ML20031H473
| ML20031H473 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 09/18/1981 |
| From: | Gordon G GENERAL ELECTRIC CO., HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20031H319 | List: |
| References | |
| NUDOCS 8110270517 | |
| Download: ML20031H473 (4) | |
Text
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September 18, 1981 i
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3
4 In the Matter of S
S HOUSTON LIGHTING & POWER COMPANY S
Docket No. 50-466 5
5 (Allens Creek Nuclear Generating S
6 Station, Unit 1)
S 7
DIRECT TESTIMONY OF DR. GERALD M. GOkDON REGARLING:
(1)
TEXPIRG CONTENTION 10 - IGSCC 8
(2)
DOHERTY CONTENTION 43 - STAINLESS STEEL CLEANING (3)
DOHERTY CONTENTION 44 - IGSCC/ WATER HAMMER 9
e 10 Q.
Dr. Gordon, have you reviewed your prior affidavit 11 on TexPirg Contention 10, which affidavit is attached hereto 12 as Attaetteertt.sMG--1?
13 A.
Yes, I have.
Q.
Are the statements contained in the affidavit 14 still true and correct?
15 A.
Yes, they are.
Q.
In its answer cpposing Applicant's motion for summary disposition o." this contention, TexPirg alleges that IGSCC is still an unresolved safety issue.
Is this allegation 19 correct?
20 A.
No.
NRC Task item A-42
(" Pipe Cracks in Boiling 21 Water Reactors") is in fact resolved through issuance of 22 NUREG-0313, Revision 1 (July 1980), which has been imposed 23 This resolution is stated on the cover and upon the ACNGS.
in the introduction to the document.
Resolution was 8110270517 810918 PDR ADOCK 05000466
_y, PDR T
1 possible through the generally accepted technical adequacy 2
of the solutions for piping stress corrosion, as used in the 3
ACNGS, endorsed in NUREG-0313, Rev. 1.
Other additional solutions are under development for earlier BWR nuclear
,4 stations, which did not have the benefit of the improved I
I materials used in the ACNGS, and will be reviewed by the NRC
,s I when they become available as stated in NUREG-0313, Rev.
1.
~
n Regulatory Guide 1.44 is consistent with NUREG-0313, Rev.
1.
3 (July 1980) in recommending the low carbon stainless steel o
materials used in the ACNGS.
~
10 Q.
In its order of September 1, 1981, the Board 11 raised a question as to whether GE has considered " superposed 12 loads" and " excessive oxygen" levels.
Would you please 13 address these points?
A.
The stainlesp, steel and carbon steel materials used 3
14 in the piping systems at the ACNGS have been thoroughly and 3_n successfully tested under oxygenated water conditions (6 parts per million dissolved oxygen) far more aggressive 17 than the dissolved oxygen levels encountered during operation 18 of the ACNGS (0.2 ppm 0 ).
The applied loads during these 2
1~9 tests exceed the maximum permitted superposed design loads 20 for actual field piping installations.
These tests have been 21 able to reproduce field cracking, which normally requires 22 years of service exposure, in 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of test time in 22 actual piping weldments fabricated from high carbon Type 304 i
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1 stainless steel.
The successful test exposure of numerous P Ping specimens with the material composition used at the i
2 ACNGS in over 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of testing under the same highly 3
4 eggressive (6 ppm 0 ) conditions represents adequate statis-2 tical assurance of expectation of several hundred reactor 5
years of service without cracking in the field.
Furthermore, 6
higher carbon gradee of Type 316 stainless steel (which 7
rep, resent less margin to cracking than tho Type 316 Nuclear g
Grade, low carbon piping used in the ACNGS) have performed without cracking for over 20 years in one domestic Boiling Water Reactor.
Low carbor, steel has also performed successf*lly 11 for many years on many applications.
Thus, the ACNGS stainless 12 steel piping material composition has shown by long term 13 tests and successful related field experience that it repre-sents resolution to piping stress corrosion.
15 Q.
Finally, the Board questioned the statement in the 16 affidavit of Messrs. Gunther and Malec to tie effect that leak 17 detection provides assurance against rapidly propagating 18 cracks.
Do you know of any documentation which substantiates 19 their conclusion?
20 A.
Yes.
Their conclusion is in fact substantiated by 21 the NRC in the introduction to NUREG-0313 (July 1980).
Furthermore, with reference to the issues raised by both 22 TexPirg Contention 10 and Mr. Doherty's Contention 44, 23 studies have been performed on the effects of overpressure 24
_ i e
1 loads and water hammer type loads on IGSCC in sensitized 2
steels.
These studies have substantiated the conclusion 3
that leakage will occur before complete break.
4 Q.
With regard to Doherty Contention 43, and the 5
remaining question at page 52 of the September 1 Order, will 6
any of the ECCS components supplied by GE be stainless steel and if so will they be coated with compounds which could 7
contribute to IGSCC?
g A.
No coatings of any type will be used on stainless g
/
steel components in the ECCS supplied by General Electric.
General Electric's stainless steel cleaning requirements are in accordance with Regulatory Guide 1.37, and Applicant 12 commits in the PSAR to follow Regulatory Guide 1.37.
Some 13 carbon steel equipment will be coated, but carbon steel is-14 not susceptible to chloride-assisted stress corrosion 15 cracking.
16 17 18 19 20 21 22 23 24
AttachInent GMG-1 f!
UNITED STAT 55 0F AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD a("
In the Matter of HOUSTON LIGHTING & POWER COMPANY
)
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)
Docket No. 50-466 1'_.T '
)
(Allens Creek Nuclear Generating
)
7 Station, Unit No. 1)
)
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AFFIDAVIT OF DR. GEP.ALD M. GORDON l
n Ik State of California County of Santa Clara
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I, Gerald M. Gordon, Manager, Plant Materials, within the Engineering and Technology, Nuclear Power Systems Engineering Department of the
..g General Electric Company, of lawful age, being first duly sworn, upon my 4
cath certify that the statements contained in the attached pages and e
accompanying exhibits are true and correct to the bdst of my knowledge and belief.
ii, Executed as San Jose, California Julyff,1980.
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Subscribed and sworn to before me this.29 day of July,1980.
$w An
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NOTARY PUBLIC IN AND FOR SAID COUNTY AND STATE My ommission expires 2n M M of 19f/_.
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AL SEAL i
v RUTHE M. KINNAMON NOTARY PUBUC - CAUFORNIA
$ANTA CLARA COUNTY My comm. expires MAR 25, 1981 175 cwne" ^
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7/29/80
Attachment GMG-1 t;
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION n
[.1 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD h
In the Matter of S
e; 5
HOUSTON LIGHTING & POWER COMPANY S
Docket No. 50-466 f'f; T
S (Allens Creek Nuclear Generating S
Station, Unit 1)
S E
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AFFIDAVIT OF DR. GERALD M. GORDON in My name is Dr. Gerald M. Gordon.
I am employed by the General Electric Company as Manager, Plant Materials
- ^
Engineering and Technology, Nuclear Power Systems Engineering Ei Department.
I have been so employed for two years.
A a
statement of my experience and qualifications is set out in N
Mf For the purposes of clarity, the following defini-I5 tions apply to this affidavit:
a.
Austenitic Stainless Steel.
Austenitic Stainless Steel (nominal 18 percent chromium and 8 percent nickel
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compositions being the most popular) are steels that possess an austenitic (face centered cubic atomic) structure at room temperature and are non-magnetic. _Austenitic stainless ~
steels are generally either normal carbon (0.08% maximum) or low carbon (0.03% maximum).
The low carbon versions are referred to as L-Grade.
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b.
Sensitization.
The intergranular precipitation of 9
chromium carbide and resultant depletion of chromium level in I
austenitic stainless steels when exposed to temperatures in 1,2/
g the approximate range of 800* to 1500*.
This depletion i
leads to decreased corrosion resistance at the grain boundaries.
c.
Intergranular Stress Corrosion Cracking (IGSCC).
Cracking occurring preferentially at grain boundaries re-i sulting from a special combination of stress, material
'g condition and environments.
Sensitized austenitic stainless r
steel is a material condition where IGSCC has occurred in L
BWRs.
I d.
Type 316 Nuclear. Grade Stainless Steel.
A classi-
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fication of steel alloys that has approximately 16 percent i
f chromium, 10 percent nickel, 2 to 3 percent Molybdenum, less than 0.02 percent carbon, and some other minor alloying ele-ments, with the remainder being iron.
Type 316 Nuclear Grade does not sensitize when welded and therefore is not susceptible
.to IGSCC under BWR conditions.
e.
Plain Carbon Steel.
A classification of non-stainless steel alloy that has approximately.2 percent carbon cnd.5 percent manganese.
This class of alloys has not exhibited stress corrosion cracking in contact with BWR f
coolant.
f.
Heat Affected Zone.
A region of base metal on L
either side of a weld which is heated above 800'F during I
L L S
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welding.
In materials subject to intergranular stress
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corrosion this results in sensitization.
sensitization of normal carbon content (0.05-g, 0.08%) stainless steels occurs in those places where pipe I
welding has heated a narrow band of the material to an elevated temperature and the material has been allowed to
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cool slowly without a subsequent heat treatment.
These g
heat-affected zones are thus " sensitized."
When these sensitized areas were exposed to a particular combination of f
stress and dissolved oxygen in high temperature waters, there zones have in the past shown some susceptibility to L
stress corrosjon cracking at grain boundaries.
Through
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July 28, 1980, only 209 out of about'34,000 stainless steel i
pipe weld heat-affected zones within the Reactor Coolant Prec$ure Boundary (RCPB) have experienced IGSCC in all operating BWRs.
Of these, the bulk have been in the recirculation m
bypass line, the core spray line, control rod drive hydraulic return line, and reactor Vater clean up lines.
Counter measures have been identified and qualified for these lines in ACNGS.
For example, the recirculation bypass line and the control rod drive hydraulic return line were eliminated from the ACNGS design and the core spray line and reactor water clean up lines were changed from normal carbon stainless
- f steel to plain carbon steel.
Plain carbon steel has not L '
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exhibited stress corrosion under BWR conditions.
The remainder of the RCPB piping is either plain carbon steel or Type 316 f
Nuclear Grade stainless steel.
The specific material changes in stainless steel g
piping for ACNGS are the result of programs suggested by.a I
special interdisciplinary General Electric Task Force investi-gation conducted in 1975 to determine the cause of cracki.g i
in stainless steel piping lines.
Potential improvements were identified and extensively tested.
Only after being g
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proven were they implemented into ACNGS and other plants.
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More recent changes made to the ACNGS plant to further avoid IGSCC incidents include the use of feedwater-L spargers and collet cylinder tube and recirculation pump
[
housing which are made of low carbon stainless steel and Control Rod Drive Housings which are fabricated from Type 316L stainless steel.
ACNGS is designed to eliminate the occurrence of 3
C IGSCC.
The most direct and certain solution to eliminate i
't-the potential for IGSCC in BWR piping, as recominended by the NRC,-5/ is the use of materials resistant to stress corrosion.
I l-To that end, all of the RCPB will be composed of materials j
not imbject to IGSCC.
The NRC Staff established in Regulatory L
5/
Guide 1.44 and NUREG-0313 the criteria for testing and i
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fabricating austenitic stainless steels to minimize the incidence of IGSCC.
These criteria are not applicable to l
the very low carbon grades of austenitie stainless steels and plain carbon steel because these materials have been i
shown to resist IGSCC.
This immunity can be produced because the most significant factor affecting the degree of sensitization is the carbon content of the alloy.
Stainless steels with a maximum of.03% carbon are essentially immune to IGSCC in a BWR environment.
At great expense, Applicant has specified that all stainless steel material in the recirculation system be made of the most impervious material available--316 Nuclear Grade stain-less steel.
This material may not have a carbon content exceeding.02 percent which is even lower than the carbon level of 316L (.03% maximum).
Hence, the entire recirculation system can be considered immune to IGSCC.
In conclusion, the materials used for ACNGS piping supplied by General Electric comply fully with 10 CFR 50, Appendix A, Criterion 31, and reduce the potential incidence of intergranular stress corrosion cracking to virtually nil.
The NRC has reviewed the substitution of materials described above and accepts this alternative as a resolution for the generic problem.
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References 1.
G.
E. Linnert, " Welding Metallurgy," American Welding
,g Society, 1949.
2.
American Society for Metals, Metals Handbook, Volume 1.
" Properties and Selection of Metals," Eighth %Jition, l
1961.
3.
"The Application of Low Carbon Type 316 Stainless Steel
,(
for BWR Recirculation Piping Systems," J. F. Copeland and E.
D.
Sayre.
Paper to be _ presented at Symposit
" Material-Environment Interactiona in Structural anc Pressure Containment Service," organized by the Metal Prtderties Council, Inc., in cooperation with the Materials Division of ASME, to be held during'the 1980
. 'l ASME Winter Annual Meeting, Chicago, Illinois, November 16-21, 1980.
4.
" Mitigation of Stress Corrosion Crasking in Boiling
[
Water Reactors" by.9. E. Hanneman and R. L. Cowan II.
Paper presented at the American Power Conference, Chicago, Illinois, April 1980, to be published in
{
proceedings.
5.
" Technical Report on Materici Selection and Processing
[
Guidelines for BWR Coolant Pressure Boundary Piping,"
,L NUREG-0313, Rev. 1 (October, 1979).
i i
6.
" Investigation and Evaluation of Stress-Corrosion L
Cracking in Piping of Light Water Reactor Plants,"
NUREG-0531 (February, 1979).
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ATTACHMENT 1 i
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GERALD M. GO.RDON 1
PRESENT POSITION Manager, Plant Materials Engineering and Technology, Nuclear Power Systems Engineering Department, General Electric Company.
I EDUCATION r
B. S. - Metallurgical Engineering, Wayne State University,1956
- i Ph.D. - Metallurgical Engineering, The Ohio State University, 1959 TECHNICAL ASSIGNMENTS t
Prior to joining General Electric, Dr. Gordon was a Senior Metallurgist at Stanford Research Institute, Menlo Park, California, from 1959-63.
He s p ied as a Project Leader on a number of anvernment and comercially sponscred programs L
in the areas of high temperature oxidation and mechanical performance of refrac-j m y metal alloys,
'1 Dr. Gordon joined the General Electric Company Nuclear Energy Division in 1964 as a Senior Metallurgist in the Reactor Materials Development Group at Vallecitos.
i He became Manager of the Metallurgy Development Component in 1969. This group had j
i' materials research and development responsibility for physical metallurgy, fracture toughness and radiat'on damage of reactor materials and aqueous corrosion and stress cracking of nuclear reactor pressure bcundary and internals materials.
1 [-
In 1973, Dr. Gordon became Manager of the Zircaloy Performance Group with respon-sibility for development and evaluation of nuclear fuel cladding and channel f~
materials.
He also served as Manager, Plant Component Behavior Analysis and j
was responsible for implementation of laboratory developments in design of reactor plants. He assumed the oosition as Manager, Plant Materials Engineering in 1976, and his current positior, as Manager, Plant Materials Engineering and Technology I
in 1978, and is currently responsible for evaluating and specifying BWR plant
'L materials as well as materials surveillance and identification and solution of potential o' actual stress corrosion cracking problems.
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Dr. Gordon is a Registered Frofessional Engineer in California and a Fellow of the American Society of Metals. He has authored numerous publications and patents and has been an invited lecturer or Session Chairman at several International Co'nfer-l e ces on Corrosion & Stress Corrosion Cracking. He is past Chairman of the National Association of Corrosion Engineers Committee T-ll A on Corrosion in High Purity Power Plant Water.
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