ML20246J094
| ML20246J094 | |
| Person / Time | |
|---|---|
| Issue date: | 05/04/1989 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Glenn J SENATE, GOVERNMENTAL AFFAIRS |
| Shared Package | |
| ML19316F857 | List: |
| References | |
| NUDOCS 8905160292 | |
| Download: ML20246J094 (3) | |
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. UNITED STATES
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' NUCLEAR REGULATORY COMMISSION pg j' '; V.-d g.
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May'4, 1989 I
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The Honorable John'Glenn, Chairman Comittee on Government Affairs United States Senate Washingten, D.C.
20510
Dear'Mr. Chairman:
~
On January 20, 1988, Chaiman Zech sent a letter to you responding to your
~1etter of; December'2, 1987, regarding' concerns over the use of the ferritic stainless steel alloys, specifically E-Brite 26-1, in comercial nuclear power plants.. The letter included a statement' that the NRC staff would develop an information notice to alert the industry to the E-Brite 26-1 issues.
The staff completed its review of the allegations concerning E-Brite 26-1 and concluded that E-Brite 26-1-is acceptable for use.in comercial nuclear power-
' plants when the design and material requirements of Sections II and III of the American' Society of Mechanical Engineers Boiler and Pressure Vessel Code are satisfied.
In addition, E-Brite 26-1 was found installed at only one plant and the material is performing satisfactorily at that plant..Because of these conclusions, the NRC staff does not intend to issue an information notice.
A sumary of the actions taken by the NRC staff in reviewing the acceptability of-E-Brite 26-1 is enclosed.
If you have any question concerning'the staff's review or the decisior, not to issue an information notice, please contact.me.
Sincerely, r
. ;/s
., /
Vfd[or Stello, Jr. C'[ /?dhj,
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Executive Director j
for Operations L
Enclosure:
i Sumary of Staff Review on E-Brite 26-1 cc. Senator William V. Roth, Jr.
L FULt. TEXT ASCil SCAN 80 n
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SUMMARY
OF STAFF REVIEW ON E-BRITE 26-1 In June 1986, the NRC received an allegation that E-Brite was a defective material because it could not be welded and its ductile-to-brittle transition temperature could not be controlled, thus making its use in nuclear power plants hazardous. The NRC staff undertook a number of actions to adoress this issue.
In August 1986, the staff met with the alleger to discuss his concerns and review any other available, pertinent information. From discussions with the alleger, it was determined that the alleger's concerns stemmed from the failure of several tanks fabricated for use at a military painting facility.
Investiga-tions conducted by the Army Corps of Engineers showed that the failures were the result of poor welding practices.
In September 1986, the staff reviewed records pertaining to E-Brite metal manu-factured by Airco Vacuum Metals (AVM) to determine whether any of the material was shipped to nuclear power plants in the 1970's as stated in the allegation.
It was deterniined that AVN did not supply E-Brite to a nuclear facility.
In May 1987, after furtter correspondence with the alleger-in which he indicated that fraud might be involved with the test data used to obtain Code approval of E-Brite, the staff notified the American Society of Mechanical Engineers (ASME)
The of the allegation and requested that ASME review the approval of E-Brite.
ASME Subcommittee on Properties met in February 1988 and accepted the position of the Subgroup on Strength of Ferrous Alloys that E-Brite was suitable for establishing design stresses and for use under the ASME Code (see attached letter dated March 8, 1988).
13, 1988, the Subcommittee on Properties issued a statement that a On September review of the notch toughness data for E-Brite provided to the ASME Code Connittee in 1971 showed that the data was representative of suitably heat treated base metal in the product thickness described and of suitably welded joints in such products (see attached letter dated November 3, 1988).
In July 1987, the issue of falsification of data to obtain AShE Code approval The 01 review disclosed was referred to the NRC Office of Investigations (01).
no evidence of fraud concerning test data submitted to ASME.
In December 1987, the NRC staff learned that Arkansas Power and Light (AP&L) installed E-Brite material in two shutdown cooling heat exchangers at. Arkansas huclear One (AN0) Unit 2 during a refueling outage in 1981.
In January 1988, the staff conducted an inspection of the supplier (the Tubular Products Division of Allegheny Ludlum (AL) in Claremore, Oklahoma) to review the quality assurance records and evaluate the process used to fabricate the E-Brite tubing for the heat exchangers at ANG Unit 2.
Activities associated with the fabrication of the E-Brite tubing met NRC requirements. The staff also conducted an inspection in March 1988 at the ANO Unit 2 site to review the procurement and operational history associated with retubing of the shutdown cooling heat exchangers with l
E-Brite material.
1
. l The results of the-inspections conducted at the AL and the AP&L facilities indicated that the procurement and manufacture of the E-Brite tube bundles met NRC requirements. Successful tube-to-tube sheet welds also were made during l
the retubing at At10 Unit 2; the welds indicated that E-Brite has adequate weld-ability for the f abrication of heat exchanger tube bur.dles. One of the tube Sundles sustained mechanical damage during installation. Although some of the damaged tubes that were repaired experienced some leakage in 1983 and were repaired again, the absence of leaks in either of the heat exchangers since 1983 provides additional assurance about adequacy of this material. Therefore, the alleger's assertions that E-Brite could not be welded and that its ductile-to-brittle transition temperature could not be controlled were not substan-tiated in the ANO Unit 2 application.
The staff determined that E-Brite material is acceptable for use in safety-related applications in nuclear facilities when the design and material requirements of Sections II ar.d III of the ASME Code are satisfied.
Attachments:
1.
Letter from ASt1E, dated March 8, 1988 l
l 2.
Letter from ASME, dated flovember 3, 1988
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ATTACHMENT 1
' The American Society of MechanicalEngineers Codes and Standards 345 East 47tt Street Pressu'e Tech 9 ology Ne* Yore NY 1D017 212 705 7087 March 8, 1988 Mr. J. Stone Acting Vender Branch Chief USNRC Washington, DC 20555
Subject:
Review of 26Cr-Mo Ferritic Stainless
Reference:
ASME Files BC87-144 & SCP 87-37 Your file RIV-66-A-06 NRC, Herschoff letter of April 21, 1987 NRC, Merschoff letter of May 26, 1987
Dear Sir:
The NRC letter of April 21, 1987 informed ASME of an allegation that the data used to obtain Code approval misrepresented the subject material.
In response to that letter we initiated a review of the issue and are providing the action taken by the Subcommittee on Properties.
The following is an excerpt from the minutes of the Subcommittee on Properties meeting of February 9,1988.
Those minutes are subject to approval and are for committee use only.
6.
SUBGROUP REPORTS A.
SG on Strength of Ferrous Alloys M. Gold reported 1.
SCP 87-37 (Review of 26Cr-1Mo Ferritic Stainless Steel)
BC87-144 (Section III, Case 1505, & I-1.0, 2.1, 7.0)
(Section VIII, Case 1496, & UHA-23)
A formal review and reanalysis of this alloy was reouested by ASME staff in response to an allegation that data supplied in the original ASME evaluation misrepresented this material.
3 C's,
C A summary of the reanalysis is shown below.
a L
The original test data (3 heats) was analyzed using the polynomial regression method which resulted in ratio trend curves for.both yield and tensile strength which were identical to the curves which resulted from the original analysis.
(
A second set of data (5 heats, all of which contained approximately 0.10% Cb) were similarly analyzed. The new analysis resulted in a ratio c'
trend curve for yield strength identical to that developed from the 3
first set of data. The trend curve for tensile strenth developed from this data diverged fron, that derived from the first set of data at 400*F, because the second set of data was from material with 0.10% Cb added, which reduces strain aging behavior.
Mr. J. Stone March 8, 1988 Page 2 Conclusion i
l The two sets of independently supplied test data' support virtually identical ratio trend curves, except for the reduced strain aging behavior apparent in the second set as a result of the 0.10% Cb addition. At temperatures above 400'F, the allowable stresses developed from the first set of data diverged from those developed from the second set of data by less than 5%.
As it is our convention to not revise allowables if a reanalysis develops differences of less than 5%, no proposal will be made to revise allowables.
tiotion: To accept the position of the Subgroup on Strength of Ferrous Alloys as stated below:
" Review of the data submitted in 1971 for the 26Cr-1Ho steel shows that it was suitable for establishing design stresses and rules for use under the Code.
A review of a secona set of data in 1984 from a different producer confirmed the original analysis."
This motion was seconded and approved unanimously.
M. Gold reported that the SG on Strength-Ferrous Alloys had determined that the tensile data displayed no unusual or unexpected features, and that a review of data from material' produced by a second manufacturer corroborated the conclusions drawn from analysis of the data originally provided.
End of minutes excerpt.
The issue of representative toughness data for 26 Cr-1Mo ferritic stainless steel will be considered at a future meeting.
Yours truly.
A
^
John Millman Secretary, SC-P (212) 705-7026 JM/km cc:
D. Canonico J. Ling D. Doty W. Mikesell W. Knecht
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