.1 3.6 0.7 25.4 5.9 1.4 3.5 0.0 0.0 100.0 800 33.0 0.0 0.0 26.4 0.8 3.9 0.6 25.7 5.0 1.4 3.2 0.0 0.0 100.0 1600 32.3 0.0 0.0 27.8 1.6 4.4 0.6 22.6 6.1 1.6 3.0 0.0 0.0 100.0 3200 1
32.3 0.0 0.0 28.6 1.3 4.6 0.7 21.9 6.2 1.5 2.8 0.0 0.0 100.0 6400 22.6 8.5 7.3 31.5 0.3 2.5 0.5 19.0 4.1 1.2 2.5 17.9 15.5 66.6 12800 19.1 11.1 11.9 38.2 0.4 2.1 0.2 12.3 3.3 0.0 1.4 16.2 19.4 62.4 20000 Table 3-14 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION B (MID-POINT ON FRACTURE FACE OF TUBE R16C74)
IN FIGURE 3-15 Atomic Percents No'malized Atomic Percents O
Cr Fe NI Si S
Cl C
Cu Zn F
Cr Fe Ni A
30.0 2.5 0.0 26.8 1.4 4.6 0.7 28.4 0.8 1.8 2.9 8.4 0.0 91.6 0
29.5 3.5 0.0 26.9 1.3 5.0 0.7 28.3 0.9 1.7 2.3 11.5 0.0 88.5 50 30.1 3.2 0.0 26.9 1.8 5.2 0.6 27.0 0.8 2.4 2,1 10.5 0.0 89.5 100 30.7 3.2 0.0 27.4 1.8 5.3 0.7 25.8 0.7 2.1 2.3 10.5 0.0 89.5 200 32.8 2.8 0.0 27.5 2.2 5.8 0.9 22.0 0.6 2.5 2.9 9.3 0.0 90.7 400 32.2 3.8 0.0 31.0 2.4 6.6 0.4 17.4 1.6. 2.5 2.1 10.9 0.0 8).1 800 30.2 4.3 0.0 32.2 1.7 5.9 0.6 21.8 0.0 1.8 1.4 11.8 0.0 88.2 1600 26.2 6.1 0.0 38.4 1.3 5.5 0.4 17.4 1.0 2.4 1.3 13.8 0.0 86.2 3200 22.5 9.6 0.0 42.6 0.9 4.9 0.4 16.2 0.9 1.2 0.8 18.4 0.0 81.6 6400 16.0 14.2 4.4 51.2 0.0 2.1 0.2 9.8 0.8 0.0 1.3 20.3 6.3 73.4 12800 10.7 16.0 6.4 59.4 0.0 0.7 0.2 6.2 0.0 0.0 0.4 19.6 7.8 72.6 20000
.i additional sputtering.
All three regions analyzed were completely depleted of Fe near the surface ranging from approximately 100A in depth in region C to 7000A in regions B and A.
All three regions were depleted of Cr.
No Cr was detected in the first 6400A of the film sputtered in region A (The film was nickel oxf e with the contaminants listed above).
In region B, the Cr, relative to the 3-20
3 j.
EPRILic:nsed Act:rti I
other alloy metals, was approximately a factor of two below that
- t in the alloy in the outer 800A of the film.
The Cr in region C,
- t relative to the other alley metals, was approximately 80s that in the alloy.
Table 3-15 AUGER IN-DEPTH COMPOSITION PROFILE OF REGION C (CRACK TIP) ON FRACTURE FACE OF TUBE R16C74 SHOWN IN FIGURE 3-15
'4 Atomic Percents Normalized Atomic Percent O
Cr R
Ni Si S
Cl C
F Cr N
Ni A
36.9 4.6 0.0 31.4 1.5 2.7 0.5 19.5 2.8 12.9 0.0 87.1 0
38.4 4.6 0.0 36.2 1.6 3.8 0.4 13.0 2.1 11.2 0.0 88.8 25 38.1 6.0 0.0 37.0 1.6 3.7 0.3 11.6 1.7 13.9 0.0 86.1 50 36.0 6.5 1.4 38.6 1.4 3.5 0.2 11.1 1.4 13.9
.3.0 83.1 100 33.1 7.6 2.6 43.5 1.0 2.6 0.3 8.5 0.8 14.1 4.9 81.1 200 24.6 8.6 4.7 51.6 0.0 2.0 0.2 7.6 0.6 13.3 7.3 79.5 400 i
20.9 10.4 5.3 55.9 0.2 1.5 0.2 5.1 0.3 14.5 7.4 78.0 800 16.5 12.3 6.2 58.2 0.3 1.6 0.0 4.9 0.0 16.0 8.0 75.9 1600
{
10.4 14.6 7.1 63.9 0.0 0.4 0.0 3.6 0.0 17.0 8.3 74.7 3200 10.3 15.7 7.5 61.6 0.0 0.5 0.0 4.5 0.0 18.5 8.8 72.7 6400 5.2 16.6 8.6 69.7 0.0 0.0 0.0 0.0 0.0 17.5 9.0 73.5 12800 Figure 3-17 is an SEM photomicrograph of tube R12C70 in the a
as-received condition showing the OD.
Table 3-16 is the Auger in-depth profile of the OD.
The surface oxide was iron oxide, magnetite, with 7 to 16 a/o Ni, Environmental contaminants were Si, S,
C, Ca, and Zn.
Cr was present after 8 m was sputtered at a relative concentration somewhat less than that in the alloy.
In the as-received condition an IGSCC crack in R12C70 had already been pulled to failure in the laboratory (Figure 3-18).
The tube had not been burst tested.
No other cracks could be found in the l
tubing received.
Figure 3-19 shows the exposed IGSCC crack, which is approximately 250 pm in depth.
Two grains were profiled, one near the crack opening, A,
and one near the crack tip, B.
3-21
EPRILicensed Material
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Figure 3-17.
SEM Photomicrograph of As-received Section i
l of Tubing R12C70 showing the OD Surface l
l' Table 3-16 i
AUGER IN-DEPTH COMPOSITION PROFILE OF A REGION ON THE OD OF TUBE R12C70 l
Atomic Percents Normalized Atomic I
Percents O
Cr Fe Ni Si S
C La Zn Cr Fe Ni A
l 1
15.3 0.0 7.3 1.5 2.7 1.4 70.6 0.2 1.1 0.0 83.4 16.6 0
17.9 0.0 13.7 2.4 4.6 1.7 57.6 0.2 ' 1.8 0.0 65.2 14.8 100 17.2 0.0 17.3 3.0 4.5 1.4 54.7 0.6 1.2 0.0 65.1 14.9 200 17.1 0.0 20.3 2.8 4.1 1.3 51.7 0.7 1.9 0.0 87.8 12.2 400 16.5 0.0 25.6 2.5 4.1 1.1 48.1 1.0 1.0 0.0 91.0 9.0 800 18.0 0.0 29.0 2.8 4.1 1.1 42.9 0.9 1.3 0.0 91.3 8.7 1600 19.4 0.0 30.4 2.7 5.2 1.1 39.1 0.9 1.3 0.0 91.9 8.1 3200 22.2 0.0 31.9 2.6 6.1 1.0 33.4 1.6 1.4 0.0 92.5 7.5 6400 26.2 0.0 32.5 3.3 8.4 0.8 25.7 1.9 1.1 0.0 90.8 9.2 12800 32.4 0.0 33.8 4.0 11.2 0.6 14.9 2.3 1.0 0.0 89.5 10.5 27500 35.1 0.0 36.9 4.6 9.9 0.1 9.8 2.5 1.1 0.0 88.9 11.1 40000 36.9 0.0 40.8 6.7 8.6 0.0 5.0 2.1 0.0 0.0,,85.9 14.1 60000 35.1 6.8 31.8 13.8 7.7 0.0 3.0 1.8 0.0 12.9 60.7 26.4 80000 33.1 8.029.321.5 5.3 0.0 1.5 1.3 0.0 13.6 49.8 36.6 97000 3-22 i
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Figure 3-18.
SEM Photomicrograph of As-received Section l.
of Tubing R12C70 Showing the Fracture Face Analyzed i
i
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h i y N..,
h~
l
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h, t%[,/..
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Figure 3-19.
SEM Photomicrograph Showing the Grains i,
Profiled on the Fracture Face of R12C70
.I 3-23 ja i,
i
l.
EPRILicensed Material I =
Figure 3-20 is a higher magnification SEM photomicrograph of the fracture face showing that bands of deposits were present.
Tables 3-17 and 3-18 are the in-depth profiles of regions A and E respectively.
No Cr was observed in the first 100A of the film.
l After Cr was detected in the film, it had a relative concentration less than that in the alloy until approximately half the thickness of the film had been sputtered, 6400A on grain A and 1600A on
~
grain B.
The relative amounts of Fe were approximately 50i more than that in the alloy.
The contaminating species were Si, S,
and j
C.
l l
f i
Gl-i s$'
i j'
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Figure 3-20.
SEM Photomicrograph Showing Typical Region l
of the Fracture Face of R12C70 in High Magnification s
j 3-24 l
1
EPRILic:nsed M:t:ri:I i
l Table. 3-17
.s.
I:
AUGER IN-DEPTH COMPOSITION PROFILE OF - REGION A 3
(CRACK HOUTH)
ON FRACTURE FACE OF TUBE R12C70 SHOWN IN FIGURE 3-19.
?
~I Atomic Percents Normalized Atomic Percents O
Cr
- Fe Ni Si S
C Cr Fe Ni A-
.l 45.3 0.0 4.2 40.5 2.2 0.3 0.0 0.0 9.3 90.7 0-t 46.2 0.0 4.9 41.7 3.3 0.6 0.0 0.0 10.5 89.5 100 0
43.7 4.9 4.8 39.6 4.0 0.6 0.0 9.9 9.7 80.3 200
.f.
42.5 5.'6 6.2 39.2' 4.3 0.4 0.0 11.1.
12.1
-7A R ana
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39.1 - 7.5 6.4 40.0 5.8 0.3 0.0 13.8 11.9 74.3 800 34.1 9.4 6.0 43.0 5.9 0.3 0.0 16.1 10.3 73.6
'1600-I.
25.2 12.3 7.4 47.6 5.7 0.0 0.0 18.3 11.0 70.6 3200 22.1 13.1 7.4 51.4 4.4 0.0
-0.0 18.3
'10.3 71.4 6400 16.0-16.1 9.0 55.8 2.3 0.0 0.0 19.9 11.2 68.9 12800 13.1 16.8 9.5 58.8 0.4 0.0 0.0 19.8 11.1 69.1 25000 1
14.0 15.8 9.4 55.4 0.2 0.0-0.0 19.7 11.6 68.7 50000 Table 3-18 AUGER IN-DEPTH COMPOSITION PROFILE. 0F REGION B'
l I
l (CRACK TIP) ON FRACTURE FACE OF TUBE R12C70 SHOWN IN FIGURE 3-19 Atomic P,ercents Normalized ' Atomic i
Percents O
Cr Fe Ni Si S
C Cr Fe Ni A
45.5 0.0 4.3 37.1 0.2 0.6 12.2 0.0 10.5 89.5 0
44.9 0.0 4.4 43.6 0.4 1.1 5.7 0.0 9.3 90.7 100 J
42.5 5.0 5.2 43.0 0.2 0.6 3.5 9.4 - 9.8 80.8 200 i
39.3 8.8 4.9 44.3 0.0 0.5 2.1 15.2 8.4 76.3 400 32.3 10.3 5.4-48.5 0.2 0.3 3.0 16.1 8.5 75.5 800 21.7 12.6 7.7 55.5 0.0 0.2 2.4 16.7 10.1 73.2 1600
]
17.3 14.1 8.8 57.6 0.0 0.0 2.2 17.5 11.0.71.5-3200 ~
11.1 16.0 8.4 62.2 0.0 0.0 2.2 18.5 9.7 71.8 6400 l
12.7 16.1 9.1 60.1 0.0 0.0 2.0 18.9 10.7 70.5 12800 s
7.0 16.9 10.2 64.5 0.0 0.0 1.3 18.5 11.2 70.4 25000.
l I
13.8 15.2 9.0.57.0 0.0 0.0 5.0 18.7 11.1 70.1 50000~
i-
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4>,.-,-
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EPRILicensed Material Figure 3-21 shows an optical photomicrograph of the OD section of tubing from Tube R20C66.
The cut across the axis of the tube extending across most of the section of tubing was made after this 3
section was received from ABB/ Combustion Engineering.
The crack j
which was analyzed is indicated by an arrow.
Shallow IGA was l
observed over most of the OD below the laboratory cut
)
(Figure 3-22).
These cracks were apparently opened when the tube was burst testud.
Table 3-19 is the in-depth composition profile of the OD obtained from an area between IGA penetrations.
The impurities in the oxide were Si, S,
C and Ca.
The major impurity is Si, which was in concentrations exceeding 15 a/o.
After sputtering 6 pm of oxide, the only impurity remaining was Ca, i
which was less than 0.3 a/o.
Considering the alloy metals, Cr had 3-
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Optical Photomicrograph (6X) of As-received Section of Tubing from R20066.
The Cut in the Section was Made After Receipt from ABB/ Combustion Engineering i
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3-26 i
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t 20prp i men Figure 3-22.
SEM Photomicrograph of the OD of Tube R20C66 i
Showing the Shallow IGA 1
Table 3-19 AUGER IN-DEPTH COMPOSITION PROFILE OF THE FILM ON THE OD OF TUBE R20C66 BETWEEN IGA i
Atomic Percents Normalized Atomic I
Percents O
Cr Fe Ni Si S
C Ca Cr Fe Ni A
i 37.7 8.5 11.4 10.2 4.9 0.1 27.1 0.0 28.3 37.8 33.9 0
,j 41.1 7.8 12.3 15.7 16.8 0.1 5.4 0.4 21.7 34.5 43.8 650
' 4. 6 1600 48.9 4.4 14.2 9.9 16.9 0.2 4.0 0.3 15.5 50.0 J
l !
48.5 2.9 13.8 13.2 16.4 0.2 3.5 0.3 9.8 46.1 44.1 6400 46.4 5.0 14.4 14.0 15.5 0.2 2.9 0.3 14.8 43.2 41.9 12800 38.4 10.4 13.5 23.4 10.5 0.1 2.4 0.4 22.0 28.6 49.4 25600 I
(
ll 21.2 14.9 11.8 47.3 2.8 0.0 1.3 0.3 20.1 16.0 63.9 50000 17.7 15.8 10.2 53.2 1.1 0.0 1.4 0.3 19.9 12.9 67.2 60000 17.5 15.3 7.0 59.4 0.0 0.0 0.5 0.2 18.8 8.5 72.7 80000 16.6 15.3 7.2 60.6 0.0 0.0 0.0 0.4 18.4 8.6 73.0 90000 4
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a minimum in concentration after 6400A of film was removed with a corresponding increase in Fe.
These extremia apparently occur at i
the scale / alloy interface since Ni, Cr, and Fe began their approach to the concentration in the alloy as additional sputtering was performed.
Table 3-20 is the in-depth profile from the wall of an IGA penetration.
The environmental species incorporated in the oxide were Si, S and C.
Aside from C, these species changed little in concentration to a depth of approximately 400A.
These environmental species slowly decreased in concentration during additional sputtering.
Carbon behaved differently.
It had a very high contamination level on the surface and remained in high concentrations throughout the film.
Relative to the alloy, Fe was enriched by approximately a factor of two and Cr was slightly depleted.
Table 3-20 AUGER IN DEPTH COMPOSITION PROFILE OF A REGION INSIDE A CRACK CAUSED BY IGA (TUBE R20C66)
Atomic Percents Normalized Atomic Percents O
Cr Fe NI Si S
C Cr Fe Ni A
27.8 11.0 7.7 32.5 1.5 1.6 18.0 21.4 15.0 63.5 0
20.9 10.3 9.5 46.9 2.5 1.1 8.8 15.5 14.2 70.3 100 2 0, ^ 10.7 10.1 48.7 2.3 1.1 7.1 15.4 14.6 70.0 150 19.6 10.3 10.7 49.0 2.5 1.0 6.9 14.7 15.3 70.0 200 19.0 11.1 11.0 50.3 1.8 1.0 5.8 15.3 15.2 69.5 250 18.8 10.3 11.3 49.8 2.3 1.0 6.5 14.5 15.8 69.7 300 18.2 10.9 11.7 51.3 1.5 0.8 5.4 14.8 15.9 69.4 350 18.4 10.2 11.8 50.4 1.9 0.9 6.4 14.1 16.3 69.5 400 18.8 10.5 12.6 51.5 0.0 0.5 6.2 14.0 16.8 69.1 450 15.9 11.5 11.9 53.3 0.0 0.7 6.6 15.0 15.5 69.5 550 15.5 11.9 12.9 54.0 0.0 0.6 5.1 15.2 16.4 68.5 650 12.9 15.3 6.4 53.8 0.9 0.1 10.5 20.2 8.5 71.3 1600 Figure 3-23 shows the grair.s profiled on the fracture f ace from i'
tube R20C66, where grain A is near the crack mouth, grain D is near the crack tip and Graina B and C are between A and D.
The j
in-depth composition profileu from the R20C66 fracture face are 3-28
EPRILic:nsed M't:rl:I l'
I i
given in Tables 3-21 (Grain A), 3-22 (Grain B), 3-23 (Grain C),
I and 3-24 (Grain D).
Considering the relative concentrations of the alloy metals, the average concentration of Cr in the films on
-l the two grains analyzed nearest the crack mouth, Grains A and B, was approximately the same as that in the alloy; whereas, the
_l average concentration of the Cr in the films on the two grains analyzed nearest the crack tip, Grains C and D, was slightly less than that in the alloy.
Fe was enriched in the film nearest the I
crack mouth, A,
and approximately the same as that in the alloy in
(
the films on the other grains.
Environmental species found in the film were Si, S,
Ti, and C.
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SEM Photomicrograph Showing the Grains l
Analyzed on the Fracture Face of a Crack in Tube R20C66, l
Where A is Near the Crack Houth, D is Near the Crack Tip and Grains B and C are Between the Crack Houth and Crack l
Tip
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EPRILicensed Material Table 3-21 AUGER IN-DEPTH COMPOSITION PROFILE OF GRAIN A t-(CRACK HOUTH)- ON FRACTURE' FACE OF TUBE R20C66 IN FIGURE 3-23 t'l Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Si S
Ti C-Cr-Fe Ni A
I 24.4 3.2 2.2 16.4 4.1 1.3 0.6 47.7 14.7 10.2 75.1 0
29.4 3.0 2.5 12.8
- 4. 0 2.2 0.5 45.5 16.5 13.8 69.6 200 29.2 4.4 4.2 15.9 5.3 2.0 0.7 38.3 17.9.17.2-64.9 400.
.g 29.4 4.9 5.3 16.9 6.9 1.7 0.5 34.5 18.0.19.5 62.5' 800 28.3 7.2 5.1 22.7 4.4 1.5 0.5 30.4 20.6 14.5 64.9 1600 27.7 8.4 6.7 22.9 6.6 1.0 0.4 26.4 22.1 17.7 60.2 3200 29.1 9.3 9.6 31.2 5.5 0.7 0.4 14.3 18.6 19.1 62.3 6400
.l 31.0 8.5 6.9 37.3 5.1 0.4 0.5 10.3 16.1 13.1 70.8 12800 30.2 9.5 7.0 39.5 5.0 0.5 0.6 7.7 17.0 12.4 70.5 12900
.j 30.0 9.8 7.7 35,9 5.2 0.6 0.5 10.3 18.3 14.3 67.3 25600 27.3 12.1 5.7 42.4 2.4 0.7 0.5 8.9 20.1
'9.6 70.4 32000 26.0 12.1 6.8 47.5 2.6 0.2 0.0 5.0 18.2 10.2 71.6 56600
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21.2 13.1 6.8 50.8 1.4 0.2 0.5 6.0 18.5 9.7 71.8 66600 27.4 11.7 5.4 44.0
- 6 0.6 0.7 8.6 19.2 8.8 72.0 75600 i.
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3-31 1
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EPRI Licen:ed Mat:ri:I Table 3-22 AUGER IN-DEPTH COMPOSITION PROFILE OF GRAIN B r
IN FIGURE 3-23 (TUBE R20C66)
Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Si S
Tl C
Cr Fe Ni A
5 27.0 5.3 0.0 15.9 1.7 1.5 0.6 48.0 25.1 0.0 74.9 0
1 30.1 7.5 2.0 24.5 2.4 1.6 0.0 31.9 22.1 5.8 72.1 200 31.9 8.3 2.6 28.6 2.7 1.3 0.0 24.5 21.0 6.6 72.4 400 7
I 29.4 8.8 2.8 29.1 1.4 1.3 0.7 26.4 21.6 6.8 71.5 800 34.5 8.1 3.8 35.2 4.9 1.2 0.4 12.0 17.2 8.2 74.7 1600 27.6 10.1 4.1 39.9 1.2 0.8 0.0 16.3 18.7 7.5 73.7-3200 33.8 13.2 4.9 39.9 1.2 0.3 0.3 6.5 22.7 8.4 68.9 6400 35.5 12.8 4.4 40.4 1.1 0.3 0.0 5.4 22.2 7.7 70.1 12800 35.8 12.1 4.6 41.2 0.9 0.5 0.0 4.8 20.8 8.0 71.2 19200 34.1 12.8 4.8 41.0 i.4 0.1 0.5 5.3 21.9 8.2 69.9 25600 36.0 11.9 3.8 42.5 1.0 0.3 0.5 4.1 20.4 5.5 73.1 32000 23.5 16.9 5.2 47.0 0.6 0.2 0.5 6.1 24.5 7.6 68.0 56600 25.5 13.0 5.9 44.1 0.7 0.3 0.4 9.9 20.6 9.4 70.0 66600 27.4 11.6 5.1 43.7 2.1 0.2 0.6 9.1 19.1 8.5 72.4 75600 i
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3-32
EPRILic:nsed M:t;rtl o
1 Table 3-23 l
AUGER IN-DEPTH COMPOSITION PROFILE OF GRAIN C IN FIGURE 3-23 (TUBE R20C66) l Atomic Percents Normalized Atomic Percents O
Cr Fe Ni Si S
Ti C
Cr Fe Ni A
I 39.7 6.1 3.6 31.9 5.4 1.5 0.3 11.4 14.8 8.7 76.5 0
41.1 6.7 3.6 29.6 4.3 1.0 0.0 13.8 16.7 9.0 74.3 200 I
39.9 7.1 3.4 31.4 4.6 1.0 0.3 12.1 17.0 8.2 74.9 400 4
42.0 10.9 4.3 37.0 0.4 1.7 0.5 3.2 21.0 8.3 70.8 800 32.2 10.3 3.0 44.3 0.0 1.4 0.4 8.4 17.9--S ? 7 c o 39.8 4.3 4.4 37.6 4.3 0.9 0.6 8.1 9.3 9.6 81.1 3200 37.1 8.7 5.3 37.4 4.5 1.0 0.3 5.6 16.9 10.4 72.7 6400 31.9 9.7 5.0 41.7 2.8 0.7 0.5 7.7 17.2 8.9 73.9 12800 33.4 8.6 3.8 40.7 4.8 0.7 0.3 7.7 16.1 7.2 76.7 19200 33.2 8.4 4.6 40.6 3.9 0.6 0.5 8.1 15.7 8.6 75.7 25600 31.5 10.7 5.4 41.6 2.8 0.5 0.4 7.1 18.5 9.3 72.2 32000 28.6 11.8 4.9 43.6 1.9 0.5 0.5 8.2 19.6 8.1 72.3 56600 27.7 11.2 4.7 43.4 1.8 0.6 0.5 10.1 18.9 7.9 73.2 66600 25.7 13.5 5.0 46.3 1.6 0.3 0.5 7.2 20.8 7.7 71.5 75600
(
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f 3-33
EPRILic:nsed M:t:ri:I Table 3-24 AUGER IN-DEPTH COMPOSITION PROFILE OF GRAIN D (CRACK TIP)
ON FRACTURE FACE OF TUBE R20C66 IN FIGURE 3-23 1
Atomic Percents Normalized Atomic i
Percents O
Cr Fe Ni Si S
Ti C
Cr Fe Ni A
42.4 4.7 0.0 36.0 0.3 1.6 0.0 15.0 11.5 0.0 88.5 0
I 41.7 6.7 4.1 39.2 0.0 1.6 0.0 6.7 13.4 8.2 78.4 200 41.1 8.3 3.5 40.8 0.0 1.4 0.4 4.5 15.8 6.7 77.4 400 i
36.8 11.2 3.4 44.0 0.0 1.2 0.4 2.9 19.1 5.8 75.1 800 28.1 8.6 3.2 32.8 1.9 1.2
-0.5 23.8 1 9. 3 -7, '
"r 18.5 11.9 4.9 60.3 0.0 0.3 0.3 3.8 15.4 6.4 78.3 3200 l
14.1 14.1 5.8 60.9 0.0 0.0 0.6 4.5 17.4 7.2 75.3 6400 12.0 15.8 6.6 62.4 0.0 0.0 0.3 3.0 18.6 7.7 73.6 12800 12.1 15.0 7.0 60.5 0.0 0.3 0.6 4.5 18.2 8.5 73.3 191200 13.8 14.2 6.0 59.0 0.0 0.0 0.5 6.4 17.9 7.6 74.5 25600
{
13.9 14.7 6.8 58.7 0.0 0.0 0.4 5.5 18.3 8.4 73.2 32000 8.2 16.7 5.5 65.0 0.0 0.0 0.6 4.0 19.2 6.4 74.5 56600 Figure 3-24 shows the section of Tube R12C8 as-received.
This section had only shallow intergranular cracks a few grains deep.
Table 3-25 is the in-depth composition profile from the OD surface.
The surface film is a Fe-Ni oxide with a few percent of incorporated Cr. Silicon is the major environmental impurity.
Ca is present in concentrations up to 3.5 a/o. zn and Cu are present in the 1-3 a/o range, and C1 is always less than 1 a/o.
The IGSCC fracture face analyzed was that exposed in the as-received condition shown in Figure 3-24 and in higher magnification in Figure 3-25.
Table 3-26 is the Auger in-depth composition profile of an area on the fracture surface.
Considering only the alloy metals, both Cr and Fe were present in relative quantities slightly less than those in the alloy.
The major impurity from the environment was Pb, which had a surface concentration of 14.5 a/o and decreased in concentration with depth.
Copper was present in the 5-6 a/o level in the outer 1600A and decreased with additional sputtering.
l 1
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5,00 m' Figure 3-24.
SEM Photomicrograph of As-received Section from Tube R12CB i
Table 3-25 l'
AUGER IN-DEPTH COMPOSITION PROFILE OF THE OD OF TUBE R12C8 i
Atomic Percents Normalized Atorruc Percents i
l O
Cr Fe Ni Si S
Cl C
Q Cb Zn Cr Fe Ni A
a 42.1 0.0 9.2 5.0 7.2 1.4 0.5 29.5 1.2 1.3 2.6 0.064.635.4 0
s 47.2 0.0 10.4 4.6 13.4 1.3 0.4 17.1 1.7 1.1 2.8 0.069.530.5 50 l
47.6 1.4 10.0 4.5 16.3 1.3 0.3 12.7 1.7 1.1 3.0 8.662.928.5 100 i
48.4 0.0 10.2 4.4 19.0 1.2 0.3 10.7 1.9 1.2 2.6 0.069.730.3 200 l
47.2 2.3 94 4.0 20.4 1.1 0.3 9.3 2.1 1.3 2.6 14.5 60.0 25.6 400 45.9 1.8 11.2 4.5 20.6 1.1 0.2 8.8 2.3 1.1 2.6 10.5 64.0 25.5 800 l
] ;
44.4 1.6 13.4 5.4 19.8 1.1 0.1 8.9 2.6 1.1 1.5 7.665.726.7 1600 3
l 44.5 1.5 16.0 5.0 18.5 0.8 0.1 7.5 2.8 1.3 2.1 6.9 71.0 22.1 3200 42.0 1.8 19.6 6.2 17.5 0.8 0.2 5.5 3.2 1.0 2.1 6.671.022.4 6400 i'
38.9 2.0 25.0 6.6 15.5 0.5 0.2 4.5 3.5 0.9 2.3 5.9 74.3 19.8 12800 j
35.8 2.7 30.9 8.5 12.9 0.2 0.1 3.0 3.3 0.7 1.8 6.4 73.4 20.2 20000 l
33.4 3.5 33.7 14.2 8.9 0.2 0.0 2.0 2.8 0.0 1.3 6.865.627.6 12800 j;
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F Figure 3-25.
SEM Photomicrograph of the Fracture Face Analyzed From Tube R12C8 l
l, Table 3-26 AUGER IN-DEPTH COMPOSITION PROFILE i
OF THE FRACTURE FACE OF TUBE R12C8
.i,
l Atomic Percents Normalized Atomic i
Percents O
Cr Fe Ni C
Cu Pb Cr Fe Ni A
l 39.7 7.3 0.0 26.2 6.6 5.9 14.3 21.9 0.0 78.1 0
41.8 6.8 1.7 29.2 2.7 6.0 11.8 17.9 4.6 77.4 50 41.1 6.7 1.8 31.5 2.0 5.9 11.1 16.8 4.4 78.7 100 l
l 38.7 6.7 1.9 33.8 2.5 5.9 10.4 15.9 4.5 79.6 200 37.0 6.7 2.4 36.9 1.6 5.6 9.8 14.5 5.3 80.2 400 36.6 7.1 3.6 37.3 1.7 5.1 8.7 14.7 7.4 77.8 800 l
l 35.0 10.3 5.0 35.4 2.1 5.3 6.9 20.4 9.8 69.8 1600 27.1 11.1 5.5 46.5 2.1 3.5 4.2 17.6 8.8 73.6 3200 i
22.3 13.7 6.4 49.8 3.7 2.4 1.7 19.5 9.2 71.3 6400 1
19.0 16.1 7.2 51.5 4.4 1.3 0.6 21.5 9.6 68.9 12800 l
17.6 17.2 7.6 52.0 4.9 0.8 0.0 22.4 9.9 67.7 25000 j
20.4 16.3 6.1 48.5 7.9 0.7 0.0 23.0 8.6 68.3 50000 i'
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XPS Results i*
The XPS analysis'of the CD surface and fracture faces identified
'l the corrosion products as mixed oxides and hydroxides.
On all surfaces analyzed, Fe had the distinctive magnetite spectrum.
Chromium and Ni were in mixed oxide / hydroxide states.
An p
approximate deconvolution of the Ni 2P lines gave the following:
tube R25C58 - 40% Ni(OH)2 / 60% NiO, tube R29070 - 35% Ni(OH)2 /
65% NiO on the OD and 40% Ni(OH)2 / 65% NiO on the fracture face, tube R30C64 - 90% Ni(OH)2 / 10% NiO on the OD and 75%
l Ni(OH)2 / 25i NiO on the fracture face, tube R16C74 - 100%
Ni(OH)2 on the OD and 70-% Ni(OH)2 / 30% Nio on the fracture face, tube R12C70 - 40% Ni(OH)2 / 30i NiO on the OD and fracture face, tube R20C66 - 100% Ni(OH)2 on the OD and 75% Ni( H)2 / 25%
O NiO on the fracture face, tube R12C8 - 70% Ni(OH)2 / 25% NiO on the OD (the fracture face was not analyzed because it was too small to give unambiguous results).
The spectra distinctive of 0,
CrOOH and Cr(OH)3 have large linewidths relative to their Cr2 3
energies of separation which makes them difficult to I
deconvolute.
However, Cr203 has a distinctive double peak which was not observed on any of the areas analyzed.
This suggests that Cr was at least 80% to 90% Cr (OH) 3 and/or CrOOH.
When observed, Cu was present either as metallic Cu or Cu20; these XPS lines are not distinguishable.
Silicon was present as 1
silicate.
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DISCUSSION The underlying hypothesis used in relating the Auger and XPS analysis of the corrosion products on the fracture faces and OD tube / tube support plate intersections to crevice' Um.Joutca
.o that the oxides found in these locations are predictable from the Pourbaix diagrams of the pure metals.
These diagrams [5] are shown in Figures 4-1, 4-2 and 4-3 for Fe, Cr, and Ni at 288'C with activities of dissolved species of 10-3 Nickel has the broadest range of stability at high pH's with nickel oxide stable starting at potentials just above the hydrogen line, "a",
(potentials below which water decomposes to form hydrogen gas) to approximately pH 15.
Ni metal is stable at all alkaline and caustic pH's at
{
potentials on the hydrogen line.
Chromium is least stable having I
only soluble species stable at pH's above 9.5 under oxidizing conditions and to potentials several hundred mV below the hydrogen line.
The extent of the stability of iron at high pH's is between that of Cr and N1.
Iron has a band of potentials both above and below the hydrogen line at pH's higher than 11 where soluble HFeO ~
2 is the stable species.
t The surface composition of Alloy 600 laboratory specimens exposed i
to caustic solutions can be predicted oy superimposing the Pourbaix diagrams for pure metals.
Figure 4-4 shows the Auger in-depth composition profile for Alloy 600 exposed to a 315*C, 10%
NaOH solution at open circuit, whian indicates a dealloyed surface enriched in Ni and depleted in Fe and Cr (1) The potential of the system is on the hydrogen line.
A comparison of the Pourbaix diagrams for Fe, Cr, and Ni predict that the most stable insoluble
{
phase is metallic Ni, which is exactly what is observed.
The 200A oxide film, which is on the surface, formed when the autoclave j
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8 12 16 pH Figure 4-1.
Potential-pH Diagram for the Fe-water System (at 2 8 8*C) with Dissolved Species Activities of 10-3 [1]
cooled down.
Figure 4-5 shows the Auger in-depth profile of an Alloy 600 specimen which was exposed to 50% NaOH at 320 C while polarized 150 my above the open circuit potential. (A) The surface t
has a film composed almost entirely of nickel oxide, again in accordance with the Pourbaix diagrams.
Figure 4-6 shows Japanese results (2) for Auger analysis of Alloy 600 which had been exposed to solutions of sulfuric acid, sodium sulfate and sodium hydroxide having varying pH's.
As the pH decreased, Cr enriched uniformly, Ni decreased uniformly and Fe increased slightly.
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Potential-pH Diagram for the Cr-water System i
(at 288*C) with Dissolved Species Activities of 10-3 [1]
Although the practice of relating the Cr concentration in the j
oxide film on alloy 600 to the pH in the tube / tube support plate crevice has a valid thermodynamic and laboratory basis, it has limitations which must be considered.
The major uncertainty arises because the thermodynamic and laboratory data were obtained for environments much simpler than those existing when the films were formed on the steam generator tubes.
T.a crevice environments are concentrated solutions of environmental species i
including high concentrations of iron ions resulting from I
corrosien of the support plate and other steel components exposed
~
to the secondary water.
Laboratory data for the corrosion products formed in these complex environments does not yet exist.
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Potential-pH Diagram for the Ni-water System (at 288 C) with Dissolved Species Activities of 10-3 g)
A common feature in the in-depth composition profiles of the CD of all the tubes plugged and removed in 1991 (Tables 3-3, 3-6, 3-12, and 3-16) is the low chromium concentration near the surface of the film.
The films on the OD of R30C64 (Table 3-8),
R16C74 (Table 3-12), and R12C70 (Table 3-16) were completely depleted of Cr during the first few thousand Angstroms of sputtering.
Although Cr was not completely absent in the in-depth composition profile of the OD of R29C70, several points on the surface were found where there was no Cr, some of which were pure r.ickel exide (Table 3-2).
The film on R29C70 was not homogeneous in the lateral direction because of spalling which occurred during handling or burst testing.
The Fe content in these films, when 4-4 i
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Auger In-Depth Composition Profile for Alloy 600 Which was Exposed to 50% NaOH at 320*C 60 V% -
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Auger In-Depth Compos'ition Profile for Alloy 600 Which Had Been Exposed to 50% NaOH at 320 C While Polarized 150 mV Above the Corrosion Potential, Ecorr 9
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Change in Fe, Cr, and Ni Composition of the H SO /Na2SO /
Surface Film on Alloy 600 After Exposure to 2
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[2.)
compared with Cr and Ni, was several times higher than that in the alloy.
This reflects the high concentration of Fe in the crevice solution as discussed above.
However, the high Ni content in the film on the OD suggests that this film is not simply a deposition product.
Even so, a high pH environment would be unfavorable for deposition of Cr containing oxides.
The profiles of areas on the fracture faces, all of which were intergranular, from cracks in these three tubes also have less Cr, relative to Fe and Ni, than in the alloy.
The high Fe levels found on the OD were not present on the these fracture faces.
The Fe, relative to Cr and Ni, on i
most areas profiled on these fracture faces was less than or approximately the same as that in the alloy.
Tube R12C8 was also removed in 1991.
However, unlike the three tubes discussed above, it was plugged in 1989.
Thus, it was exposed to the secondary water chemistry and not to the l
concentrated crevice solution for two years prior to its removal.
It is not clear what changes, if any, this exposure to the I
a secondary water chemistry would introduce to a film formed in the
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EPRILic:nsed M:t:ri:1 5
crevice solution.
The in-depth composition profile of the film on the CD (Table 3-25) was similar to those plugged and removed in 1991 as discussed previously.
The film profiled on the fracture
~
face (Table 3-26) had Fe and Cr levels slightly less than those in the alloy.
Thus, the chemistries of the films on the OD and fracture faces of tubes R12C8, R12C70, R16C74 and R29C70 suggest the existence of strongly alkaline or caustic tube / tube support plate crevices.
However, the results from R20C66, while not suggesting an acio I
crevice, do suggest a crevice chemistry with a pH lower than those which existed during the formation of the films on R12C8, R12C70, R16C74 and R29C70.
This is based on the profiles of the films on the OD (Table 3-19), in an IGA penetration (Table 3-20) and on a fracture face (Tables 3-21 through 3-24), which are not depleted in Cr, but have a concentration approximately the same as the alloy, relative to Fe and Ni, when averaged over all depths analyzed.
Tube R20C66 was plugged in 1990 and removed in 1991.
Thus, it was not exposed to the concentrated crevice environment l*
for a period prior to its removal.
It is not clear why the Cr levels are different for R20C66.
It could be explained by the presence of deposits.
There is considerably more scatter in the data in the profiles than that from the other tubes.
The film on the OD of R25C58 (Table 3-1), which was removed in 1986, had relative Cr concentrations as high as those observed on R20C66.
The film on R25C58 dif fered f rom those on the OD of the other tubes since the relative Ni concentration was lower, silica was not ancorporated in the film (the CD films on other tubes had maximum concentrations of 10 to 20 a/o silica), and the Zn content was several times higher than that in the films on other sections of tubing.
The presence of Zn suggests that the crevice chemistry was neither highly acidic nor strongly alkaline since zine oxide is soluble in these conditions.
Zine has been observed in' films on alloy 600 which had been exposed to mildly acidic solutions at 288*C i6).
Since the Trojan condenser was replaced with a Ti 4-7
EPRILicen:ed M:teri:1 condenser during the 1987 refueling outage, nothing can be concluded from the absence of Zn in the films on tubes removed after their date.
The presence of silica in the crevice could be
?
an indication of a highly alkaline or caustic condition, under 1
which conditions silica is soluble, or a result of reduced demineralizer efficiency resulting from continuous on-line borate
{
treatment, which was initiated in 1989.
4 P
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EPRILic:nsed M:t:rt:I l.
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'4 Section 5 CONCLUSIONS l
The tubes removed in 1991 had Cr depletion in the films of corrosion products on the OD and fracture face of sections of i
i t
tubing from tube / tube support plate intersections.
The tube removed in 1986 had a Cr enriched corrosion product film on the OD of a tube / tube support plate intersection.
Available laboratory data and thermodynamic considerations suggest that Cr depletion correlates with a caustic to highly alkaline environment and Cr enrichment correlates with a neutral to acidic environment.
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Section-6 REFERENCES
.l:
1.
J.B. Lumsden, " Mechanisms for Formation and. Disruption of Surface Oxides," EPRI Final Recort NP-5369, Electric. Power-Research Institute, Palo Alto, CA; August 1987 l
K. Onimura, K. - Arioka ~,
j i-2.
H. Takamatsu, K. Matsueda,
.T S. Tokunga and K. Katsura, " IGA /IGSCC Propagation Behaviors
]
of Alloy 600," Proceedings of the Fourth International s
Svroosium on ~Environcental Decradation of Mat'erials in Nuclear Power Systems-Water Reactors. 1999,. Jekyll Island, i
4 USA; 1990; pp. 7 7-44.
i I
t s
3.
J.B. Lumsden, " Oxide Film Compositions and Morphology.on Alloy 600 Tubes from Steam Generators-North Anna-Unit,1 and Point Beach, Unit 1,"'EPRI Final' Report NP-5712, Electric Power Research Institute, Palo Alto, CA; April 1988.
/I 4.
J.B. Lumsden, "The Relationship Between Surface Oxide
}
Chemistries and the Chemistries in Steam Generator
-y Crevices," Proceedinos of the Fourth International Svm_osium on Environrental Decradation of Materials in j
c Nuclear Power Systems-Water Reactors.'1999, Jekyll Island, i
USA; 1990; pp. 6 6-51.
i 5
P.L. Daniel and S.L. Harper, "Use of Pourbaix Diagrams'to Infer Local Pitting Conditions," EPRI Topical Recort-NP-4931, Electric Power Research Institute, Palo Alto,'CA';-
October 1986.
6-l'
EPRILicen::d M:t:ri:I a
g e.
6.
A.M. McKay, " Mechanisms of Venting in Nuclear Steam Generator," Corrosion /82, Preprint No. 214, NACE, Houston, TX.
(1982).
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%-Leadership in Science and Technology Bs
- , 1 ABOUT EPRI E
l The mission of the Electric Power Research Institute is to discover, develop, and deliver g
advances in science and technology for the benefit of member utilities, their customers, and society.
i Funded through annual rnembership dues from some 700 member utilities, EPRI's work covers a wide range of technologies related to the generation, delivery, and use of electricity, W
with special attention paid to cost-effectiveness and environmental concerns.
At EPRl's headquarters in Palo Alto, California more than 3SO scientists and engineers i
manage some 1600 ongoing projects throughout the world. Benefits accrue in the form of products, services, and information for direct application by the electric utiltty industry and i
its customers.
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(continued from front cover)
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This license and this agreement are effectrve until terminated You may terminate them at any time by e
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60 -34'l i
SAFETY REVIEW OF TROJAN PLANT RESTART:
Steam Generator Deterioration and Interim Plugging Criteria s-j Stephen H. Han~auer
. i.
January 18,1993 I
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Technical Analysis Corporation
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6723 Whittier Avenue, Suite 202 Mclean, Virginia 22101 9 03) 883-3700 9 rs } IM f)[' )'
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e TABLE OF CONTENTS Section East 2
1 Introduction and Summarv......................................
2 1.1 Scope of t his Review.....................................
1.2 Incomplete Nature of this Reoort.............................
3 3
13 Summarv.............................................
2 Method of An alysis...........................................
4
)
3 Ste am anmtor wbe Bss....................................
6 6
3.1 Initia tin g Event.........................................
3.2 Significant Functions Recuired to Keen the Core from Melting........
7 33 Ev en t Se cu e nces........................................
8 3.4 Release of Radioactive Materials.............................
8 4
M ain St e am Pipe Bre ak........................................
9 9
4.1 Initiating Event.........................................
4.2 Sienificant Functions Recuired to Keep the Core from Melting........
9 43 Event Se cu ences.......................................
10 4.4 Tube Leakage Ratt.....................................
11 4.4.1 Criteria Before 1991
...............................11 4,4.2 Additional Decradation Discovered in 1991................. 12 4.43 Additional insoections in late 1992....................... 14 4.5 Release of Radioactive Materials............................. 15 4.5.1 Core Melt AccMents...............................
15 4.5.2 Events that do not Melt the Core........................ 16 5
Uncertainties and Marrins....................................... 17 i
i 1
SAFETY REVIEW OF TROJAN PIAhT RESTART:
Steam Generator Deterioration and Interim Plugeing Criteria Stephen H. Hanauer Technical Analysis Corporation January 18,1993 1
Introduction and Summarv 1.1 Scone of this Review In the 1991 outage at Trojan, the usual steam generator tube inspections were supplemented with a more sensitive motorized rotating pancake coil eddy current inspection probe.
These The results of an supplementary inspections identified several hundred tubes with cracks.
extensive program of additional inspections were combined with data from other plants and industry programs.
Additional data were obtained from both in-plant and laboratory examinations of tubes removed from Trojan steam generators. Portland General Electric (PGE),
the utility operator of the Trojan plant, proposed, and the NRC accepted in February 1992, a revised interim tube repair criterion that leaves unrepaired several hundred known cracks that may be deeper than the previous limit of 40% of the tube wall thickness.
The plant was restarted but shut down on November 9,1992, to repair a leak caused by inadequate stress relief after a sleeving repair during the 1991 outage. During the outage that began in late 1992, additional steam generator tube inspections were performed, whose results raised additional concerns.
Staning in December 1991, Dr. Joram Hopenfeld and Mr. Joseph Muscara, NRC Staff technologists, questioned the validity of the NRC-approved February 1992 repair criterion and its basis in data and analysis. On January 5 and January 15,1993, the NRC issued documents setting forth a technical resolution of the concerns expressed by these NRC Staff members.
'Ibe Oregon Department of Energy engaged me to advise whether, in my opinion, Trojan is safe to restan after the current outage that began in November 1992. At the time of this engagement, the apparent issue was the questions raised by Dr. Hopenfeld and Mr. Muscara. An additional issue, questions raised by the additioral deterioration detected in the late 1992 inspections at Trojan, is also discussed in this report.
1
1.2 Incomplete Nature of this Reoort The annauncement on January 4,1993, by Portland General Electric that operation of Trojan was being terminated came while the review described in this report was in progress. He technical work by PGE and its contractors to support restart safety review was terminated, so we will not I
obtain, on this Trojan review at least, the technical information that was still to come on January I
- 4. Derefore, our conclusions and recommendations are necessarily based on less information--
in particular, less documented information--than we would need for a true restart safety l
cvaluation. This report is to be viewed as an interim evaluation that will not be completed.
Since PGE has decided that restart will not occur, even if it were decided that Trojan is safe to restart, my conclusion and recommendations have no practical significancefar Trajan ^~M^
nis report provides a record for the Oregon Energy Facility Siting Council and the Oregon Department of Energy.
o 13 Summarv I conclude that Trojan is safe to restart.
More precisely, I conclude that, if the PGE/ Westinghouse analysis underway on January 4,1993, had been completed and documented, and had confirmed what we were told in conference calls with PGE, Westinghouse, the NRC and Oregon representatives, Trojan would have been safe to restart. I also recommend ther the safety of Trojan be augmented, for restart, by decreasing the operating limit on radioactivity allowable in the primary coolant and by developing plans to replenish the supply of borated water in the refueling wr-r storage tank, if needed during an accident.
My evaluation included the effect of Trojan steam generator tube degradation on the probabilities and consequences of possible accidents in which the degradation might play a significant role.
Such accidents could be initiated by rupture of one or more steam generator tubes during operation, or by a break in one of the main steam pipes, in which possible tube leakage would influence the outcome.
For possible accidents whose initiating event is rupture of one or more steam generator tubes, neither the frequencies of such initiating events nor their consequences is significantly influenced by the Trojan tube degradation. Several such events hsve actually occurred, but the causes have been unrelated to tube deterioration such as that at Trojan.
Similarly, the courses and consequences of possible tube rupture accident event sequences would not be significantly affected by the Trojan tube deterioration. The actual events have been controlied without significant releases of radioactivity.
Another class of possible accidents begins with an initiating event of a break in a main steam pipe, or the sticking open of one of the large safety and relief valves connected to these pipes.
Any significant leaks in the degraded Trojan steam generator tubes during such an event could let the radioactive primary cooling water into the secondary system and out the break (or the 2
lVm
l' stuck-open valve) in the steam pipe, which is outside the reactor containment. This would constitute a release of radioactivity, and also a loss of water available to cool the reactor core and g
keep it from melting.
A recent NRC analysis shows that the Trojan tube degradation would likely lead to increased l
tube leakage in such an event, with a leakage rate predicted to be about 1000 times larger than PGFJWestinghouse's prediction. I recommend that the NRC leakage rate prediction be used in analyzing these possible accidents, and that the leakage rate be assumed to be as large as rupture l
of a single tube, which is about 600 gallons per minute.
The supply of borated water available to replenish the primary coolant includes the 400,000 l
gallons m the refueling water storage tank, and about an equal amount available from other sources. Operator actions can reduce the leakage rate, conserve the available water supply, and bring the sequence to successful termination.
Greater attention should be paid, in my opinion, and plans developed, for ways to replenish the g
refueling water storage tank if it turns out to be needed. There is plenty of time and plenty of 4
water; what is needed is a source of boric acid, a place to mix it, and a way to get the solution into the tank. This does not seem too difficult to me.
I conclude that in spite of the tube degradation, the core melt probability for main steam pipe break accidents, estimated by the NRC at one in 1,000,000, is satisfactorily low.
For main steam pipe break event sequences that do not involve core melt, the calculated radioactivity release is proportional to the assumed tube leakage rate. Using the Westinghouse
]'
calculation, but the NRC leakage rate prediction, and my recommended assumption of an assumed leakage rate for one tube rupturing, the calculated radiation dose at the site boundary is unacceptably high. However, the calculated dose can be reduced by a modest lowering of the
]
allowable operating limit on radioactivity in the primary cooling water during normal operation.
I recommend that this be donec 2
Method of Anlysis Safety evaluation involves analysis of postulated sequences of events that could threaten the P
defense in depth provided for safety assurance. Rather than use the NRC design basis accidents, L
I have based this analysis on the methods of probabilistic safety analysis, in which both the probabilities and the consequences of the events are taken into account. The intent is to use realistic values of parameters and probabilities.
All nuclear plant accidents that can hurt people offsite or impact the environment begin with an initiatine event, which can be an operating mistake, a piece of equipment breaking, a fire or 3
.-r=--
u
severe storm or earthquake, a disturbance on the power grid away from the plant, or some other oCCurTCDCe.
The initiating event, whatever it is, creates a disturbance in the plant, whose temperatures, pressures, flows, etc., go awry. A secuence of events ensues. He plant trips off-or fails to I
trip when it should. The operators and the automatic controls are used to bring the plant into a safe, quiescent state--or fail to do so. De outcome is described as the _elant state.
r In any real accident sequence, the outcome depends on not only the type and severity of the initiating event, and the characteristics of the plant, but also the successes and failures of the operators and the systems as they attempt to manage the course of events. The outcome thus depends on the capabilities of the systems called on to function and whether these systems work when they are called on, and on the actions of the operators and the automatic controls to call on safety functions and systems.
Similarly, for the plant states that involve the release of radioactive materials out of the nuclear b
reactor primary system (or from other sources, such as spent fuel), the characteristics of the R
release, the response of the containment, the actions of the operators and the functioning of systems (such as containment isolation valves, containment cooling and containment spray) determine the extent to which the radioactive materials are controlled or released to the environment.
l Possible outcomes of different sequences of events include:
The potential accident is arrested without plant damage (investment protected) and a.
l without release of radioactive materials (public health and safety and environment protected). Almost all initiating events actually experienced have had this outcome.
I b.
The plant is damaged (investment not protected) but releases are prevented or limited to insignificant amounts (public health and safety and environment l
protected). He Browns Ferry fire and the Bree Mile Island accident are in this class.
f ne plant is damag:d (investment not piotected) and radioactive materials are c.
released (emironment not protected; public health and safety may not be
]
protected). De Chernobyl accident is in this class.
We can't know in advance what, specifically, will happen.' We can't predict which initiating g
event will occur, or when. We can't predict,if an initiating event occurs, what the operators will do and whether the systems will function. We do know which initiating events the plant is desiped for, what the operators are supposed to do, and how the systems are desiped to
]
function. He NRC safety requirements are framed in terms of plant desip basis (events to provide for), system desip basis (performance analysis, high quality, redundant components,
)
4 i
u=-
^
qualification of equipment, to enhance reliability of function) and operator qualification and training (to enhance likelihood of correct action). But events will occur how and when they happen, operators can perform well or poorly, equipment can function or fail.
In order to analyze plant safety and public or environmental risk in the face of all this uncertainty, the problem can be approached using pmbabilities. We don't know what will happen or when, but we can predict the frequency of events of different severity, the probability of the operators acting correctly or making mistakes, and the probabilities of systems functioning adequately or inadequately. 'Ibese probabilities, plus some complex calculations, yield the calculated probabilities of aniving at the various outcomes; that is, of protecting the public health and safety, the environment and the investment in the plant.
Steam generator tube integrity or leakage is significant in several possible accident scenarios--
event sequences. Some amount of tube leakage is often experienced during operation. Leakage is monitored during operation and if it exceeds Technical Specification operating limits, the plant must be shut down. Tubes must be inspected regularly or when the leakage rate gets too high, and repaired as needed. By itself, tube leakage within the operating limits has little safety significance.
For less probable events such as steam generator tube ruptures and main steam pipe breaks, estimates of the frequencies of occurrence are derived from experience and analysis. Estimates obtained from different sources are not always equal, showing the approximate nature of estimating infrequent and hypothetical events.
3 Steam Generator Tube Breaks 3.1 Initiating Event On several occasions worldwide, tubes have broken during operation, with leakage rates of several hundred gallons per minute, far larger than the operating limits. While each such event has been analyzed for its root cause, none of the actual occurrences was apparently " caused" by any operating event that happened at the time.
A 1988 NRC study gives the average frequency of steam generator tube ruptures during normal operation as one in 67 years for one tube ruptured, one in 1250 years for 2-10 tubes ruptured, 5
t and one in 50,000 years for more than 10 tubes ruptured.' Another recent (1989) NRC study gives the frequency of 1-tube ruptures during normal operation as one in 100 years.2 nese tube rupture frequency estimates are much higher than the estimates reviewed in section 4.4.2 of this report for tube rupture in steam pipe break events. He actual tube rupture events experienced in other plants have been caused by things that are unrelated to the Trojan tube degradation that is the subject of this report. Even though the degradation observed at Trojan has been going on for years, at several plants, it has not caused any tube ruptures. Therefore, I believe that the contribution of the recently discovered Trojan tube degradation should not significantly change the estimated frequency of tube rupture initiating events.
3.2 Sienificant Functions Recuired to Keep the Core from Melting De following are the safety functions required after a tube rupture or a steam pipe break, to keep the reactor core from melting.
Shut off the neutron chain reaction by inserting the control rods, stopping almost o
all reactor core power generation, leaving only (unavoidable) decay heat.
Reduce primary pressure as fast as allowable (keeping pressure high enough to o
provide core cooling and also avoiding reactor vessel overcooling) to prevent steam generator tube leaks from developing or increasing in size, and to decrease the flow rate through any leaks that do occur or were already present. For the tube break initiating event, the purpose is to minimize the flow rate through the break in the tube.
Replenish primary cooling waf.er lost through leakage.
o Maintain core suberitical, inhibiting neutron chain reaction and core power o
generation, by keeping the control rods inserted and the cooling water adequately borated.
o Remove heat from primary system NRC, NUREG-0844, September 1988, pages 3-19 through 3-21.
8 NRC, NUREG/CR-4550, Volume 7, Revision 1, September 1989, page 4-5.
2 6
33 Event Sequences ne following descriptions briefly outline the sequences of events that would ensue following this initiating event. Only the most significant sequences are described; there are many other possibilities, wherein the equipment or the operating team is unsuccessful in accomplishing one or more of the essential safety functions. In general, the probabilities of these failure paths are low.
The event sequence begins with the initiating event: One or more tubes break, with a spectrum of break sizes from small leaks (which have little safety significance), to large leaks, to complete rupture of one tube (such that primary water can flow unimpeded out of both ends of the broken tube), to rupture of multiple tubes.
The operations team must cool and depressurize the primary system to decrease the flow of primary fluid out the leak or rupture. The sequence is terminated when the primary system coolant has been depressunzed, cooled, and its level lowered below the tube leak, so the leakage flow stops. He secondary system holds the leakage fluid and pmvides containment, at least for a while. If the primary cooling and depressurization is successful soon enough, the leakage fluid is contained indefinitely; otherwise, some primary fluid eventually overflows the steam generator and goes to the condenser or out the steam relief valve.
3.4 Release of Radioactive Materials For event sequences that do not result in core melt, the release of radioactivity is negligibly
]
small. His has been true of the tube rupture events that have actually occurred.
Some core melt sequences would be predicted to result in substantial radioactivity releases.
However, the probability of occurrence of such releases is estimated to be very low. Two 1989 NRC risk studies give core damage probabilities for event sequences that begin with steam generator tube rupture as one in 500,000 years' and one in one in 800,000 years.' ' Core damage" in these risk studies is not necessarily core melt, and not all core melt events result in large releases.
In any case, since the tube rupture initiating event ikquency is not significantly affected by the present state of the Trojan tubes, and the event sequences are also not affected, I conclude that the risk from steam generator tube break events is not affected by the tube degradation experienced at Trojan.
NRC, NUREG-1150, Second Draft, June 1989, page 3-5.
NRC, NURUG/CR-4550, Volume 7, Revision 1, September 1989, page 4-69.
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Main Steam Pioe Break 4.1 Initiating Event In addition to a " spontaneous" tube break, a break or large leak in a main steam pipe is a i
possible initiating event where tube leakage would have a significant influence on the outcome of an event sequence. Such a large steam pipe leak could be the result of either an actual break in one of the 4 large steam pipes or the sticking open of one of the main steam safety valves or relief valves. Only the ponions of the pipes that are outside the containment,'um udutc wc wou
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steam isolation valves, are imponant to steam generator tube leakage, for reasons discusse( in section 43. The steam safety and relief valves are connected to this safety-signi5 cant ponion of the steam pipes.
This region of the main steam pipes is designed and inspected as "superpipe", just because a break here could bypass containment. At least three breaks have been experienced in steam pipes: Relief valve headers failed at H. B. Robinson Unit 2 and one of the Turkey Point units; the bypass steam pipe failed at Fermi Unit 2. All three failures were caused by design errors; all three occurred early in plant life. Trojan has operated for over 15 years; there is no reason to believe that the Trojan design has any problem.
NRC studies give estimates of the frequency of breaks plus stuck-open valves in the safety-significant ponion of one of the 4 Trojan main steam pipes occurring during normal operation as one in 1000 yeazs and one in 500 years.8 He very recent NRC re-evaluation gives one in 700 years.' This evaluation points out that a precursor of this event occurred at Trojan in 1984; a steam safety valve opened in a plant transient, stuck open for a while, and then closed.
4.2 Signiricant Functions Recuired to Keep the Core from Melting The required functions are essentially the same as for the steam generator tube break, see section 3.2. The event sequences are somewhat different, as can be seen in the next section.
i 1000 years: NRC, NUREG-OS44, Section 3.4,1988; 500 years: NRC, NUREG/CR-8 4550, Volume 7, Revision 1, September 1989, page 4-5.
NRC memo, E. S. Beckjord to T. E. Murley, January 15, 1993, Enclosure 2, (un-numbered) page 2.
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43 Event Secuences ne event sequence begins with the initiating event: A main steam pipe breaks in the region outside containment but inside its main steam isolation valve, or one or more of the large safety and relief valves connected to the pipe sticks open. Because the break or leak is ahead of the isolation valve, closing the isolation valve doesn't stop the flow of steam out of the system.
He pressure in the secondary system decreases as the steam escapes. For a large pipe brea the secondary system pressure falls from about 1000 pounds per square inch to below 200 poun per square inch in a few minutes. He primary system pressure, initially 2200 pounds per squa inch, decreases almost as quickly, initially, as a result of cooling via the steam generators and also by loss of water through any tube leak. However, the primary system pressure rises aga because the emergency core cooling systems replenish the primary coolant. If nothing intervenes, the pressure that the steam generator tubes experience (equal to the primary coolant pressure inside the tubes minus the steam pressure outside the tubes) therefore doubles its normal operating value after atout 1/2 hour.'
If the steam generator tubes leak significantly as a result of this increase in effective pressure, the leakage fluid flows from the primary system to the secondary system and out the steam pipe break, which is outside contamment. He fluid so lost can constitute a release of radioactivity, and also cannot be recirculated to replenish the p:imary fluid needed to cool the reactor core.
He propensity of the tubes to leak will be decreased, and the flow rate through any leak that develops will be decreased, by lowering the primary system pressure as much and as fast as allowable. The pressure must be maintained high enough to cool the reactor core, and the rate of decrease must be limited to avoid thermal shock to the reactor vessel. He recen calculations show that the effective pressure difference seen by the tube starts at the normal operating value of about 1300 pounds per square inch, and that the eventual maximum pressure can be reduced from 2600 to about 1800 pounds per square inch by operator action, without jeopardizing the core or the reacer vessel.'
He primary system pressure must also be maintained higher than the secondary system pressure if there is any steam generator tube leakage, as long as there is any unborated secondary system water in the steam generator above the leak. Such unborated water must be prevented from being injected into the primary system. He primary system water must be borated to keep the reactor suberitical when it cools down, even with the control rods inserted.
Eventually, if significant tube leakage continues, the source of borated water for replenishing the primary system coolant-the refueling water storage tank-will become depleted, and core NRC memo, E. S. Beckjord to T. E. Murley, January 15,1993, Enclosure 3, Figure 4.
7 NRC memo, E. S. Beckjord to T. E. Murley, January 15,1993, Enclosure 3, page 2.
cooling water injection will stop unless another source of borated water has been provided in the meantime.
If the supply of borated water runs out and core cooling is interrupted for too long, the core will melt. A leakage path for radioactivity will exist from the core, via the primary system piping, through the leaks in the steam generator tubes and out the broken steam pipe, bypassing containment.
The operator can influence how long the supply of borated water will last, by (1) cooling the primary system using the undamaged steam generators, with injection of auxiliary feedwater and opening the steam relief valves in the undamaged main steam pipes; (2) controlling and limiting the pressure of the injected borated water to limit the pressure seen by the tubes, and so limit any tube leakage, and (3) conserving borated water by controlling the flow of borated water being injected into the primary system to just the amount needed to cover the reactor core and maintain cooling.
The sequence can be terminated by switching the core and primary system cooling function to the residual heat removal system, which can function with the primary system water level lowered below the level of the tube leak (s). In order to accomplish this, the primary system temperature and pressure must be reduced, by cooling the system, to allow the switchover in cooling, and then the water level must be lowered. This takes time, and cannot be speeded up too much without overheating the core or overcooling the reactor vessel.
Clearly, the parameters that control whether the core melts are the steam generator tube leakage rate, as compared to the available supply of borated water, and the time required to reduce the primary system temperature and pressure, switch cooling modes, and lower the primary system water level.
Radioactive materials that escape from the broken steam pipe (or stuck-open valve) are released outside the containment and constitute a potential hazard to the plant staff and the public.
4.4 Tube Leakage Rate This is the core of the controversy.
4.4.1 Criteria Before 1991 As a result of much experience, testing and research, the nuclear industry and the NRC developed inspection methods for steam generator tubes and criteria to determine when tube repair is required. In general, degradation that results in an effective tube thickness less than 60% of the original thickness (such as a crack deeper than 40%) requires repair, usually plugging or sleeving.
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One of the NRC studies previously cited gives probabilities for tube rupture as a consequence of a steam pipe break event as one in 40 (for 1-10 tubes ruptured), and one in 2000 (for more than 10 tubes ruptured).' These probability estimates were based on the strength, leakage and rupture propenies of steam generator tubes as perceived when the repon was developed in 1988.
4.4.2 Additional Degradation Discovered in 1991 Tbc corrosion that was discovered in the Trojan steam generator tubes in 1991 occurs on the outer tube surfaces, at locations wherb the tube is inside the holes in the tube suppon plates.
Each steam generator tube is U-shaped, with the bend on top and the ends of the " arms" anchored in the tubesheet, which is a thick piece of metal with holes in it for the tube ends. De tube arms also pass through holes in 7 horizontal tube suppon plates spaced over the length of the tubes.
De region between the outside of the tube and the inside of the support plate hole is a crevice, open at the top and bottom. Impurities in the secondary system water tend to concentrate in these crevices, sometimes a million times more, compared to their concentrations in the water outside the crevices. Some of these concentrated impurities are believed to cause the Trojan corrosion.
The corrosion process results in two different crack patterns: (1) Many small " axial" parallel cracks oriented venically, along the axis of the tubes; and (2) a pattern of cracks that looks (under a microscope)like an irregular mosaic. Some corroded areas look like one pattern or the other; some look like a mixture of the two.
It is believed that this corrosion process started before 1986. The cracks grow slowly with time, and are inspected periodically.ne growth rate has been estimated from in-plant inspection data and the results of laboratory examinations. larger cracks are detectable with conventional eddy current inspection probes. The new rotating pancake coil eddy current probe found many cracks that the conventional probe is unable to detect, thus cracks are detected by the rotating pancake coil probe at an earlier stage of the corrosion process.8" He existing, 40% criterion for tube repair was judged not to be applicable to the newly discovered cracks. Newly developed " interim plugging criteria" were approved in early 1992.
Tbc NRC safety evaluation associated with approval of the new criteria states that 428 Trojan tubes have known flaws left unrepaired at the end of the 1991 outage, and that these may have maximum depths exceeding the old 40% limit.8' NRC, NUREG-0844, pages 3-21 and 3-22, September 1988.
Westinghouse Trojan Dbe Repair Criteria, Repon WCAP-13129, Rev 1, pages 5-11/12.
NRC Safety Evaluation, pages 5 and 11, February 5,1992.
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Westingbouse states that,in spite of the known degradation of the Trojan steam generator tubes beyond what was known in 1988, a steam pipe break event would result in very small tube leakage, less than 1 gallon per minute, and that the probability of tube rupture in a steam pipe j
event is less than one in 30,000.
It is obvious that the current Westinghouse view is that the 2
tubes are stronger today--much less likely to rupture in a steam pipe break event-than the NRC estimated in 1988. Tbc Westinghouse estimates use correlations based on data on the measured burst strength and leakage of tubes removed from Trojan and other plants, degraded as they are j
today, plus laboratory tests on arti5cially degraded specimens.
Dr. Hopenfeld and Mr. Muscara say that the probability of significant tube leakage following a steam pipe break is higher for Trojan tubes with known through-wall cracks and other cracks deeper than 40% than the earlier estimates for tubes without these defects. They state that this leakage will shonen the available time the operators have to control the plant and organize an additional source of borated water. The shorter available time will increase the chance of operating errors and failure to make more borated water available in time, and thus increase the probability of core melt in steam pipe break event sequences.
The NRC Staff based its early 1992 acceptance of Trojan operation, for Cycle 14 only, with the presently xnown degradation and allowance for growth during this cycle, on detectability of large cracks, margin in tube repair criteria, high measured strength of degraded tubes, and tightened operating leak rate limits." A later memo from the NRC Office of Research, Division of Engineering, reaches conclusions similar to those in the NRC Staff Safety Evaluation, based on similar reasoning, with the added basis of the low probability of a steam pipe break."
Two very recent (January 1993) NRC Office of Research memos reiterate their earlier evaluation, give the technical bases for their evaluation, and respond specifically to Dr. Hopenfeld's arguments."
'Ite recent NRC memos contain estimates of steam generator tube leakage rate following a main steam pipe break. The leak rate estimate for Trojan *D" steam generator is somewhere between Westinghouse, WCAP-13129, Revision 1, pages 10-11 and 10-14, December 1991.
12 NRC, Safety Evaluation Related to Amendment No.178 to Facility Operating I.icense No. NPF-1, February 5,1992.
NRC Memorandum, L C. Shao to E. S. Beckjord, " Interim Plugging Criteria for Trojan Nuclear Plant," December 9,1992.
NRC memos, E. S. Beckjord to T. E. Murley, " Interim Plugging Criteria for Trojan Nuclear Plant", January 5 and 15,1993. Enclosure 3 to the January 5 memo is " Division of Engineering [ Office of Research) Responses to Comments of J. Hopenfeld."
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33 and 1350 gallons per minute, with a "best estimate" value of 145 gallons per minute." The evaluation is for the full effective tube pressure of 2600 pounds per square inch, with no credit for operator action to reduce the pressure. Any one of the four steam generators would give approximately the same results". De main steam pipe break or stuck-open valve is assumed to involve only one steam generator.
De hTC "best estimate" leakage rate is almost 1000 times higher than the Westinghouse estimate, and seems to me to more reasonably represent the probabilities and uncertainties than the Westinghouse value. In fact,I suggest that 600 gallons per minute, the leakage flow rate for one tube mpturing as a result of a main steam pipe break, should also be considered in evaluating the safety of Trojan restart, as well as the NRC "best estimate value of 145 gallons per minute.
De NRC Office of Research memos cited just above also contain a calculated core melt probability of one in 1,000,000 years, for all event sequences initiated by steam pipe breaks or stuck-open steam safety valves." He sequences included in this estimate include the various possible failure paths, with their probabilities. His core melt probability is satisfactorily low, in my opinion.
4.4.3 Additional insoections in late 1992 During the forced Trojan outage to repair the leaking sleeve that began November 9,1992, PGE performed additional inspections of some tubes in the steam generator that had to be opened to repair the leak. Inspections included the other sleeves, to confirm the adequacy of the 1991 repairs, and some tubes. Additional tube inspections suggested that a different pattern of corrosion might be present. Still more inspections were performed, including ultrasonic probes as well as eddy current probes.
Westinghouse reported that the patterns of corrosion were not new, but resembled tubes pulled earlier from other plants, with acceptable measured burst strength. His information was reported in a series of telephone calls from PGE and its contractors to the NRC and State of Oregon representatives, ne data and analysis were to be documented in a submittal from PGE to the NRC scheduled for January 8, but the January 4 PGE announcement terminating Trojan operation forestalled this, nerefore, we have only the information transmitted in the phone calls.
" NRC memo, E. S. Beckjord to T. E. Murley, January 15,1993, Enclosure 1, section 6 (no page numbers).
" Ibid., section 2.
" NRC memo, E. S. Beckjord to T. E. Murley, January 15,1993, Enclosure 2, page 3 (unnumbered). The January 5,1993, memo gave a slightly higher core melt probability estimate, one in 700,000 years.
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j If the documentation had confirmed the analysis transmitted orally in the phone calls, I would have recommended that the additional infonnation on Trojan steam generator tube degradation r
does not denote significant additional deterioration. Derefore, the 1992 steam generator tube data do not indicate that restart is unsafe.
4.5 Release of Radioactive Materials 1
Of course, the reason that reactor accidents are of concern is the possibility of releasing
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radioactive materials out of the plant; to the potential detriment to the health and safety of the public and to the environment. For steam generator tube degradation, the potential impet A from any increases in either the amount of radioactivity predicted to be released or the probability l
of signi5 cant releases. Two sets of accident sequences were analyzed: (1) Core melt accidents; j
(2) steam pipe break events where the core doesn't melt.
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]I 4.5.1 Core Melt Accidents l
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- l De principal risk to the public from reactor accidents arises from those event sequences where I
the reactor core is predicted to melt. For core melt sequences, the resulting hazard depends on the functioning of the containment and the systems provided to cool and control the fluids in the containment space.
J Not all core melt accident sequences create actual public hazards. We now know that about 40%
of the reactor core at nree Mile Island was melted during the accident in 1979, but the radiation hazard to the public from that accident was negligibly small. The successful functioning of the i
containment and the resumption of cooling after 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> kept the radioactivity released from harming the public and the environment.
De effect of the degradation of the steam generator tubes at Trojan on the probability and consequences of core melt events is small. Any effect would be on the probability of tube rupture or leakage, either as an initiating event or as a possible consequence of a steam pipe break sequence. In section 4.4.2, above, I recommend using a tube Icakage rate based on the NRC analysis, which is much larger than the Westinghouse value, for evaluating steam pipe break events. However, the calculated core melt probability for such events is satisfactorily low, I
fn my opinion.
For main steam pipe break event sequences involving high tube leakage rates, the ability of the operating staff to replenish the refueling water storage tank if needed might be critical to preventing core melt. PGE procedures identify some other potential on-site sources of borated water that roughly double the 400,000 gallon capacity of the refueling water storage tank.
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Greater attertion should be paid, in my opinian, rad plans developed, for ways to replenish the refueling water storage tank if it turns out to be needed. Tuere is plenty of time and plenty of water; what is needed is a source of boric acid, a place to mix it, and a way to get the solution into the tank. His does not seem too difficult to me.
4.5.2 Evenir that do not Melt the Core Event sequences that do not melt the reactor core can still result in the release of radioactive materials. He potential sources of such releases are: (1) ne water and steam initially in the secondary system, some fraction of'which escapes immediately. His fluid has very small radioactivity content. (2) ne primary cooling water, flowing through the tube leak into the steam generator and thence out the steam break. His fluid has a larger radioactivity content during normal operation. (3) If the core melts, a much larger au - nt of radioactivity is released into the primary system, from where it can leak into the secondary system and escape out the steam break. De radioactivity released depends on how much fluid is released and how large its radioactivity content is. De hazard depends on how much is released, the pathways to people and the environment, aad the protective steps taken.
For steam pipe heak event sequences that do not involve core melting, any significant radioactivity relear vill originate in the release of primary system coolant through any tube leakage path into the secondary system and out the break or stuck-open steam safety valve.
Westinghouse gives an analysis of the relationship between the tube leakage rate in a steam pipe break event and the calculated radiation dose to a person located at the plant site boundary. The Westinghouse calculation is reported to give 30 rem to such a person's thyroid for a leakage rate of 100 gallons per minute. My evaluation in section 4.4.2 concludes that analysis of steam pipe break accidents should make allowance for tube leakage or rupture much larger than the Westinghouse analysis. He NRC "best ertimate" leakage rate is 145 gallons per minute; I recommend also considering 600 gallons per minute, which is typical of one tube rupturing. For my suggested assumed leakage rate of 600 gallons per minute, the applying the Westinghouse calculation would give 180 rem.
i I recommend that the operating limit on allowable radioactivity in the primary cooling water be lowered accordingly, to limit the calculated dose for such an event. In plants where the primary coolant radioactivity is high during operation, its principal source is minute failures in the metal cladding on the fuel, allowing the radioactive gases generated in the fuel during operation to leak into the cooling water. He fuel in Trojan the past few years has had little such leakage. I 1
believe that the tighter limit I am proposing would not limit Trojan operation significantly.
Westinghouse, WCAP-33129, Revision 1, December 1991, page 9-7.
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5 Uncertainties and Marrins There are many uncanainties in any analysis of improbable accidents, such as the ones discussed in this repon. The use of available data, often sparse, and calculated plant responses and event probabilities results in known and unknown errors in the calculated results. Applying such results to safety analysis should include allowance for such uncenainties.
Both POE (with Westinghouse) and the NRC have added margins for uncenainties in, for example, the Interim Plugging Criteria and the assumed tube leakage rate in steam pipe break events.
I I have recommended an additional margin in the assumed leakage rate, leading to a reduced allowable primary coolant radioactivity during operation. In addition, I have recommended an additional " margin" of a different kind in recommending development of provisions for replenishing the supply of borated water, if it were ever to be needed.
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In addition to the uncenainties in all accident analysis, there are substantial uncenainties in the detection, measurement and prediction of flaws in steam generator tubes like the ones found at Trojan. The history of the discovery, measurement and analysis of the actual flaws at Trojan, j
and how our present understanding developed with time, illustrates my point. He Westinghouse analysis shows a wide scatter of the available data." These uncertainties appear to be a good
.I part of the basis of Dr. Hopenfeld's and Mr. Muscara's difficulties with the present hTC position.
I believe that the Trojan flaws are not very well characterized, and that there are substantial uncenainties in the current Westinghouse analysis. To this extent, I agree with Dr. Hopenfeld and Mr. Muscara. However, PGE and the NRC have recognized this, and have provided large margins, particularly in the flaw size for which repair--plugging or sleeving the tube--is required. De high burst strengths measured in corroded tubes removed from Trojan are a principal source of the margin provided in the Trojan interim plugging criteria.
My recommended " margins" in assumed leakage rate, operating limit on primary coolant radioactivity, and provisions for replenishing the refueling water storage tank, are in addition to PGE's and the NRCs margins..
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" Westinghouse, WCAP-13129, Revision 1, December 1991, Figures 5-2,5-11, and 6-7; other examples are proprietary.
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