ML20096A619

Rebuttal Testimony of B Erler Re Stokes Allegations Concerning Evaluations of Discrepancies in Calculated Actual Stress Performed by Sargent & Lundy.Related Correspondence

ML20096A619
Person / Time
Site:	Byron
Issue date:	08/30/1984
From:	Erler B SARGENT & LUNDY, INC.
To:
Shared Package
ML20096A623	List: ML20096A619 ML20096A626 ML20096A639
References
OL, NUDOCS 8408310192
Download: ML20096A619 (10)

Text

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= UNITED ~ STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

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'BEFORE THE~ ATOMIC SAFETY AND LICENSING BOARD sMF In The Matter-of )

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COMMONWEALTH EDISON COMPANY ) D kd 00s AI@,k454-OL

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' (Byron Nuclear Power Station, ) yfGCE OT HCy r3 2 s Units 1 & 2) ) -"HT M 4 39 ," ;-

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- Rebuttal Testimony of Bryan Erler Q.l. Please state your full name and placelof employ-ment for the record.

A.l. Bryan A. Erler, Associate and Structural Design Director, Sargent & Lundy, 55 East Monroe Street, Chicago, Illinois 60603.

Q.2. Please describe your job responsibilities.

A.2. As Structural Design Director I am responsible for the overall coordination and management of four of Sargent &

Lundy's Structural Divisions. These divisions are: the

' Structural Engineering Division; the Structural Engineering Specialist Division; the Structural Drafting Division; the Architectural Design Division. These Divisions have the responsibility for preparation, review and approval of all l Structural design engineering calculations and civil /

architectural / structural drawings.

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4 Q . 3. . ^Please describe you educational'backgrcund and work experience.

A.3. I graduated from Purdue University with a BSCE-in 1969, and an MSCE.in 1970. I have 14 years' experience in-the field of civil / structural / architectural engineering and design of fossil and nuclear plants. I am a. registered Structural-Engineer in the State of Illinois. I started as

.a Structural Engineer at Sargent &'Lundy in 1970.. I worked-on the containment design of several nuclear power plants

. including Zion Units 1 and 2, Fermi Unit 2, Zimmer Unit 1, and LaSalle Units 1 and 2. In January, 1973, .I was promoted to the position of' Supervisor of Special Structures Section, and'in July, 1973, I was promoted to Assistant Chief, I

Structural Engineer Specialists, responsible for the contain-ment design and seismic analysis of several nuclear power i plants, including Byron Units 1 and 2, Braidwood Units 1 and 2, and Clinton Unit 1. In 1976, I was promoted to Chief

) Structural Specialist, responsible for all contain~'nt and seismic analysis at Sargent & Lundy. In 1977, I was promoted i to-the position of Head, Structural Design and Drafting Division, responsible for all structural design engineering calculations and civil / architectural / structural drawings at Sargent & Lundy. In 1979, I was appointed Associate in the firm and in-1982 was promoted to my current position of Structural Design Director. I l

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' *; l I am presently a member of the American Concrete Institute, the American Society of Civil Engineers, and the A

. Post-Tensioning Institute. I also serve on the-following two technical committees:

ACI/ASME Joint Technical Committee on Concrete Pressure Components for Nuclear Application (the National ASME Code committee responsible for developing criteria for design of reinforcsd and prestressed concrete containments and prestressed' concrete reactor. pressure vessels in ASME Section III Div. 2); and ACI-348, Structural-Safety (responsible for establishing appropriate safety margins for reinforced and prestressed concrete structural designs and provides input for the ACI-318 design group).

Q.4. What is the purpose of your testimony?

A.4. I have undertaken to address various allegations by Mr. Stokes relating to evaluations of discrepancies performed by Sargent & Lundy. Specifically, I will confirm that the calculated actual stress of the discrepant items fell within the allowable stress limit of the American  !

Institute of Steel Construction (AISC) Code. I will also report on the inspection of welds on neighboring connections for the cable tray that contained a cracked weld that confirms the validity of a particular Sargent & Lundy judgment.

My testimony also discusses various allegations by Mr.

Stokes with respect to flare-bevel groove welds, fatigue ,

loading on pipe supports, and use of the AWS Code, AWS Dl.1-83, for evaluation of welding discrepancies.

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Mr. Stokes-in his testimony' states that in his l I

' review of Sargent & Lundy evaluations, he'found instances r.

- where Code stress.allowables appeared to be exceeded.

Did the evaluations of the Hatfield, Hunter and Systems Control Corporation. visual weld discrepancies performed by Sargent &

- Lundy reveal any instances where the actual stresses . exceeded the-AISC' Code allowable?

e A ." 5 . No. The evaluations' revealed that none'of the Hatfield, Hunter or Systems' Control Corporation.. visual weld

- discrepancies resulted.in actual stresses exceeding the-allowable stress limit of the AISC Code.

Q.6. Mr. Stokes in his testimony states that a "10%

overstress factor" was used by Sargent & Lundy.during intermediate steps of the calculations conducted to disposition various weld discrepancies under the Reinspection Program.

Could you explain Sargent & Lundy's use of this factor?

A.6. The 10% overstress factor refers to a 10% limit where'Sargent & Lundy engineers are allowed to use their

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knowledge of'the structural analysis to decide, when the

- calculated stress is less than or equal to 10% greater than allowable, that the calculated stresses have sufficient conservatisms in them to meet the AISC Code stress allowable.

In the case of the Reinspection Program, some of the initial calculations of the Hunter and Hatfield weld discrepancies showed an overstress less than 10%. This

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.was not relied upon, however, when the calculations were 1 i

I refined. In no instance was the 10% overstress factor applied in' the final determination that the capacity of the various connections met allowable stress limits.

Q.7. Mr. Stokes alleges in his testimony that it appears that Sargent & Lundy's judgments and evaluations fe'11 short of the degree of objectivity and impartiality required of an independent review. Have any judgments or evaluations been performed by Sargent & Lundy to verify that a particular judgment was appropriate?

A.7. Yes. I have reviewed a number of judgments and evaluations performed by Sargent & Lundy to verify that its judgments were appropriate. For example, a judgment was made by Sargent & Lundy with respect to the evaluation of a crack in a cable tray holddown weld. This discrepancy was one of the 187 discrepant Hatfield welds included as part of the sample when, in response to NRC questions, additional inspections were made of welds not initially covered by the Reinspection Prog::am. This discrepancy was evaluated by assuming the cracked weld had a 100% reduction in capacity. The calculation involved transferring the load to the other weld on the cable tray support and to four welds on neighboring connections.

It was determined that the other welds could easily sustain the additional load transferred from the cracked weld because their calculated stress was very low compared to the i AISC allowable. This determination was based on a judgment l l

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. $ that-the' neighboring welds were in a nondiscrepant condition.

LThis judgment was verified by a recent inspection. ,

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. l Q . 8. - What was the result of the inspection?

i A.8. No~ discrepant--conditions-were present in any of

-the welds.

Q.9. Mr. Stokes testified to concerns about flare-bevel w'elding at Byron. Were flare-bevel' groove welds included in

-the Byron reinspection program and, if so, how many contained-

' discrepancies that required calculations for evaluation?

A.9. Flare bevel groove welds were captured.by th'e reinspection program. The discrepancy evaluations performed by Sargent & Lundy included 30 flare-bevel AWS welds produced by Hatfield Electric Company.

Q.10. Were the tubes to which these welds were made inspected for a determination of the actual radius?

A.10. Yes. An inspection was performed of each of the tubes. The measurement yielded radii at least two times the tube wall thickness (2T) for all tubes except one which had a radius equal to 1.75T. The stress of each weld was conservatively evaluated using the AWS formula for effective throat of 5/16R with the smallest R measurement of 1.75T.

This demonstrated that the AWS allowable stresses were met.

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Q. ll. - Were these welds produced under a qualified' procedure?

=A.ll. Yes, these welds were produced by Hatfield welders

-under-qualified procedures.

'Q.12. Mr. Stokes testified that flare bevel groove welding was included in a Hatfield prequalified welding procedure designated as 13AA. How-do you reconcile this testimony with your previous answer?

A.12. The Hatfield AWS flare bevel welds' captured in the Byron Reinspection Program were produced during the period May, 1978 through September, 1982. During that period of time, flare bevel groove welds were produced under qualified procedures 13Q and 13AB. Procedure 13AA, a prequalified welding procedure, was approved on December 30, 1983, and flare bevel groove welding was erroneously included in its procedure. This is being rectified and the procedure for flare bevel groove welding is being issued as a qualified procedure.

Q.13. Mr. Stokes opines in his testimony that fatigue ,

I loading should have been considered with respect to pipe  !

I supports. Was fatigue loading considered in the analysis of  !

pipe support loading?

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A.13. Yes. Fatigue loading was considered in the analysis of these welds. However, in accordance with the AISC, it is

not necessary to reduce the. allowable stress in a weld for fatigue loading-until.the number of stress cycles exceeds 20,000.- The number of stress cycles experienced by pipe

~ supports at' Byron is substantially less than 20,000.-

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I Q.14. Is there an inconsistency between fatigue requirements i

L for piping and those for pipe supports? j i

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.A.14. No. Both piping and supports require. consideration of fatigue. Due to the nature of-loading on a piping I system, the requirements may' vary depending on the class of the system. For example, a Class 1 system requires explicit calculation for the piping while Class 2 and 3 piping are

.affected by cyclic loading only if the number of cycles exceeds 7,000 (ASME Section III NC 3611.2) . For pipe supports with respect to Class 1, 2 and 3 piping, both ASME and AISC are consistent in not requiring any reduction in allowable stress for less than 20,000 cycles.

At Byron, for Class 1 piping systems, the analysis has accounted for the number of cycles as required by the code. For Class 2 and 3 piping systems, the number of cycles experienced is less than 7,000. Accordingly, no reduction in allowable stress for fatigue is required. The supports are subjected to less than 20,000 cycles and, consequently, no allowable stress reduction due to fatigue is required in the design of the supports.

Mr. McLaughlin, as a structural engineer, was guided by the 20,000 cycle requirement of the AISC Code, and

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1 he properly concluded in his testimony that convexity weld 1

discrepancies could be neglected because structural elements, such as pipe' supports, erected in Byron did not experience 20,000 on and off loadings. Similarly, as a mechanical engineer, Mr. Branch properly conducted a fatigue analysis for Class I piping in accordance with the ASME Code. These are consistent with. standard design practices. Mr. Stokes' contrary viewpoint is erroneous.

Q.15. Is a waterhammer loading on a piping system a

' loading which could cause a fatigue problem?

A.15. No. Waterhammer is a dynamic pulse loading with low frequency of occurrence. Therefore, the number of stress cycles is extremely low and fatigue is not a problem.

Indeed, Attachment 10 to Mr. Stokes' testimony clearly states that very fact.

Q.16. Mr. Stokes expresses a concern about the fact that Sargent & Lundy's evaluations of weld discrepancies were performed pursuant to the AWS Dl.1-83 Structural Welding Code while the welding was performed pursuant to earlier editions of the code. Do ycu believe the use of two editions of the AWS code present the concern articulated by Mr.  !

I Stokes?

A.16. No. AWS Code Dl.1-83 was used for design assessment of the discrepancies in the reinspection program. It should initially be pointed out that with the exclusion of the 1

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. year- 1978, a revised version of the AWS Code has been published every year from-197 5 to-the present. The design requirements

. have not changed significantly.since the, issuance:of AWS Dl.1-75, which was the Code in effect at the- time of initial

- construction. The allowable stresses are the same. The few changes that-have been made with respect to calculation of stresses have all been more restrictive;with regard to weld 1

capacity. These stricter weld design requirements-in no way require less-demanding-calculations for evaluating a' discrepancy.

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If anything, it is conservative to use the-latest edition of AWS Dl.1 for evaluation of discrepancies.

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