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EQ24' Are thereLother areas you.are. concerned with? | EQ24' Are thereLother areas you.are. concerned with? | ||
JA24: Yes. InLan NRC letter dated January 30, 1984 to Common- | JA24: Yes. InLan NRC {{letter dated|date=January 30, 1984|text=letter dated January 30, 1984}} to Common- | ||
: wealth Edison, on page 11 there was a referer ce to.an allegation ' received by the. NRC regarding undersized welds where tube steel was'used. That document is attached to my testimony as Attachment 8. I do not believe that allega-tion was substantiated at'the time. Also attached to my | : wealth Edison, on page 11 there was a referer ce to.an allegation ' received by the. NRC regarding undersized welds where tube steel was'used. That document is attached to my testimony as Attachment 8. I do not believe that allega-tion was substantiated at'the time. Also attached to my | ||
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Latest revision as of 18:13, 24 September 2022
| ML20094P599 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/16/1984 |
| From: | Stokes C AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| OL, NUDOCS 8408170296 | |
| Download: ML20094P599 (36) | |
Text
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- REMiTEDC M 5 0$ -b UNITED STATES OF AMERICA .
NUCLEAR BEGULATORY COMMISSION , 00CMETED UDE In the Matter of: )
) Docket No. 50-454 OL
~ COMMONWEALTH EDISON COMPANY ) 50-455 OL '84 AGO 16 R2 d6 (Byron Nuclear Power Station, ) "
~ '
Units 1 and 2) ) - DIRECT TESTIMONY OF CHARLES C. ' STOKES ON BYRON REINSPECTION PROGRAM Q1: Please state your name and present employment. A1: My name is Charles Cleveland Stokes. I am a nuclear engineering consultant. My partner and I do consulting work in the nuclear construction industry under the name P/S Associates. Q2: Please describe P/S Associates. A2: P/S Associates is a newly formed firm that offers consulting services to those differing entities involved in the nuclear construction industry. Our initial work has been on behalf of intervenors in NRC proceedings relating to the Diablo Canyon and Byron nuclear power plants. Q3: Please describe your job responsibilities. A3: I am a technical consultant in utility production facilities from the standpoints of design calculation, code and federal requirements, and quality assurance. I also design and review existing structures and mechanical components. Q4: Please describe your education and professional background. 8408170296 84 {DRADOCM050 1 ' D(;)
r _ y
, o. .. .
f' L A): -Ilam ~ a member of the National Society of Professional y Engineers =and a registered Professional Engineer in Alabama, t [ i Florida and Georgia. I' began my' career as a co-op student employee ~with the Civil-Structural Department of Southern [ Company Services from November 1972 to June 1975. I was
' assigned various duties, primarily those'of a draftsman and detailer for the Fossil-Hydro Concrete, Structural Steel, and Nuclear Concrete Departments.. I worked on Farley Nuclear Plant Unit 1 and 2 miscellaneous outdoor concrete structures and HVAC, 'and Miller Steam Plant concrete base slabs. I detailed reinforcing and bolted connections.
I graduated in May 1975 with a BCE degree from Auburn University. I took a job as an Assistant Engineer with l Southern CompanyLServices Civil-Structural Department. In this position, I designed outdoor structures on Miller Steam Plant, anc designed sub-structure concrete and checked super-structure calculations on Harris Dam. I held this position from June 1975 to July 1987, when I was promoted to an Engineer II. In this position, I was Civil Material Coor- ! dinator and designed miscellaneous items on Vogtle Nuclear Plant, performed 79-02 and 79-14 analysis on Parley Nuclear Plant Unit 1, and redesigned the precipitator structural steel on Miller Steam Plant, designed structural steel for l coal conveyor on Schereer Fossil Plant, and wrote two speci-fications (one for modifications to the Reactor Heat ; Discharge System on the Hatch Nuclear Plant). f I held this position from July 1978 to May 1980. [ 2 , l
o . In May 1980, I graduated from the Birmingham School of Law with a Juris Doctorate degree. At this time, my resume was submitted to Bechtel Power Corporation' o'vil Struc-tural group in Gaithersburg, Maryland working on the Davis-Besse Nuclear Plant. I was accepted and worked from June 1980 until October 1980, performing 79-02 and 79-14 calcula-tions. I then worked from October 1980 to May 1981 for the Nuclear Services Corporation, a division of Quadrex Corpora-tion, on the Zimmer Nuclear Plant, during which time I was also assigned to the LaSalle Nuclear Plant. In June 1981 I . began work for the Mechanical Engineering Department of the Lawrence Livermore National Laboratory as the stress analyst on the injector of the Advanced Test Accelerator (ATA). I also made some design changes to the Experimental Test Accelerator (ETA). I left the Livermore Laboratory in February 1982 to work for Reactor Controls, Inc., on Grand Gulf Unit 1 Control Rod Drive System. In November,1982, I was accepted by Bechtel to work on Diablo Canyon Nuclear Plant. I worked on Unit 1 until March 1983 when I was assigned to Unit 2. In October 1983, I was laid off two weeks after writing three Discrepancy Reports against both Units 1 and 2. (See my resume, Attachment 1.) 05: Are you familiar with the Byron Reinspection Program? A5: Yes. I have spent more than 300 hours reading and reviewing o Commonwealth Edison documents, NRC documents, Sargent & Lundy documents both at their of ficen and my of fice, weld 3
r procedures of Hunter and Hatfield, and miscellaneous other documents on the Byron facility. I have reviewed documents reflecting many of the engineering evaluations performed by Sargent & Lundy. I have also reviewed the prefiled testimony of Edison witnesses Branch, McLaughlin and French, in addition to listening to portions of their testimony and the testimony of other Edison witnesses at the Atomic Safety and Licensing Board hearing on Byron in July of 1984. I have reviewed Edison's February 1984 Reinspection Report and the June 1984 supplement thereto, as well as engineering packages and other proprietary design documer.ts obtained from Sargent & Lundy. In my review I have reworked a number of engineering calculations and have cross-referenced formulas used by Sargent & Lundy to perform their evaluations. I have spent some time reviewing parts of the Byron FSAR, various codes, NRC NUREGS, and researching formulas in the design criteria of Sargent & Lundy or which were referenced in Sargent & Lundy calculations on discrepancies found in the Reinspection Program. Q6: What is the purpose of your testimony? A6: My testimony addresses the engineering evaluations performed and the use of engineering judgment by Sargent & Lundy in its attempt to show that there are no safety-significant constauction problems at Byron. The main purpose of my O testimony is to suggest the need for an independent engi-neering analysis of the safety significance of the problems found in the Reinspection Program, as well as an independent 4
analysis and examination of certain hardware at Byron where evidence indicates possible safety problems. My testimony is not intended to show conclusively that Byron is not safe to operate. A nuclear power plant is a 1arge and complex facility requiring extensive time and resources for a conclusive engineering assessment; even an assessment limited to the Hatfield, Hunter and PTL discre-pancies found in the Byron Reinspection Program, together with certain Systems Control Corporation discrepancies dis-cussed in Edison's pre-filed testimony, would require far more time than I have had, as well as a range of engineering skills and experience including but not limited to those which I possess. However, even in the limited time I have had to review Sargent & Lundy documents relating to the engineering i evaluations, I have seen numerous indications of issues which, in my judgment, collectively require further exploration and resolution before there can be reasonable assurance that Byron is safe to operate. For this reason, the purpose of my testimony is to suggest that there are enough signs of possible safety problems at Byron, and enough legitimate concern about Sar-gent & Lundy's engineering calculations and use of engineer-ing judgment to require an independent engineering analysis of the discrepancies found in the Reinspection Program, and of certain other hardware issues, prior to a determination by this Board on whether there is reasonable assurance that Byron can be operated safety. 5
F My testimony also includes instances in which I am 1 unable, based on the limited review I have conducted, to
]
agree with Sargent & Lundy's calculations or with its , evaluation of safety significance. In addition I discuss miscellaneous hardware issues, which in my engineering judg-ment, based on the limited documents I have reviewed, do not appear to have been properly dispositioned. Lastly, I com-ment on those areas in which I perceive signs of possible problems, but as to which, due to time constraints and/or incomplete documentation, I have not yet reached any further conclusion other than my opinion that further review is warranted. Q7: Why do you believe an " independent" engineering analysis is needed? A7: A letter from Chairman Nunzio J. Palladino of the Nuclear Regulatory Commission to the Congress (Attachment D to the proposed prefiled testimony of intervenors' witness Dr. William II. Bleuel) specifies criteria for an independent design review of Diablo Canyon. In that letter, Mr. Palla-dino indicates that an independent design review must be done by an entity not previously involved in the activity under review. In my opinion, whenever the review is signi-ficantly judgmental, independence is needed. Q8 Do you believe Sargent & Lundy's engineering evaluations were significantly judgmental? A8: Yes. As illustrated in various instances cited throughout 6
.g .
, )
r.- .... the' remainder of my ' testimony, an analysis of the findings t of.the reinspection program for safety and design signifi- i
- cance, which is what Sargent, & Lundy did, is comparable to a design analysis in.the degree of' judgment required.- *-
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IQ9: In your opinion, why was not Sargent & Lundy's' analysis
" independent"?
A9: For a number of reasons.. Sargent & Lundy has been and-remains the' architect-engineer on Byron from the plant's
' inception to the.present date, in addition to'being involved in a number of other Edison projects, such as the Braidwood 1
plant. Sargent & Lundy thus has a. direct economic stake in the outcome of the evaluations.- Moreover, my review of the firm's evaluations reveals repeated instances in which in my opinion, based on the limited documents I reviewed, it appears that Sargent & Lundy's judgments and evaluations fell short of the degree of objectivity and impartiality required of an independent review. Sargent & Lundy has evaluated thousands of discrepant conditions and yet'did.not find one single item to be safety-significant. As I reviewed a limited portion of Sargent & Lundy's calculations I found instances where the allowable stress
. appeared to exceed code requirements. I also found instances in which certain elements, minor by themselves, appear to have been omitted, non-conservatively,- from calculations,. including instances where if these factors had i
7 L _. __m..-._______._...___._____.__________._m__-_______________-_m___ _ _ - _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _
.* s been added, it appears that actual stress would have ex-ceeded allowable stress. I also found instances where it appeared, on the basis of the calculation itself, that the .
equipment would fail. Yet the Sargent & Lundy evaluations with respect to each of those instances concluded that , there was no safety significance. Q10: What did your review of the design criteria (Sargent & Lundy's Structural Project Design Criteria BYRON AND BRAID-WOOD Nuclear Power Station Units 1 &2 (DC-ST-03-BY/BR) REV.
- 12) for Byron reveal?
A10: The design criteria were one the standards guiding calcula-tions of design significance in the engineering evaluation of Heinspection Program discrepancies. In my review I found instances in which formulas appeared to be incorrect, and instances in which equations appeared to be missing elements. I also found instances of design assumptions that in my opinion are faulty and should not have been relied on in the design of the plant, being used as refer-ences in Sargent & Lundy evaluations of discrepancies found in the Reinspection Program. In sum, I found a number of problems in the design criteria for Byron. Q11: Can you elaborate on the problems you found in the design criteria? Alls, Attachment 1 to my testimony is a non-exhaustive summary of what appear to be erroneous formulas, questionable design assumptions, and apparently faulty equations I found in my l 8 l 1 i I
n b Lbrief. review of the~ design. criteria for Byron. Following p L that summary . in Attachment 1 are copies of'those sections-f of'the design criteria that are listed in the summary. , Point 1 -in the summary is .section 12.2.4. In this_ L ' formula the lambda symbol appears to be missing. This is i-important because these formulas should be correct. l Although some. engineers might know the correct formula without having to see it, others may not. 'If the wrong L formula is utilized, faulty design may result. In 12.2.4 the formula is P AE = 1/2( )ll 2 ovk AE. l ! The section of the design criteria where this apparent error is contained is o otherwise so thorough that someone relying on it might not check the formula in another book. If the criteria had omitted the formula and simply referenced it, that would have caused to engineer to check the formula thereby finding the correct one. That would have been the better alternative. By making that section so thorough, then leaving out the lambda symbol, the likely result is that the formula may be used correctly. This omission bears on the inferences about the plant which may be drawn from the Reinspection Program. This
- sect. ton of the design criteria relates to below-grade l
structural building outside walls - which were not included in the Reinspection Program and could not have been, because they are generally inaccessible and under the ground. Thus any error due to this omission would not have been detected in the program. 9 i f
7 Q12: What are the other problems you find in the design criteria? - A12: The second point in my summary refers to Section 19.5.d. The equation appears to be missing a summation symbol ! before the b2 (squared). Without the summation symbol it is not obvious that'one should sum up the totals.before b 2, It would make a significant difference in the calculation if such summation were not done. This equation relates to l the summation of torsional stresses for the concrete turbine foundation - another example where any resulting error would not have been detected in the Reinspection ! Program, which did not tear up and reinspect concrete. l l Also in Section 19.5.d the equation, I believe, should l be the square root of F prime C. This is important because l l of the significant difference between the square root of the number and the number itself. Particularly, Section 19.5.d talks about allowable stresses. If the calculation is made l without the square root, the result would be a higher l allowable stress. Section 32 3 2 also has an apparent error. It says 25 fy when it should be .25 fy. This section relates to buried piping. Again, any resulting error would not have ( been detected in the Reinspection Program, which treated l buried piping as inaccessible. l Q13: Isn't that the kind of error an engineer would recognize l0 right away? 10
A13: one would think so, but then why did no one correct the design criteria? Q14: Are there other apparent errors you have found in the design criteria? A14: Yes. Number 4 in my summary is Section 32.4.2., which also relates to buried piping. Spangler's equation appears to l be listed as D.061, whereas it should be 0.061. Also, I believe that R 4 should be R3 in the denominator. I do not know whether anyone took the D to mean diameter and used I the diameter in the calculation, but this appears to be a problem. The difference between using R 4 as a denominator as opposed to R3 is substantial. Although my review has l been limited by time the apparent errors just discussed in l l addition to others listed on the summary in Attnehment .1 l suggest the presence of defects in some of the formulas Sargent & Lundy has employed at Byron. 015: When you stated that there are design assunptions that in your opinion appear t,o be faulty, to what asnumptions did you refer? A15: Section 34.2 of the design criteria states that embed plates are designed for 10 kips per foot tension load and 12 kips per foot ahear load. I have attached Section 34.2 as Attachment 2. Q16: What do you think the problem in with this section? 0 A16: This a major concern that I have from my discussions with Sargent & Lundy peopic and from what I saw in the field 11
f While on the site visit. It appears that Sargent & Lundy , has procedures to hang conduit, HVAC pipe supports, both small and large, off embed plates. If they hung every-thing off embed plates, as I understand Sargent & Lundy to say, there could be serious safety problems. I believe there are legitimate doubts as to whether the embed plates would survive a seismic event. These would affect, for example, Hatfield conduit supports, and !!unter pipe supports, which may be hung from embed plates, but again, the embed plates themselves were not, reinspected in the l Reinspection Program, and inadequacies in t. hem thus may have gone undet,ected. Q17: Please explain why embed plates might not, survive a seismic event.. A17: Ten kips is, I believe, too small for the design of the l plates. For example, on the field trip I saw a 12 inch l line t, hat had a st. rut to the embed plate on the wall. If t,h e re is a large load on the strut, as 1 observed in the field (the st, rut nppears to have 15 or 20 kips based on t,he size of it), and 1, hat load is applied to t,he pinto, the bolt, strengt,h can be exceeded and the whole plate can pull off t,he wall. If that occurred throughout, t,he pi n n t, pro-bloms could be widespread. If a large number of embed plates are questionable, t.he plant could not undergo n safe shutdown earthquake. Q18: llow do you know that every embed plitte in designed to 10 l and 12 kipn? 12
A18: I do not; however, that is what this document appears to say. I did not find any calculations for embed plates, but this document appears to stato that all embed plates are designed in this manner. Q19: Do you have other concerns with the design critoria? A19: Yes. In Section 37.2.1, reinting to mechanical component supports such as !!unter pipe supports, there is a listing s of design offects that are to be ignored when performing calculations, for examplo, toratonal stressoa, axini soir weight, and assumptions that all masnes are lumped at, the shear cent.or. The stressos are nmall, but they are non-conservativo. If those stresson woro added t,o the cnicula-tions some of thom, I believo, would fail. This name problem occurred at, Diablo, namely, ignoring minor but non-connorvativo design effoots. At Diablo the NilC required that, all the minor st.ronnen bo included and all cnicula-t,lons where ninor stronnes woro ignored be rocniculated. I have included a summary of thono proceduros and the proco-duros thonsolven as Att.nchment, 5 to my test,1 mony. I l 020: Do you have other concerna with the design critoria for l Byron? A20: Yon. I lint.ed the major concerna above, but I have many other concerns that I have not had nn opportunity to review in dept,h. Attachment 6 to my tont,1 mony lint.n thone potential problem nronn I have encountered with t,he Byron dontgn criterin nnd my initint concerns. It may be that t.hore nro annworn to ny concernn with the proceduron listed 13
in Attachment 6, but my questions have not been resolved by my review of the documents thus far. Q21: What other documents have you reviewed that you have concerns with? A21: One of the documents is entitled Seismic Subsystem and Equipment Response Spectra Design Criteria Byron and Braid-wood Nuclear Power Stations Units 1 and 2 DC-ST 04 BB REV.2 Copy 48. This document is attached to my testimony as Attachment 7 It relates to buildings and would affect calculations for each component in the building, including Hatfield, Hunter and PTL inspected components; Reinspection Program calculations relied on this document. In Section V.B. the document states, "The horizontal seismic model of the nuclear power plant complex involves many degrees of dynamic freedom; theoretically a response spectra could be generated for each degree of freedom in addition to the horizontal model, a separate vertical 3 del was developed for the vertical direction of excitation, so additional degrees of freedom for which response spectra could be generated ',:ere introduced into the analysis." (Em phasi s mine) It appears that only one model was done for both the horizontal EW and NS, even though the building cross sec-tions were different and only one vertical model was made. For each building about its respective center of gravity six (6) seismic loadings should have been applied: acce, lerationsalongthreeaxisesandmomentsaboutphosethree axises (Fx, Fy, Pz, Mx, My and Mz). In the spectra, only 14 l J
l l accelerations for EM, NS, and vertical are shown. At least one other spectra per building should have been made, one for the torsional acceleration to building components at their radial distances from the center of gravity. For 'f many component.s, this torsional component is significant on loads, stressos, and ultimately to a conclusion as to safety of the plant. Q22: '4hy is this important at Byron? A22: At Byron it appears that the torsional component is being ignored. In every plant I've worked in, the torsional component has never been ignored. I might add that neglect of the torsional component is consistent with the FSAR, Section 3.7.2.11, where torsional effect was listed as insignificant. Nowhere else where I have worked has the torsional effect been considered insignificant. Now this was apparently detected by the NRC in 1982. The NRC apparently did not altogether approve of the practice, but did find it to be in compliance with the FSAR. Q23: In your opinion, is it reasonable to assume that at Byron the torsional effect is insignificant? A23: If all of the items in tile building are within 10 feet of the center of gravity, then it probably is insignificant. But many of the buildings have items as f ar away as 50-60 feet from the center of gravity of that building. For example, the turbine building is a long rectangular building, in which the torsional effects are likely to be substantial. 15 m
7 7 ,. . r s ,e
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EQ24' Are thereLother areas you.are. concerned with? JA24: Yes. InLan NRC letter dated January 30, 1984 to Common-
- wealth Edison, on page 11 there was a referer ce to.an allegation ' received by the. NRC regarding undersized welds where tube steel was'used. That document is attached to my testimony as Attachment 8. I do not believe that allega-tion was substantiated at'the time. Also attached to my
^
testimony as Attachment 9 are the Sargent & Lundy calcu'la-- tions for the weld survey project. Page110 of that Sargent
& Lundy document, titled Flare Bevel Groove Welds, states,
" Typical field measurements indicate that the actual radius is between T and 2.5T, where T is the tube wall thickness.
Therefore, the design assumption of R = 2T and effective throat equal to 5/16 R per. AWS is not applicable." This quote in my opinion appears to substantiate the NRC allega-tion contained in Attachment 8. The document-itself (Attachment 9) raises further questions. It was' prepared by one D.J. Sheahan. The document contains no apparent indication that it was ever checked or approved. It has no page numbers, no calculation number, and no book number. The only other thing that helps one trace it is on the ; fifth page where.there appears the name "D. Patel - 28." I looked up this name on a list provided to intervenors in discovery and found that he or she is one of the structural leaders at Sargent & Lundy for the reinspection program. ,
.i 1
In my opinion.what Attachment 9 suggests is that a design assumption of R = 2T is not valid. However,-I am 16
m
- . y not- certain what design: assumption' was in fact used. l l
Again,, this same problems was prese~ntlat the L Diablo plant.
~
j WeLused the design assumption of R = 2T and found i't was ,
.: wrong.
Attachments -8 and 9 thus potentially bring into ques-- tion every weld-to tube steel in the plant. It may be the: case that large numbers of' welds of this type are as much-as 50% deficient. .This problem would'potentially affect, for example, .Hatfield conduit and cable tray and Hunter
. pipe supports.
Q25: Do you have any concerns relating to welds reinspected in the Reinspection Progam and repaired prior to engineering evaluation? A2S: In my review, I came across a document labeled Reinspection' Program entitled Daily Inspection Report Hunter Corporation Inspector Mark. M. Tabbert 8/16/83. This is a-list of.118 ASME, LISC deficient welds, mainly on a feedwater system. This document is attached to my testimony as Attachment 10. I have compared the list of 118 welds with the listing in t'ne ASME categcry that was reviewed by Sargent & Lundy and ! it ' appears from the documentation available to tLe that none of the 118 were Jn the group reviewed by Sargent &'Lundy. It thus appears that these 118 welds may have been fixed before any review was done. The list of the 118 welds describes the. deficiencies, and in my opinion, if'there had been an engineering evaluation performed on these weld deficiencies some would likely have been found to be 17 I'
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safety-significant. _ Q26:~ What is the next area of your concern?
-A26: Calculation book 19.1.2, the Design Procedures and Assump-tions for' Evaluation of As Built Welds with AWS Inspection Discrepancies,.p. 14 No. 7 is attached to my testimony as Attachment 11. The document states that " Convexity is only considered a defect on welds'with fatigue load application and i does not effect welds at Byron /Braidwood s ta t ions."
This quote is consistent with the testimony of John McLaughlin, who testified that pipe supports were not sub-ject to fatigue analysis. In my opinion, pipe supports are subject to fatigue and thus convexity should have been considered a defect. Donald L. Leone, in his testimony at page A.20 (fourth
. line) (later adopted by Mr. Branch) stated that, " weld discrepancies involving ASME Class I piping were evaluated against the fatigue analysis for the piping system."
, This may evidence a lack of communication between the structural group and the mechanical group. If mechanical designs account for fatigue in the piping system, then the structural group should take that into account when design-ing those respective pipe supports. Pipe supports are subject to load reversal many times. In the NRC document Review of U.S. Nuclear Regulatory Commission 1983 Annual Report (Attachment 12), on page 17, an article entitled l Water Hammers speaks of " water hammer", a condition that causes fatigue loading and states, "The frequency of occur-l 18
7 l..' rence .is low and: damage has. generally been limited to
. piping -supports." .As the NRC article points ~out, pipe supports are'thus subject to fatigue loading.
I disagree with - Sargent - & Lundy's; apparent position. that fatigue load-ing does not', affect AWS welds at. the Byron plant.
'Q27: Do you'have.other. concerns affecting the Reinspection Program?
m A27: Yes. Continuing in Sargent &-Lundy Calculation-book
-19.1.2; Design Proce'dures and Assumptions, page 20, No. 5.
(attached as Attachment 13)-it states, list D1.1-83 as the structural welding code. In my review of the FSAR' .the
-1983 code was not listed. This would relate to all as built welds with AWS code discrepancies, whether Hatfield or Hunter.
Q28: Why is-this a problem?
'A28: There are potential problems when one installs or builds under one code and then attempts to pass the weld under a later code. For example, under the 1983 code it may be easier to pass a weld by calculation because the installa-tien criteria in the new code are more demanding. In contrast,- welds may be more difficult to justify by calcu-lation under the earlier code because the installation criteria were then'less demanding. Therefore, if one builds under an' earlier code (with less demanding installa-
. tion criteria) and attempts to pass the weld under a later code (requiring less demanding calculations) one may end up 19 i
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passing.a weld that may_have failed under either code. As a standard practice if one installs a weld under one code, one should qualify that weld under the same code. Since the-1983 code is not referenced in the Byron FSAR, it-appears that the weld was not installed per the 1983 code and therefore should not be qualified under that code.- Q29: Do you have additional concerns about the Sargent & Lundy calculations relating to the Byron Reinspection Program? A29: Yes. I have questions ou-calculations and assumptions
'found in Calculation book 19.1.2, Sections 2.1 (p. 5) and 4.1 (pp. 7-11), Section 19 (pp. 1-5), Section 21 (pp. 77, 78, 78A), Section 21 (p. 97 A), Section 21 (p. 109) and Section 21 (p. 113).
Q30: What is your concern with Section 2.1 and Section 4.17 A30: This relates to a PTL inspected weld in the Reinspection Program. On weld No. 140, Beara No. 33601-L on page 8, the weld appears to be shown to be overstressed 1.18 to ' 1.1. This is true even though Sargent & Lundy used a 10% over-stress factor for as built conditions. On page 11 the engineer writes, "Ry is assumed to be taken by check pla t es." On the same page actual stress divided by allow-able is shown to equal .996, less than 1.10 It appears - that this joint would fail if it were recalculated including Ry. (Reaction R in the vertical (y) direction.) . l 20 4
g A '~~ l s a. l Q31: What is your concern with Section 21 pages 77, 78 and 78A7 A31:: 1This, too, relsces to a PTL reinspected weld. These pages concern a combination bolted and welded connection. The engineer who performed this calculation makes_ an assumption thats the bolts take part of the load and that there is a minimum pretension in the bolts. My concern is that this assumption is, to my knowledge, not consistent with indus-try practice.. When one designs a welded and bolted struc-tural plate the weld is designed for 100% of the shear load. In the case of pull out or tension load the anchor bolts are designed for those loads. One cannot reasonably assume that a certain amount of load goes to the bolted connection. In this case, looking at the drawing of the weld and bolts, the bolts would not catch very much load until the weld fails. Even if there is bolt pretension initally, the bolts will relax over time. For the engineer to make the assumption that the bolts take the load, he or she would have to show that those bolts could take the load for the life of the plant. This assumption is not conser-vative, nor is it proven in the calculation. However, it
. appears that without this assumption, the weld would not have passed the code.
F Q32: What is your concern with Section 21, page 97A? A32: The calculation in this section, relating to a PTL rein-spec td weld , show s "I = 2.13 grea ter than 1.0 n.q." If this means, as it would appear, that the interaction value is equal to 2.13, which is greater than the code allowable 21
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i i _ _ - _l
o f.11.0 0, then this equipment does not qualify under the Leode. 'This-is among.the worst cases of apparent failure I ; have come across in my review. The document goes on to , say, "Therefore, an Rz must be added per Phase III modifi-cations." It is not. apparent what is a Phase III modifica-tion.' If this did fall under the Reinspection Program, why was it not recorded as a Reinspection Program failure? A-failure of this apparent dimension would, in my opinion, likely be safety-significant.
-Q33: What'is your concern with Section 21, page 109?
A33: This, again, concerns a PTL reinspected weld. The respon-sible engineer-states, "This portion of the load will be taken from load "D" and it will be distributed to weld "A" and "B". In my opinion this should not merely be assumed to happen; it should be proved. There are reliable ways _to distribute loads based on fixity or on the: strength of the section. The fact that the engineer makes_this assumption does not-necessarily mean that the load would in fact react in that manner. I do not see any apparent foundation for this assumption. A calculation should have been performed to justify this questionable assumption. The only apparent way this assumption could be valid is if one assumes "D" failed. That would ensure'"A" and "B" would assume'the load, but that in itself f ails "D". Q34: Do you have concerns relating to Systems Control Corpora-tion (" SCC")? 22 -n __ e__ _._- - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
y , 9 lA34i Yes.- I.have'a series:of questions regarding SCC' supplied control boards. Commonwealth' Edison's NCR 695 Attachment A
-(my Attachment - 14)l shows that three main control board .
L sections . (1PM02J , 1PM02J, AND 1PM05J) has.been repaired with "Bondo," an . auto- body type repair compound, and by-tack welds, rather than by the full peneteation' weld speci-fled in the design. An April 28, 1982 SCC letter to Sar-gent & Lundy-(Attachment 15) revealed that SCC had used such auto body filler in many panel face repair applica-tions. In the third-paragraph,. SCC states, "We can only conclude that.the area of the board containing the cracks may have been subjected to abnormal thermal or-structural stresses."' An April 30, 1982 Sargent & Lundy interoffice memoran-
~
dum from J.A. Schwin to B.G. Freece in response (NCR number F-695; my Attachment 16) has a note at the bottom which states, "The use of body filler material' (Bondo, etc.) is a standard- practice of control board manufacturers in repair-ing blemishes to their boards." The documents I have reviewed do not disclose the final resolution of this problem, but the documents do raise serious questions. The drawing called for full penetration welds, not tack welds and bondo. The question then arises
-- What is the-function of bondo on these control boards?
Is it for strength? Or is it a: sealant? These main con-trol-boards are Class I safety equipment. Is it possible that-the bondo could again' crack and a particle become 23
~
r I
i' lodged in the contact' switches? I'would also. question
~
whetheryproceduresfare in place for-the design,-installa-tion and qualification of'bondo in Class-I safety related controls. In sum, I-would state that-if the use of bondo is as a' sealant, high temperature silicon would be better. If the use is.for strength,' welding would In3 better, even taking into account the potential for warping the panel. l Q35. Do you have other concerns with. SCC control board panels? A35 Yes. There is another issue rela' d to SCC control boards which arises from my' review of Edison's NCR F-544 (Attach-ment 17). NCR F-544 indicates that. main controlcboard panels OPM01J, OPM02J, 1PM01J, 2PM01J, 1PM04J, 1PM11J, and 2PM11J did not meet the AWS D1.1 code criteria for welds. As I understand AWS D 1.1, it is not a stringent require-ment, but is among the easier AWS code requirements to meet. In an effort to correct'the situation NCR F-544 indicates that SCC was allowed to write its own acceptance criteria. This is an issue which would appear to deserve investigation.
.Q36: Were there other Sargent & Lundy calculations for the Reinspection Program that caused concern?
A36:- A. review of Sargent & Lundy Calculation book no. BRP-1 for
~
Hunter Subjective Welding also raised concern on my part.
-This calculation book reviewed 60 AWS type discrepancies and 49 ASME; weld discrepancies,. of which only two were on 24 l C
)
f:_ :.;
. the feedwater ? system and five on main steam.
(This con-
~
trasts 1with the 118 ASME welds referred to earlier. in .my testimony which may haveLbeen fixed without a Sargent & Lun'dy -review.) i Specifically, ASME : welds Nos. , 62 (S-CC-100-11 A) and 63 1 l (S-CC- 100-33) '( Attachment' 17) were accepted despite the-I fact that the accuracy of the gauges supplied'for measuring the welds was only 1/64 of an-inch, whereas ASME requires machine shop type accuracy to the thousandths. The infor-u mation I have reviewed suggests that this is an impermis-2 sible practice. Q37: Do you have additional concerns over documents you reviewed. at Sargent & Lundy? 1 I h' ave not had time to discuss them all in this i- A37: Yes. testimony, but I do have time for one more. I reviewed a i drawing in Sargent & Lundy's home office, Review of Cate-gory I Conduit Supports Typical Support Types and Load Tables
' DWG. 6E-3393B. This document depicted support type CF and I MCP (floor to ceiling conduit support) and type CC and CP i
f~ maximum load table. When I reviewed the tables I was immediately concerned. In the plant visit I observed can-tilevers coming off the ceiling for 18 feet:using unistrut. I found this alarming. 'The CF and MCF uses approximately 2
- inch to 4 inch tubing. I saw conduit supports in the field
- that did not meet' design requirements. .It'was floor'to !
f , ceiling and was weldedcat the top and at the bottom. i According to the drawing it should have had a slip or pin , I
- 1 1
c, .g. n connection ~'at.the. top and no weld.. Beyond that, it appears
- that:the KL/R, which is the slenderness-ratio, which is a factor that comes' out.of the UBC, AISC and a number of other documents' including the unistrut catalog, which the-actual supports came.out of,. states that the limit for KL/R is-200'. feet. I-reviewed many designs that exceeded the 200
-foot factor. .One I noted was 300 feet.
It appears there is a problem with the K factor used in the equation. The unistrut . catalog uses .8 'as a built in factor within-that document. A pin. connector uses a factor-of 1.2, and a cantilever uses n~ factor of 2. When one uses any of those other factors instead of the .8 for unistrut, these supports are substantially outside of the length-requirement. The unistrut book says that when one exceeds 200 for KL/R one no longer has a yield stress failure,-but a buckling failure. I have viewed this problem both in the tables at Sargent & Lundy and in the field at Byron. l 26
r.__..- m,% , .- A11=JA .1 TED CORR 2 S?OECW RESUME CHARLES C. S1MES P. E. SSt 416-72-9963 Yp[ ROUTE 1 BOX 223 COTIUM000,AL. 36320
, TEL. - (205) 677-5078 '84 AGO 16 Pl2:37 PER904AL Imm: , ,y y7--
Dat3 of Birth - 03/12/51 - SINGLE - U.S. CITIZEN - EXCELLENTf[d4L W O M p g W .E i PROFESSIG4AL EXPERIENCE: FIELD ENGINEER - (NOV. 8,1982 % OCT.14,1983) Acspted assignment to Pacific Gas and Electric Companies DIABLO CANYON NUCLEAR PROJECT UNITS 1 & 2. Placed in on-site engineering group. Performed pipe stress and pipe support design calculations. Wrote raper on how to design and represent flare, flare-bevel, Ekewed welds and other partial and full penetration welds on drawings to comply with AISC cnd AWS prequalified welds for structural and tube steel. Was assigned to Pipe Support Design Tolerance Clarification Group to authorize changes required for installation of supports and was responsible for snubber substitution on both units. PIPE STRESS / SUPPORT ENGINEER - (2/82 'IO 5/82) Field consultant on Mississippi Power & Light's GRAND GULF 1 for RCI Inc. Assigned to Control Rod Drive System to assist ECHO pipe stress group and RCI hanger group in resolving interference problems by suggesting alternate design. Responsible for ECN's of as-builts and alternate designs and supervising drafting. Assisted QC and Construction personal in interpretation of drawings. BWR Plant and class 1 pipe. MECHANICAL ENGINEER - (6/81 TO 2/82) Assigned to the Mechanical Engineering Dept. of the Lawerence Livermore National Labor-ctory as a stress analyst on the injector of the Advanced Test Accelerator (ATA). Performed calculations on the injector housing, epoxy insulators, accelerator cells, cat-hods, anode, support structure and handling fixtures for fabrication and installation. System involved vacuum-oil interfaces and extremely strong magnetic and radiation fields. Injector constructed of aluminum and stainless steel with in<.,ulators of a special fill-cpoxy compound. Also made design changes to epoxy insulators on Experimental Test Accelerator (ETA). PIPE STRESS / SUPPORT ENGINEER - (10/8010 5/81) Contracted to Nuclear Services Corporation, a division of Quadrex Corp. in San Jose Cclifornia. Performed pipe stress calculations and design of safety related small bore piping supports. SIGS program was used in analysis of complex supports. Was assigned to ZIMMER NUCLEAR PLANT as a member of special pipe stress and hanger analysis group. Class I, II, III pipe. PIPE SUPPORT ENGINEER - (6/801010/80) Assigned to Bechtel Power Corporation's Civil Structural group in Gaithersburg, Md. , working on the DAVIS-BESSE PROJECT. Checked and made base plate and anchor bolt stress j calculations and modifications for anchors and pipe hangers. ANSYS finite element pro-gram utilized to account for plate flexibility and bolt elongation. Strudl was used for 'l cnalysis of complex frames. Other in house programs were also used. *; Y j PROJECT / DESIGN ENGINEER - (7/75 TO 5/80) Southern Compar.y Services Inc., Birmingham, Alabama. Wr:t3 two specifications concerning modifications to Georgia Power's HATCH NUCLEAR PLANT. 1
continued: The main item modified was the Reactor Heat Discharge System in the Torus. Designed the structural steel truss for Georgia Power's SQlEREER PLANT coal conveyor system Unit No. 2, including details and bents. Redesigned the precipitator structural steel on Alabama Power's MILLER STEAM PIANT to add precipitator roof enclosure. Elastic analysis performed to allow for thermal growth and to resist wind forces. STRUDL analysis, code check and design was used. Acted as a nuclear. pipe support stress analysis, designer and checker on Alabama Power's FARLEY NUCLEAR PIANT. Performed stiffness calculations and checks by hand and computer. STRUDL was used for analysis of complex structures. Also worked in the field supplying support information to office personal. Work performed in accordance with NRC 79-02 and 79-14. PWR class I, II, III pipe. Sarved as civil material coordinator on Georgia Power's VCX3TLE NUCLEAR PLANT. Was responsible for civil quantity take-offs for project construction scheduling, financing and material purchases. Computer storage and retrieval of information was used. Did ANSYS finite element analysis of powerhouse substructure on Alabama Power's HARRIS DAM. Supervised draf ting, checked drawings and checked calculations on superstructure Concrete. Designed outdoor structures on Alabama Power's MILLER STEAM PIANT. These included railroad, truck and ash pipe bridges, ash trench system and off-site make-up water system. Responsible for checking calculations, supervising drafting and coordinating field and inter-office disciplines. PROFESSINAL LICENSES NO AFFILIATIWS: Registered Professional Engineer - State of Alabama - (12786) Registered Professional Engineer - State of Florida - (29985) Registered Professional Engineer - State of Georgia - (12340) EDUCATIGi: Birmingham School of Law, Birmingham, Al., Juris Doctorate degree, May 1990. Auburn University, Auburn, Al., BCE degree, May 1975. Massey Institute of Technology, Jacksonville Fl., correspondence accounting.
'IEE FACTS STATED ABOVE ARE TRUE AND ACCURATE Charles C. Stokes P.E.
i l l
AHerhment 3
+
RELATED'CORgESPOMDEEE S & L DOCUMENTS REVIEWED-
-STRUCTURAL PROJECT DESIGN CRITERIA BYRON CM TERAIDWOOD NUCLEAR POWER STATION UNITS 1 & 2 (DC-ST-03-BY/BR)$REV. 12 .
Sect.1.1, para. 2 No exceptions to the Final Saf Report and Enviromental Report is per&dtdG036 R2:3)ty Analysis. Sect.7.4.1.b ," cong. on metal deck-category I Interior floors 1B", walls roof l{6hhhtkpJ.;tglabs 14", 12", 4" and rojm ceiling category II slabs 6" -(? anchor bolt problem)2! Sect.7.4.2.b exterior walls below grade 15" min. thick and above grade 24" min, thick.(? reversed) Sect.8.1.a ACI 318-71 (? FSAR REQUIREMENTS) (? USED IN j DESIGN) Sect.8.1.b ACI 322-72 (? FSAR REQUIREMENTS) (? USED IN DESIGN) , Sect.8.1.c AISC-69(ELASTIC DESIGN) (? PLASTIC DESIGN) (? USED IN DESIGN) Sect.8.1.d UBC-73(SEISMIC ANALYSIS CATEGORY II STRUC-TURES) (? FSAR REQUIREMENTS) (? USED IN DESIGN) Sect.8.1.e AISI-68 (DESIGN COLD FORMED STEEL STRUCTURAL
- MEMBERS) (? FSAR REQUIREMENTS) (? USED IN DESIGN)
{ l Sect.8.1.f 73 ASME Sect. III Div. 2, proposed staridard f ' code for concrete reactor vessels and containments. (? FSAR RE-QUIREMENT3) (? USED IN DESIGN) Sect.9.5 CATEGORY II DEFLECTION WAIVED (? WITHOUT SOME LIMIT IMPOSIBLE TO DETERMINE WHSN CAT. II EFr'ECTS CAT. I) TABLE 9.3-1 NOTE 5 1,67 AISC <a .55 Fy (? FSAR REQUIREMENTSJ TABLE 9.4-l' DESIGN STRESSES 1.75 AISC ? Fy (? FSAR RE-l QUIREMENTS) Sect.10.2.1.1.3.4 In all cases, structural members will be checked for the loads obtained from the pipe and cable pan hanger drawings. (? table diff.) Sect.10.2.2.1.1 33 hz or less or increase acceleration 50 % ! Sect.10.2.2.2.1 LEEWARD PRESSURE IS SUCTION NOT APPARENT IN TABLE Sect.10.2.2.2.2 LEEWARD PRESSURE IS SUCTION NOT APPARENT IN TABLE l Sect.10.2.2.2.3 LEEWARD PRESSURE IS SUCTION NOT APPARENT IN
. TABLE
. -_ .-~ ._ _ . . _ .._ .
m 19 # Sect.10.2.3.3.1 FOLLOWING PARA. EXTREME ENVIROMENTAL (1.67 AISC allow. ? .95Fy) (? FSAR. REQUIREMENTS) Sect.10.2.3.4 a. 1.6 AISC allow.. .95Fy-(? FSAR REQUIRE-MENTS) (? USED IN DESIGN) see other sections. Table 10.3-1 DESIGN STRESSES COLUME 1.6 AISC. allow. .95 Fy (<= left out) Sect.12.2.4- FORMULA Pae=l/2 ? HXH Kae~ (missing lambda symbol) Sect.18.1.1 ALL DESIGN ASSUMPTIONS, METHODS, REFERENCES AND MATERIALS SHALL BE DEFINED FOR EACH AREA OF DESIGN USING STANDARD CALCULATIONAL
SUMMARY
SHEETS.
' Sect.19.5.d EQUATION MISSING SUMMATION SYMBOL BEFORE THE b SQUARED Sect.19.5.d EQUATION SHOULD BE SQUARE ROOT OF F'c Sect.20.3.1.d MAX. WT. OF CONDUIT & CABLE DIFFERS FROM NEC 71 VALUES IN UNISTRUT-CAT.
FIGURE 21.8-3 ?NF TO WELD FIGURE 21.8-4 ?NF TO WELD FIGURE 21.8-5 ?NF TO WELD Sect.32.3.1 ? EQUATION NOT ABLE TO VFRIFY (REF. STEEL PLATE ENG. DATA-VOL.3 WELDED STEEL PIPE AISI) Sect.32.3.2 WALL THICKNESS SHOULD BE CHECKED FOR INTERNAL PRESSURE AND EXTERNAL LOAD BEFORE INTERNAL PRESSURE IS APPLIED, NOR HAS A MINIMUM THICKNESS BEEN CHECKED FOR SAFE HANDLING Sect.32.3.2 :25 fy SHOULD BE .25 fy Sect.32.4.2 SPANGLER'S EQUATION D.061 SHOULD BE 0.061 AND R TO THE FORTH SHOULD BE R TO THE THIRD IN' DENOMINATOR Sect.34.2 EMBED PLATES DESIGNED FOR 10 KIPS PER FOOT TENSION LOAD AND 12 KIPS PER FOOT SHEAR LOAD (? PLATE SAFETY i FACTOR WITH CRITERIA THAT ALMOST EVERY THING IS HUNG FROM THEM) i Sect.35.3.1 STRESS LIMITED TO 1.0Fy FOR LOADING AND Fy/sq. root.of 3 FOR SHEAR (? .95Fy for tension loading) Sect.36. ************* Sect.37.1.2 (? NO LIMIT OF DEFLECTION ON NON-SAFETY HANGERS IN SAFETY RELATED AREAS) WHAT CLEARANCE CRITERIA WILL BE USED TO ENSURE THAT NON-SAFETY DOESN'T DAMAGE SAFETY? 3 -
n: c c A%.h'med 6 RELATED CORRESPONDENLL
~
Sect.37.2 NO DEFINITIVE: STATEMENT THAT-TORSIONAL' STRESSES SHOULDJBE CHECKED 4ect.37.2.1.'f DEFLECTION AND ROTARRQ#7pFTPRIMARY STRUCTURAL . STEEL IGNORED IN DEFLECTION CHECK (? MEMBERS WITH PINNED ENDS) Sect.37.2.1 9 1.B. IGNORE AXIAL SE ' WRGJrf [? MAGNITUDE OF LOAD AFFECTING MEMBERS AND CONNECTIONS) WU 16 M2.36 Sect.37.2.1.g.1.C. TORSION.ANALYSLSFNOTE REQUJRED (? MAGNITUDE'OP LOAD AFFECTING MEMBERS AND CONNECTfdSS')T!'1G & SEh>0 BRANCH Sect.37.2.1.g.2.8 AXIAL SELF WEIGHT MAY BE IGNORED (? MAGNITUDE OF LOAD AFFECTING MEMBERS AND CONNECTIONS) Sect.37.2.1.g.2.C. TORSION INCLUDED flERE ? LOGIC Sect.37.2.1 9.~3.A.- ASSUME ALL MASSES LUMPED AT THE SHEAR CENTER S ec t . 3 7. 2.1.' g . 3 . B . AXIAL SELF WEIGl!T MAY BE IGNORED Sect.37.2.1.g.3.C. TORSIONAL ANALYSIS IS NOT REQUIRED Sect.37.2.1.g.4.A. ASSUME ALL MASSES LUMPED AT SHEAR CENTER Sect.37.2.1 9 4.B. AXIAL SELF WEIGHT MAY BE IGNORED Sect.37.2.1.g.4.C. TORSIONAL ANALYSIS NOT REQUIRED Sect.37.2.1 9.5. EXACT ANALYSIS MUST BE PERFORMED FOR LOADS GREATER TilAN 20 KIPS Sect.37.2.1.g.S.A. ASSUME ALL MASSES LUMPED AT SilEAR CENTER Sect.37.2.1 9 5.B. AXIAL SELF WEIGHT MAY BE IGNORED Sect.37.2.1.g.S.C. TORSIONAL ANALYSIg NOT REQUIRED Sect.37.2.1.g.6.A. ASSilME ALL MASES LUMPED AT SHEAR CENTER Sect.37.2.1.g.6.b. AXIAL SELF WEIGHT MAY BE IGNORED Sect.37.2.1 9 6.C. TORSIONAL ANALYSIS NOT REQUIRED Sect.37.2.1.g.7 A. ASSUME ALL MASSES LUMPED AT SHEAR CENTER 1
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