ML20198K527

Rev 1 to Instruction CP-EI-4.1-1, Field Support Design Group Design Control Instruction. Related Info Encl

ML20198K527
Person / Time
Site:	Comanche Peak
Issue date:	07/13/1979
From:	TEXAS UTILITIES SERVICES, INC.
To:
Shared Package
ML20197J316	List: ML20197J311 ML20198K377 ML20198K383 ML20198K450 ML20198K456 ML20198K467 ML20198K472 ML20198K479 ML20198K489 ML20198K502 ML20198K506 ML20198K511 ML20198K513 ML20198K518 ML20198K527 ML20198K537 ML20198K556 ML20198K562 ML20198K568 ML20198K574 ML20198K581 ML20198K588 ML20198K600 ML20198K844 ML20198K853 ML20198K872 ML20198L117 ML20198L169 ML20198L186 ML20198L203 ML20198L212 ML20198L218 ML20198L235 ML20198L257 ML20198L261 ML20198L275 ML20198L291 ML20198L303 ML20198L317 ML20198L325 ML20198L337 ML20198L367 ML20198L379 ML20198L384 ML20198L388 ML20198L395 ML20198L401 ML20198L406 ML20198L420 ML20198L431 ... further results
References
FOIA-85-59 CP-EI-4.1-1, NUDOCS 8606040047
Download: ML20198K527 (36)

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TEXAS UTILITIES' SERVICES INC.

INSTRUCTION REVISION

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1DATEh7 FIELD SUPPORT DESIGN GROUP PREPARED BYun DESIGN CONTROL INSTRUCTION APPROVED BY DATE[ N

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.T.0 REFERENCES

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- *. ::.M: CP-EP-4.1 Design Planning j y

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'7.i? CP-EP-4 3 Design Input.

M;CP-EP-4.4 Control of Design Documents -

' (:[' 9 Y.1 j...E':-l:.-X#%$%?@ CP: EP-4.5M. Design. Veri fication Mt.

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EO GENERAL

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-.'. 4..d. net.The purpose of this instruction is to establish a standard method,-format and sequence for the generation and/or revi-

.h J7.7.M,,.7 gp$g.;e,hy' Tsion ofLdesign documents anithe method of control of these s j.

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c q N1... y:7.r.,cir.d$iw. documents by the' Field. Support Design Group.

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x, der, pipe supports will, be the only items affected).

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RESPONSIBILITY v.

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'- Supervisor to assure that documents are generated, revised Y

and controlled in accordanca with this instruction.

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~ 2.4 DESIGN INPUT y N,f.',

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Y 2.4.1 For design documents included in the scope of this procedure, d

the original design organization shall culations, etcetera)put documentation (provide th d.

sufficient design in criteria, data, cal-.' N E

to allow a consistent basis for making

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design decisions, accomplishing design verification and evalu-4g[,,

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ating design changes.

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EOR !NFORMAT10H dsti it.M:.A I

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TEXAS UTII.ITIES SERVICES INC. INSTRUCTION REVISION PAGE

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7-13-79 2 ofl0 s

2.4.2 When the Field Support Design Group revises or develops new criteria necessary to accomplish design activities, i

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' 1 the engineering organization developing the original cri-

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teria will be consulted prior to revision. Changes in cri-teria and new criteria shall be documented in writing to c.-)

the Comanche Peak Engineering and Construction Manager.

l Copies will be forwarded to any other affected engineering

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organization.

f 2.4.3 All design inputs are controlled in accordance with CPP-

-N,j EP-13.

.p 2.5 PERSONNEL QUALIFICATIONS M,

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The Design Engineering Group Supervisor shall maintain a Q

list of engineering personnel qualified and authorized to check, approve and design review documents.

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, Designation of an individual to perform any of the above W

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engineering functions shall be based on his experience.

Q performance and overall job knowledge.

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-;, INSTRUCTIONS

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- Design documents which are to be generated 'or revised on

- ~ 3.TJ * ' jsite shall be maintained by the Field Support Design Group

~ J irr locked cabinets. Record copies of these documents will jj

~~' be. maintained on site.

,6 7.Z' Requests for generation or revision of design documents O

may be initiated by my design or construction personnel y

involved in the Comanche Peak Project.

, wj 3.3 All requests for generation or revision of a design docu-cf ment shall be processed by the Design Engineering Group f{

and shall be in a suitable format, acceptable to that group, to provide sufficient details to evaluate the request.

913 3.4 The Design Engineering Group shall review all requests and Jy perform any design calculation required to assure its q !

acceptability. The group is responsible to assure that all F.i design criteria and code and regulatory requirements per-tj taining to the design document are adhered to.

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3.5 If the request is found acceptable, the appropriate design d.

.td document will be generated or revised. All changes from ff previous revisions shall be clearly indicated. The drawing T.i shall be initia11ed by the draftsman or preparer.

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7-13-79 3 ofl0

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! O 3.6 The engineer perfoming the review and/or design calcula-d tions shall check the document for accuracy and shall i

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initial the " Checked" block indicating his approval.

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'i 3.7 Following the engineering check in 3.6 the document shall be approved by an individual authorized by the Design Engine-1':

ering Supervisor. Such approval shall be indicated by 2.]

initialing the approval block.

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performing construction, inspection or other.' elated inter-2

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.'l_J.$2 face activities. Each controlled drawing issutd shall be 4l...-)l.

^ p,/..J..c...u.-:sGQnumber of the document holder shall be entered on the cr

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g. stamped with the stampface shown in Figure 6 and the control

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WQT'.' p.6%' ment will be maintained on the site.

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hatched area of the stampface. A record copy of each docu-q l

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All distribution shall g

v be controlled by the Design Engineering Group. A log (Figure e.

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f.e - d,.V.sM/> distribution, number of copies, issue data of each revision g/g and. document holder mceipt acknowledgement date.

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Ef$l signed by the file custodian for the document holder.

rev* 1' 7 ~ ISM?E5'hlt.sha11 be the responsibility of' the individual signing H

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-. G'p Dhh--Y the form to: disp'ose of any superseded issues.

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- 3.100 f.In: addition to the controlled distribution above, any O

.:, / m affected Design Organization will be forwarded an informa-

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,- 3 tion copy of each document stamped "For Engineering and M

. a Office Use Only". Additional infomation copies may,

, upon request, be provided to other parties.

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3.11

. DESIGN REVIEW

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Each design document generated or revised under the juris-diction of this: procedure shall require design review.

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A Design Review Log (Figure 3) shall be maintained by the Desiys Engineering Group. It shall indicate the document yl",

and re' vision nunter, responsibility for design review, date design review complete and design reviewer. Design re-view perfomed on site shall be in accordance with CpP-EP-12.

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3.11.3 When the responsibility for design review is assigned to an off-site party, the Design Engineering Group shall infom the cd organization performing the review in writing of the items assigned to them. The reviewing organization shall femally Q}

acknowledge completion of the design review to the Design d

Engineering Group to close out each item in the Design Review Log.

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INSTRUCTION REVISION PAGE DATE e

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CP-EI-4.1-1 1

7-13-79 4 of 10 i

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. 3.12 DESIGN DEVIATIONS l

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3.12.1 When occasions arise to deviate from design documents controlled within the scope of this instruction, a DCA or CMC may be used to document such deviations.

3.17.2 Deviations will not be incorporated in the design rev. 1

document,

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DCA's-and CMC's shall be prepared, approved and dis-

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3'.12.5 The Destgrt Engineering Group will maintain a log fj.

(Figura 4) for each drawing indicating the deviations affecting the drawing.

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OCA or CMC by initia11ng-and dating the Deviation

.QTf.7.1-1._. Distribution Sheet (Figure 5)..

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- ~. -. ~ fy. I .s_ 1.2 ,si. N I. During what time period.did you work-in the drafting department? i.: Please be specific and give dates from _ t _ f l f. hanger sketches BRH, or support / piping isometric 2. Did you wo drawings, BRHL? I I Can you identify any specific piping systems, pipe lin p,. the hangerstocations? Were the BRHL drawings used to fabricate and install the hangers? fMA / M *f 4 0 $ V 0' W S V y M M Lsf g 71 is l$,f r uses. 1,

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~ ~~ f CASE ATTACMMENT 11 - Page 1 ~[C recardine his feelines followinc meetino with NRC; 12/13/82, anc 12/20/82: Cormients of relayed to CASE President Juanita Ellis by enone I was especially concerned about the Q (safety-related) cable that was in the cable sprdad room being spliced, and one of the main reasons we went there was -E. to see what the Nuclear Regulatory Comiss. ion was g We wanted to know what the NRC i-- Q cable. dt the Nuclear Regulatory Ca=ission hearing. was going to do about the spliced Q cable. I have other infomation that I believe shows that pipe succorts and pipe hancers are not located in the proper areas.' that tnene are f aulty welds onThey tried ~ the supports. _They kept asking us questions about it.They tried to sa I As far as I'm concerned, we were badgered in the conversations that we my conversations. We kept trying to find out what they were going to do about that spliced true. The meeting went on for about an hour and forty-five minutes and had. Q cable. Juanita Ellis and I decided to go ahead and leave. I don't think that they tried to take I felt they were trying to intimidate us. I think what they really wanted They tried to feel me out. me sericusly at all. to do was te find out what other information I had, if I had a grudge against I them (the Nuclear Regulatory Cora.ission or Brown & Root), which I did not. came out of Brown & Root with a very high classification, they were very muc I left of cy own accord. to school and also, the other main reason that I left was that I couldn't condon willing to reh1re me. I thought that they were not treating the what was going on at the plant. I knew that these pi oblems or faults were being problem areas as they should. conducted or were going on, and I tried to stop it, and they just told m keep my mouth shut about it. ahead and quit. Bob Stewart telephoned me after the meeting, I believe on Friday (11/12/82 He asked me why I didn't come cown and give them the papers, and I told him that at this time I have no inter.tions of giving him the papers, for the reason The that I felt that he wasn't goirq :o do anything about the spliced Q cable. He made conversation went something lib this, to the best of my recollection. several com1ents like, "Well, if we don't have that infomation, how do you ex / And I said, "Well, I gave you a location, I expect you to locate i He said, "Hoa can we go through a couple of thousand cables in that us to do it?" l this area." You have a job to do it. You're And I said, "Well, it's a spliced cable.He said, "Well, how can we do it wit area?" being paid to do that job."I said, "I gave you the location of where it was, you c I said, "No, ings?"He said, "Well, are you going to bring those drawings to us?"And I asked h it." I'm going to turn them over to Juanita Ellis." His answer was no, ll else was listening in on the phone conversation. I told hiin that I had talked to an engineering professor at UTA and let him listen t'o the tape and look at my drawings and he suggested not to t -- not to let any of the papers out of my hands -- in fact, he suggested not to And to just go ahead and hold them but even turn them over to Juanita Ellis. scatter them to different areas where I can locate them sometime if anyth t does happen like a fire or accident at my home, or if a thief accidentally do And Mr. Stewart also tried to figure take one box of information that I have. i out what papers I did have and.J. told him I wasn't going to talk to him ab d3 Y

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CASE ATTACHMENT 11 - Page 2 And he said, "Well, you made statements and you're going to have it any more. And I said, "I have all the infonnation. You know, my state-to back them up." You know, you're badgering me now, you badgered me during 'i 4 that meeting, you're trying to discredit me now in this telephone conversation ment's backed up. i that we're just having, and you tried to discredit me I said, "I'll do what 4 feel that way, you go ahead and do what you have to do."And that wa The total I have_ to do." conversation lasted about ten minutes or less. When people that come in and have something to say, some infonnation to give e; the Nuclear Regulatory Comission, I think that the people are discredited. I think that they're not taken seriously, that the Nuclear Regulatory Comission "Well, that's no big deal, and you knw. that's just a routine -- could ~ And the people think "Well, it's possible; these guys

says, be a routine thing."

ought to know, they're a government agency, they're paid by the government, The way I see it now, they're independent, and they have no reason to lie." At least that I didn't !~ I believe that the NRC's not telling us the full story.And I think if they talked ,j get the full story. everybody else that way. j I don't know what I would have done if Mrs. Ellis hadn't been th a hadn't had that tape recorder, because I'm sure that they would have made ne i I literally cussed, I said some foul language, and I d very mad. The guy would don't get that mad. They were very, very tactful in doing it.There is one guy, Mr. Stewart out. make a very good attorney in cross-examination. i ld, who sat across the table and kept on popping questions at me as fas he popped questions to me. more logical questions. He felt me out. While Stewart badgered these questions one after another Tom 1'inson as specific questions, and a little bit more in detail, and that's where Stewart 1 also just kept on popping questions, you know, to relate to these cuestions. And then, Richard Herr asked very classic questions, very detailed, direct,H l'- and he did more cut down than anything. And if I were to cross-examine somebody, that's the way I would cv. wet it. Very I thought it was very good the way they did it, but very poor I consider myself up, you low class. (; that I do not consider myself a low-class person. I think the NRC representa-knm, as high as the NRC people are if not higher.tried to make m rj .l. tive treated me as such, and I think I'm pretty well educated. in college, I'm working on my second degree, and I didn't appreciate his attit , j ', 1 I would have If Mrs. Ellis hadn't been there with me, I would have walked out.I would have said, l told them, "To hell with it" and walked out. no more donversation, nothing else to say " And I would have probably gotten up - ll and walked out because they were treating me very poor, they were treating m 't I didn't like the way that they conducted themselves.

I very bad.

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.. =. : : _ =-- + l r. CASE ATTACHMENT 11 - Page 3 i I don't If I had it to do over again, I don't think that I would have gone.If scmebody else had a i think that anybody else would go if they find out. situation that they knew that there was a fault in the plant that might cause i 4 a problem'at some time and they knew that the Nuclear Re l I I wouldn't go back there to1: hem. I don't think I'd turn my inforg .} I think I'd turn it over to Mrs. Ellis. I gave some of w go to them. mation over to them. the information to the Armadillo Coalition and I thought that they might be able to do something about it. Regulatory Comission down at Brown & Root had a notorious reputation as supp being the arm of -- or the negotiator in between the people and the -- doing what The people that worked there felt that the Nuclear i right and doing what's wrong. I believe that there are people Regulatory Comission was not doing their job.at the plant now that i Regulatory Commission didn't have the reputation that it does now. {. I think that it jeopardizes the people that work at the nuclear power plant when Brown & Root puts out their newsletter and makes statements about people that are talking against them like the gentleman that has been fired a nu=ber of times. Also my name was written up and many other people that did make coments or did talk to the Nuclear Regulatory Comission were written up in the newsletter. This newsletter is put out onsite to all the employees or a'vailable to all the j And most of the people that did make conversation or did talk i employees onsite.to the Nuclear Regulatory Comission live in that area, and some They're used to living and maintain-are not used to making this kind of money. ing a 53.00 an hour job and living in a one-room house with barely one car run-an hcur ning, and now they have a job at Comanche Peak making And not one or two beers in the i new clothes and having parties every weekend.And when they see that their friends hav l refrigerator, but a case of beer. turned against them, which is the way that they really see it, they set that if they talk against Brown & Root, they talk against their brothers and relatives (B&R), and they feel that, " Heck, we don't need you anymcre," and they do not counsel one another anymore and they do not talk to them. I l I O I l ~= m-y -~ -.; y -., ~

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g, .~ .g. i ? MECHANICAL & PIPING ALLEGATION REVIEW CATEGORIES l i l Category Est. Handays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks .+ l 15. Defective hangers AH-3, AH-4, AH-6, AH-13 16. llanger design AH-5, AH-8 nroblems 17. Hanger damage AH-9, AH-10, AH-11, j' 18. llanger construction AH-7, AH-12 i t \\ 19. Ililti bolts AB-4, AB-5, AB-6, AB-8 } 20. Concrete anchor bolt AB-9, AB-10, AB-11 j i 21. Nuts-torquing AB-7 l Q i y Y D [ m, c cc V

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,..._y. 'a.. i ' < FOLDER NO. 16 ^ Oraft No. Date COMANCHE PEAK ALLEGATION WORK PACKAGE MECHANICAL & PIPING Category 16 - Hanger Design Problems Allegation Numbers: AH-5, AH-8 Statement of Allegation: Improper design and design problems on pipe hangers with procedure concerns involving: a. Incorrect design b. Design problems, including threat that an alleger was not designing enough hangers per unit time Reference Documents: See source documents m_arked on attached pages from allegation list. Source of Allegation: Various - see enclosed allegations list Date Received: 1982 -1984 4 The above information prepared by D. M. Hunnicutt 6/4/84 Name Date '{ Group Leader Name Date Assigned Team Members Date Assigned I Date Assigned ? i Date Assigned I .i/ I f /f-960

t 9 l g-r2- ?* 1 Status: l Review Lead: i Support: 1 1 4 Estimated Resources: t 4 Estimated Completion: l CLOSURE: ? Reviewed by: 4 4 . 4 4 I/ P ' s 's e I i i i A i l i i -,-- u - - .~ e

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_. ~ _.. f-f I 6 cp 4 SSER 1. Allegation Group: Mechanical and Piping CateE.ory No. 16 - Intimidation to Hanger Redesign Suggestions 4 2. Allegation number: AH-8 redesign suggestions and failure to design #aa**g existed due to hanger Characterization: Itisallegedthatintimidgi 3. to hangers. 675 < M 4. Assessment of safety significance: The NRC Technical Rev_i_ew Team (TRT) reviewed the allentions _ as characterizeM?ne 4e&c.La1' allegation { implies that due to the inadvertent closing of a tunnel at one end, 7arge. bore pipe supports would not fit as ori; int.lly designed. ibbs and t

ll and TUSI engineers had concluded that an alternative design of the I

supports supposedly had to be devised. The allager indicated that he was laid off following a suggested redesign of the supports in questions. The allegation is very. vague and makes no reference to specific hangers or a specific tunnel location. The yR6 TRT reviewed the TUEC procedures pertaining to support fabrication, design control and various similar i Brown & Root (B&R) procedures. The yR6 TRT also discussed design of j large bore supports and the revision cycle of., support design and construction with TUEC pipe support engine.ering (PSE) and QC inspectors. Large bore pipe supports,a,r,egopqulv.redege,d due to interferences and pipe rerouting during the compfete, cycle or one design / construction process according to those interviewed. Based upon an extensive review of similar type supports for the assessment of the alle p ions in Category #31, the documentation in the hanger packages U HP) verifies these statements. j The second part of the allegation specifies that the alleger had been threatened with termination because he was not designing enough hangers. l By his own admission the alleger indicated that the alleged threats did not cause him to compromise his work and that he performed his work according to procedures. ,.,w j#. TheN/CTRTcouldfindnoevidenceofincorrectdesignordesign practices. Since no specific evidence was given, a review of the design procedures indicated that redesign of hangers is common and is controlled by approved procedures. The NR6 TRT was unable to talk to the alleger. After this assessment, theN/CTRTconcludesthattheallegationcenters ) around intimidation and not poor design practices. rs.: - ( ( y n m.m. g~~~- c.. u.~a a in,~,,,,. ;., .~. FOM 85-5s -a m ~

~ ^ -.-- -~ ' { I - Ii. 5. Conclusion and staf f positions: . i 6. Actions Required: Refer the allegation of intimidation to TUSI personnnel for their investigation. ' l I I 8. Attachments: } 9. Reference documents: TUEC Procedures CP-EP-2.1, CP-EP-4.0, CP-EP-4.6 ? Brown and Root Procedures CP-QAP-4.1, QI-QAP-il.1-28

10. This statement prepared by:

i Name Date 1 Reviewed by: Group Leader Date Approved by: Project Director Date

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. _... _. ~. _ _. ~. ,1.__.._.. - ~ ~. _. ff f f I V?t 4d ( w I A1'egation Group: Mechanical and i t to 1. l Hanger Redesign SuggeTcions \\, 2. Allegation number,: AH-8 3. Characterization: It is alleged that redesign suggestions and failure to ~7 4. Assessment of safety sistnificance: 'l reviewed the allecations as. character implies that due to the inadvertent c. large. bore pipe supports would not fit utbbs and ' Hill and TUSI ergineers had concluded. tve design of the supports supposedly had to be' devised. ..Aleger indicated that he was . laid off following a suggestediredesi of the supports in questions. i The allegation is very' vague and makes no reference to specific hangers s or a specific tunnel location. The'NRC TRT reviewed the TUEC procedures pertainingtosupportfabricatioN'e,designcontrolandvarioussimilar Brown & Root (B&R) procedures. NRC TRT also discussed design of large bore supports and the r ision cycle of support design and construction with TUEC pipe upport kugineering (PSE) and QC inspectors. Large bore pipe supports are comnglgredesigned due to interferences and j pipereroutingduringthe[complet'e,cycTeofthedesign/ construction } processaccordingtothoseinterviewedj Based upon an extensive review l ' of similar type supporys for the assessment of the alle p ions in Category #31, the documentation in the hanger packages uBP) verifies these statements. The second part of' the allegation specifies that the alleger had been threatened with termination because he was not designing enough hangers. } / By his own admission the alleger indicated that the alleged threats did l not cause him'to compromise his work and that he performed his work according cd procedures. / The NRC TRT could find no evidence of incorrect design or design 7 practices. Since no specific evidence was given, a review of the design procedures indicated that redesign of hangers is common and is controlled byhprovedprocedures. The NRC TRT was unable to talk to the alleger. After this assessment, the NRC TRT concludes that the allegation centers around intinidation and not poor design practices. P;M * ? Asf

'/ i l, _2-e l J 5. Conclusion and staff positions: ') 6. . Actions Required: Refer the allegation of intimidation to TUSI .; i personnnel for their investigation. 'W_ 1d j 8. Attachments: 1 9. Reference documents: TUEC Procedures CP-EP-2.1, CP-EP-4.0, CP-EP-4.6 Brown and Root Procedures CP-QAP-4.1, QI-QAP-11.1-28 1 '4 10. This statement prepared by: i Name Date - I

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f h -QQ& -.m gjd i Telechonic Interview of On August 28, 1 @3 M) a previously employed TUSI hanger engineer l was telephonically work 1ng as a tated he had interviewed by NRC Investigator H. Brooks GRIFFIN. testified at the ASL'8 hearings og CPSES during 1982 on the subject of personnel qualifications [ Atated he was tenninated during a ROF in July 1982, and took his case to " equal emplovment" (not further identified) and requested permission to sue TUSI.' aid he was advised he did not have sufficient grounds to sue and his case was not pursued. ( stated the events he believed resulted in his termination began with a lem with hangers located in a tunnel inadvertently closed at one l desi n p said the consensus of opinion among the Gibbs & Hill and TUSI large l end. l f pipe engineers was that the previously designed hangers would not fit in the tunne, and an alternative method for supporting the pipes had to be devised. f l ting, he recomended a solution to a level III tated th during a solution involved the redesign of the existing engineer named hangers. aid that. and another engineer named asked him ndicated g { to leave the meeting and shortly thereafter, he was laid off. rmination was discriminatory, and that he was laid off i, i he believed hat his k because and 'were jealcus of his suggestion. tated that on one occasion nd another engineer (not further identifie ) threatened him with termination because he was not designing enough s threat discourgace him, but did not cause him to 7d indic ed hangers. ~ ted ad been giving him audit work for compromise his work. stated while he worked at CPSES, he per#crmed which he received no credit. - his work properly according to the d sign criteria he was given, but stated the .itated he "got alcng" with the other engi-lt criteria was always changing. neers.

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1 v assi p.' g [- 9 1 OSHA regulations, which I understand are in effect cut 2 there. 3 Why don't they want their inspectors at 4 Comanche Peak to know what's going on? 5 At the South Texas Project, when I worked 6 there, they had a large technical library which was 7 available for us to use. Us being the inspectors. 1 question. we wanted answered, we j i l 8 When we had a t ? I d l c[ 9 had the library available there. Why don't they want i z I g 10 people to think and to learn at Comanche Peak? e j 5 11 Material verification. In the material I y M 2~

• verification group, I have probably come across fifteen tc

,g ,/ p p q, j j 13 twenty hangers whicli lacked completely any material gl i g 14 traceability. They had no heat numbers of any kind. I'd 8 .t.- E. E 15 turn them in to the material verification group and told j t 11 5 k.j' 16 Yhem I wouldn't sign them off, usually gave them to M ed I h, 17 ! I don't know what happened to them. As far as I f^, G\\ h. 18 know, they were never corrected. t e I9 my supervisor in,the hanger t 20l group, told me to buy off Hilti bolts which had already t 2I been torqued and had ' torque seal applied by someone else 22 when I came to make the final inspection. I followed his (., 'f ' 23 , request s. . g _, i i I and I \\'f ',. ! f 4 I've read the testimony of ,1 -(' n5 b /' can testify to many of the same thines 4.,aihed' h m.7 f y a ,.4].,% + g Q w,, ( l i ALDERSON REPORTING COMP ANY. INC. ,,,g

o - APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION REGION IV Report: 50-445/82-11 & 50-446/82-10 Occkets : 50-445 & 50-446 Category: A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower , Dallas, Texas 75201 Faciiity Name: Comanche Peak, Units 1 and 2 Inspection At: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection Conducted: April through September 1982 Inspector: b /2-3*fL-R. G. Taylor, Senior Resicent inspector - Date Construction (SRIC) i Approved: M /) 8L.- T. F. Westerman, Chief, Reactor Project Section A Date j Insoection Sumary Insoection Conducted During the Period Acril through Septemcer 1982 i (NTC Reoort 50-445/82-11) i Areas Insoected: Routine and special inspection, announced, by the Senior Inspector (Construction) including facility tours, as-built pipe and support l pro.; ram, investigation of allegations, and participation in 'ASLB hearing, The inspection involved 94 inspector-hours onsite and 510 hours on ASLB by 1 ona NRC inspector. 1 i Results: Within the areas inspected, two violations were identified: (1) failure to inspect installation of seismic shims in polar crane brackets, and (2) failure to properly train and indoctrinate QC personnel. 'j Inspection Conducted During the Period Aoril through Seotember 1982 (NRC Reoart 50 446/82-10) Areas Inscected: Routine, announced inspection by the Senior Resident i Inspector (Construction) consisting of plant tours. The inspection involved 35 inspector-hours by one NRC inspector. Results: Within the areas inspected, no violations or deviations were l icent1fied. o I

o. 16 '} Page 4851, lines 19 through 23: " Bill Snellgrove, my supervisor n v in the hanger group, told me to buy off Hilti bolts which had already been torqued and had torque seal applied by scmeone else when I came to make the final inspection. I followed his request." Finding - The SRIC has accepted this allegatten as true without .+ furtner interview or investigation since the alleged statement ' ~ is consistent with his observations of the installation and QC inspection process of hangers during the past several years. By way of explanation, the installation of the Hilti bolts into their holes is a very early event in the installation of a hanger since the bolts must be placed in holes that have been ~ drilled without cutting the embedded reinforcing steel. This requirement frequently results in bolt patterns somewhat nonsymetrical. Generally, after the bolt pattern is established, the pattern is measured and the baseplate drilled to fit. Very frequently, the baseplate will be then installed { - and the Hilti bolts tightened (torqued) to set the locking wedges. When this is the case, QC inspection is performed during the tightening process and the torque seal is applied. The QC inspection records for the support are also annotated that reflect that torquing has been accomplished and accepted. With the baseplate now in place, the installaticn of the remainder of the hanger would then take place and might extend i 9 over a period of several days, weeks, or even months. When the hanger is finally all complete, a final insoection would take place but without the necessity to reverify Hilti bolt i tightening if the torque seal has not been disturbed, hence, the idea of the use of torque seal. The alleger did not state that the supervisor's instructions extended to the acceptance of s j. obviously loose Hilti bolts or to those where the torque seal had been disturbed; he simply stated what is a nomal installation 3 j process with accompanying inprocess inspections. The allegation is substantiated but has no " technical merit." ohPage 4852, lines 19 through 25; page 4853, lines 1 through 7: "In a deposition of Gordon Purdy, he states: ' Reg. Guide 1.58 i says the only acceptable alternative to a high school education is the General Education Development Equivalency.' In my 4 conversation with Mr. Taylor, with the NRC, about this, Mr. Taylor indicated to me that as long as prior service has been acceptable to Brown and Root, the Grandfather Clause j generally would allow people to work without a high school education or a GED, to work as QC inspectors. What is the jj Grandfather Clause? Where is it written up? I'd like to see it, but Mr. Taylor mentioned this to me." 4 9 9 I w

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~.. . l,b L8P-83-60 h h,)3 UNITED STATES OF ANERICA NUCLEAR REGULATORY COMMISSION Before Administrative Judges: M Peter B. Bloch, Chairman i Or. Kenneth A. McCollem \\ Dr. Walter H. Jordan Docket Nos. 50-445 50-446 In the~ Matter of i TEXAS UTILITIES GENERATING COMPANY, et al. (Application for Operating License) e (Comanche Peak Steam Electric Station, Units 1 and 2) September 23, 1983 1 MEMORANDUM AND ORDER (Emergency Planning, Specific Qualiti' Assurance Issues and Board Issues) Our Proposed Initial Decision of July 29, 1983 invited the parties to coment on our tentative conclusions concerning emergency planning, Each of certain quality assurance contentions and four Board questions. the parties submitted objections. Applicant and CASE also submitted l replies, pursuant to our authorization. Th'is decision resolves the objections that were filed. In particu-lar, we have clarified our earlier intention t,o dismiss Contention 22 (emergency planning). We also have clarified the extent of our continu-This decision is called ing interest in the emergency planning issue. a " Memorandum and Order" because its effect is to affirm the declaration l of a default on some issues and to make interin factual findings that do The parties are the Staff of the Nuclear Regulatory Comission 1 (staff), Citizens Association for Safe Energy (CASE) and Texas Utilities Generating Company, et al. (applicant). PUlwoo-os 837569 "M3 .~..

Il 1 ~ Objections: 30 7# specific instances be identified, as applicant argues. Nor does testimony of a applicant's testimony concerning its procedures _ rebut contrary practice. Except to the extent indicated, applicant's objection is denied. M. Hilti Bolt Inspection of our decision is Applicant's objection concerning this aspect We accept the testimony of Mr. Ronald G. Tolson that Hilti i sustained. 8 1 that bolt inspection requires. the. completion, of.an inspection report i requires the reporting of four attributes, including as one attribute 70 Given that the " torque wrench number" and the " calibration due date." testimony we reject any possible inference from Mrs. Stiner's testimony j f that quality assurance inspectors would fake these numbers because they bolts on which they found had received an instruction to " pass" Hilti torque seal. We are confident that a review of these inspection reports should indicate whether torquing has been proper. Consequently, since applicant's inspectors do review the

reports, the objection is sustained.

l i 74 A plicant's Objections at 53. 75 Applicant's Objections, Attachment C, " Affidavit of Ronald G. Tolson" (Tolson Affidavit) at 2. e e*

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t.....-..... i .. ~ ATTACHMENT C / b khl0 er 'Vb 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) Docket Nos. 50-445 and TEXAS UTILITIES GENERATING ) 50-446 COMPANY, et al. ) ~ ~ ~ ~ ) ( Application for (Comanche Peak Steam Electric ) Operating Licenses) Station, Units 1 and 2) ) AFFIDAVIT OF RONALD G. TOLSON REGARDING "OPEN ISSUES" IDENTIFIED IN PROPOSED INITIAL DECISION I, Ronald G. Tolson, being first duly sworn do depose and state as follows: I am employed as the Site Quality Assurance 6 Supervisor at Comanche Peak. As such, I am familiar with the QA Program at Comanche Peak, including the QC inspection processes. I have previously testified in this proceeding regarding the QA Progran at Comanche Peak. A statement of my educational and i { professional qualifications was received into evidence as Applicants' Exhibit 20. This affidavit addresses two "open { issues" identified by the Board in its July 29, 1983, Proposed Initial Decision. The "open issues" which I address concern the Hilti Bolt verification process and an allegation regarding nondestructive examination on the fuel pool liner. Hilti Bolt Verification Process l The Board has identified as an "open issue" the adequacy of the Hilti Bolt verification process. Specifically, the Board { states that it is not able to determine whether thatl=

~ / / _2_ / f would disclose instances in which Torque Seal had been applied to Hilti Bolts by craft personnel, rather than by or in the presence of OC inspectors. As I discuss below, the record of Hilti Bolt inspection involves verification of several attributes other than the application of Torque Seal. The procedures applicable to Hilti Bolt inspection provide that QC inspectors either witness directly the torquing operation or verify that the proper torque value was achieved following torquing by craft personnel. Specifically, the following information is to be documented on the inspection report: (1) Verification that the torque wrench is cali-brated and the calibration is current. (2) Verification that the torque wrench setting is in accordance with the applicable re-quirements for the bolt being torqued. (3) Verification that the correct size socket for i the bolt being torqued is used. (4) Both the M&TE number (torque wrench number) and calibration due date. The inspection report for each support or fixture which is secured by Hilti Bolts w'ill also include a unique identification number for that support or fixture. Torque Seal is applied to the bolt following the torquing operation. See QI-OP-11.2-3, i " Torquing and Spacing of Concrete Anchor Bolts" (copy attached). 1 I

, _. ~ _. u.. O : In sum,'whether a QC inspector personally applied or observed a'pplication of the Torque Seal on the bolt is not alone significant. The significant consideration is that there must be i e a complete inspection report for each support or fixture or a reinspection and reverification of torquing will be required. Fuel Pool Liner Another "open question" ident'ified by the Board concerns an allegation that a craft person was performing liquid penetrant testing on the fuel pool liner. Apparently, the Board had drawn the implication from the testimony in this proceeding that that proposed person was performing the testing for QC purposes. Initial Decisior. at 39. However, as I discuss below, an investigation of this allegation revealed that this testing was not being performed for QC purposes. ,The NRC conducted 'an investigation into -this particular 1 8 allegation. During that invastigation it was determined that the testing craft person who was observed performing liquid penetrant wau performing in-process examination of work for craft purposes i prior to formal QC inspection a.7. testing. Craft personnel may d although that test-perform such testing for their own purposes, .i t ing is not recognized by QC for inspection and testing purposes. i t l l Y ~" t y .q _. - - g_,__

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~4-l i i The fuez. peo2 IIxer has been inheestently Sacrectid Ja accarnace with the j I gc ine2.czio ragmire=emen. i I A_ Kon32e c., M oa v ~ j .g.I h sescribes and swarm to befera sur tMs.O.' day of Jtapst, 29a3. E+ / rr Qg / &% 451mor $c V ,qwey a cf PrP rr e & ' i *no.,. y

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e mas.y er succean n State of 7enas ) Y g e h 5 Os M ~- ,, ' t ) I i ci k I "I This is a telecopy facsimile. The original will be transmitted under separate cover. y i Ei l n. I tl i. YE.i 1 e 1 ! i i i A

A 48- / Y~U-"' f ' Y-11 ~ 14 i On March 1, 1983, , Hilti bolt QC Inspector, CPSES, was interviewed on site by NRC Investigator R. MEEKS @ stated is g supervisor and that ghas experienced problems in working for him. related that on several occasions, craft personnel have presented Hilti bolt documentation packages which they represented as being satisfactory to W for QC inspection. g stated Welieves that based on what the craft personnel had told him, the inspection of these items should be approved despite the inspectors recognizing that f the documentation packaga does not conform to CPSES procedures. O stated ghas never approved a package which did not meet procedural requirements. W was unable to provide any additional information pertinent to this investigation. C f 9 l~ L F01A-85-59 l- - = - -

......s. 9 ~ Qm \\. i h UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before Administrative Judges: i Peter B. Bloch, Chairman i Dr. Kenneth A. McCollom Dr. Walter H. Jordan In the Matter of ASLBP No. 79-430-060 1 i TEXAS UTILITIES GENERAT'ING COMPANY, et al. Docket No. 50-445-OL l 50-446-OL i l (Comanche Peak Steam Electric Station, Units 1 and 2) July 29, 1983 PROPOSED INITIAL DECISION (Concerningaspectsofconstructionquality) control, emergency planning and Board questions i This is the first initial decision in this operating license proceeding. The issues we decide today relate to particular allegations" of failure of Texas Utilities' Generating Company, et al.'s (applicant's) quality assurance program during ' construction of Comanche Peak and'to I ~ j ~ Hearings on allegations concerning the adequacy of emergency planning. ~ these particular allegations were completed on September 17",19fi2 and~ proposed findings of fact were received frain the parties on February 25, 1983.2 1 Although " quality assurance" and " quality control" sometimes are used in a specialized way, we generally will use these words interchangeably in our opinion, p ens The State of Texas did not file p Association for Sound Energy (CASE or rvenor) filed proposed g

1. ~

h '" Initial Decision: 2

i

~ 3 In this decision we have declared CASE to be in default on each allegation on which it has not filed findings of fact. However, we also have examined each important allegation that is in default in order to 3 determine whether to raise any of these defaulted issues by ourselves (sua sponte). See 10 CFR 5 2.760a. In a few instances, we require some additional evidence before determining whether or not to declare a.sua 1jj sponte issue. Two of the three members of this Board were added to it after the ] j hearings on the matters we address. Consequently, we have adopted the unusual procedure of issuing a proposed decision. The consequence of calling this a proposed decision is that the Board recognizes that its record is complex and that it is wiser, under the circumstances, to , ~ ' invite coments on our tentative conclusions before we become committed j' to theat. We expect the parties to object to any aspect of this decision which they belfeve to be in error. Objections must clearly and logically explain what the suspected error is and the legal and factual arguments on which the objection is based.3 Failure to comply with any l, t i l l findings on some of the allegations discussed in this decision. ~ CASE indicated that it wished an extension of time to file proposed findings on additional issues. However,' the parties had been granted a six week extension for filing of proposed findings. Reconsideration of December 7, 1982 Order at 1-2 (December 21, 1982). Under the circumstances, good cause has not been shown for granting CASE more time in which to file its proposed findings. 3 We received Applicants' Summary of the Record Regarding Weave and Downhill Welding (July 15,1983) during the preparation of this .-_w

- ^ 2. t Initial Decision: 3 l aspect of this required format may resul t in a default on the cb-jections. f The Board had posed certain questions and taken some limited evidence, to determine whether there is a serious question that the Board should raise by itself. See 10 C. F.R. I 2.760a. The Board concludes that it is not now raising any of the questions sua sponte, il 'I I. Contention 5 and the Applicable Law Contention 5 states: i The Applicants' failure to adhere to the quality assurance / quality control provisions required by the construction permits for Comanche Peak, Units 1 and 2, and the requirements of Appendix B of 10 C.F.R. Part 50, and the construction practices employed, specif-ically in regard to concrete work, mortar blocks, steel, fracture toughness testing, expansion joints, placement of the reactor

3 vessel for Unit 2, welding, inspection and testing, materials used, t

craft labor qualifications and working conditions (as they may affect QA/QC), and training and organization of QA/QC personnel, have raised substantial questions as to the adequacy of the con-i- l struction of the facility. As a result, the Commission cannot make l the findings required by 10 C.F.R. 5 50.57(a) necessary for issu-ance of an operating license for Comanche Peak. c{ i' lI 4 l decision. Since other parties have not had an opportunity to consnent on this filing, we have not considered it. However, the I}l filing is a model of the kind of specificity we expect in objections to this decision. It includes footnotes to the record. 5l' It also uses an affidavit, which parties may file in support of 'l their objections if they consider it helpful. lll j l r t 1 d 1 4 [; iL .. _. _ _. =... = _ _ _ _ _,

~ Initiaf Decision: 4 - o l I Order subsequent to the Prehearing Conference of April 30, 1980, slip op, at 11 (June 16,1980) (unpublished). As the Board has previously indicated,4 this is a broad contention calling into question the applicant's entire quality assurance program. The Atomic Energy Act of 1956, as amended, i 103, 42 U.S.C. 9 2133 (1976), authorizes the Nuclear Regulatory Comission to issue licenses j for nuclear power plants to applicants "who agree to observe such safety standards to protect health and to minimize danger to life or property as the Commission may by rule establish...." The Comission has, by i rule, required that "[s]tructures, systems, and components important te safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be perfonned." Criterion 1, Appendix A - General Design Criteria for Nuclear Power Plants,10 C.F.R. Part 50. A quality assurance program is required to ensure that the safety functions will be properly performed. i, Id. Criteria for the quality assurance program are set forth in i Appendix B to 10 C.F.R. Part 50 and must be discussed in the Preliminary Safety Analysis Report (PSA,R) for construction permit applications and in the Final Safety Analysis Report (FSAR) for operating license applications. 10 C.F.R. I 50.34. The chief concern of the quality assurance program is to identify and correct problems that arise during plant construction or operation. 4 Tr. 714. i 'd_ _ _ eyyew - a m --=s-empe ww-M-sr-we - -- - +s

~ ^ Initial Decision: 5 3 Indeed, a quality assurance program that failed to find problems would undoubtedly be ineffective. A problem identified by the quality assurance program may cause concern for the public safety if it cannot be satisfactorily resolved. A program may also cause concern if it identifies an extraordinarily large number of deficiencies, casting doubt on the plant's design and construction processes. Additionally, if a quality assurance program identifies extraordinarily few deficiencies or if we were to find that substantial numbers of deficiencies have been overlooked, that may raise questions about the adequacy of the quality assurance program. At this stage, we are not evaluating the overall efficacy of the quality assur-ance program, but, rather, whether any of the alleged deficiencies are 3 sufficiently serious and uncorrectable that the plant, due to those deficiencies, cannot operate with the requisite degree of safety. In other words, we have considered each allegation independently, without regard to whether it may represent a pattern related to the adequacy of the quality assurance program. In addition, there are particular allegations which have been or will be the subject of 1 hearings held after September 17, 1982. These questions are not re- ~ I solved by this decision. A. Rock Ove.rbreak j One of the specific allegations encompassed within Contention 5 was that during construction blasting, rock which was supposed to remain 4 .__n ,,.-..1,-w- --n,- wm m ~m r.~

'i Initial Decision: 6 l t intact was displaced and cracked. The foundation for Comanche Peak is 1 set on a rock structure known as the Glen Rose limestone. This is a marine formation of the Cretaceous age.5 The Glen Rose limestone is " soft" or " weak" rock and is not homogeneous.7 Both of these factors 6 make it difficult to predict and control the effects of blasting in this particular rock. The applicant's geotechnical consultants, Mason-Johnston and

l Associates, Inc., recommended that all safety-related structures be placed against intact rock 0 It was planned that once the site had been o

leveled off at plant grade (i.e., the mountain top had been removed down 9

i to a specified leve1 ) excavation for the reactor buildings would assure that their bases would be 40 feet below plant grade.10 Explo'sives were

~: used for the purpose of creating a crack in the rock around the } -- - J perimeter of the hole for each containment building. Further blasting was contemplated to break ~ up the rock within that perimeter crack.11 Il lr 5 Tr. 803. 6 l Tr. 835, 946. 7 Tr. 957-8, 1210. O Tr. 809. i 9 Tr. 806-07. 10 j Tr. 808. 11 Tr. 809-814. 6 m m.. __ -er*'%?_*'-~~

~ ~ Initial Decision: 7 I u ll Une::pectedly, the blasting to create the hole for the Unit I containment caused approximately the top - ten feet of the rock wall I around the planned hole to be displaced and to suffer both horizontal, and vertical cracking.12 Although changes were made to the blasting procedures when the hole for the Unit 2 containment was blasted, a f similar overbreak pattern emerged.13 Overbreak also occurred in excavating for other safety-related buildings.14 Indeed, the overbreals 4 was so extensive that there was no point in associating particular j fractured rock with the excavation of a particular building.15 Applicant reported the overbreak to the NRC pursuant to 10 C.F.R. 9 50.55(e).10 The extent of the overbreak was determined by digging trenches at ~'i increasing distances from the excavation wall and examining them for cracks. When a trench was dug for which any cracks on the side closest to the excavation were found not to have been propagated to the far side of the trench, that was determined to be the edge of the overbreak.17 f N 12 Tr. 815-16. .s 13 Tr. 829. 14 I Tr. 831,1208-09,1269-1272. 15 Tr. 1270. 16 Tr. 845-46, 1270. I 50.55(e) requires that serious breakdowns be reported to the NRC. 17 Tr. 819-21. i I

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Initial Deci'sion: 8 i Overbreak was detected up to approximately thirty feet from the exca-vation.18 While some overbreak is common during construction, the amount of overbreak at Comanche Peak was unexpectedly large.19 To " repair" the overbreak, all rock within the area of the i overbreak was removed utilizing techniques which did not require .l)t blasting.20 Once this rock had been removed, some cracks were found in ) the newly created walls.21 These cracks did not appear to be associated with displaced rock and were -grouted.22 The original shape of the .a excavation was restored by filling with concrete the area from which fractured rock had been removed.23 This " dental" concrete was less strong than that used in construct-ing the reactor buildings; nevertheless, it was stronger than the fractured rock had been.24 There was uncontroverted testimony from the applicant's panel that, once these repairs had been effected, the i foundation was.actually improved from what it would have been if the i 18 Tr. 820. 19 Tr. 835,1115,1209-12. 20 Tr. 817, 821-22. j; 21 Tr. 823-24. ( 22 Tr. 832-33, 1272. 23 Tr. 817. 24 Tr. 827, 955-56. Ii I qu-MLMey' * - F_-- e 1. N*eMM

___1--- Initial Decision: 9 a i! t i e i: overbreak had not occurred.25 The applicant's witnesses also testified, l' lj without contradiction, that the seismic capacity of.the site was not impa' ired by the repair.20 The NRC cited the applicants for failing to have and to utilize quality control procedures for excavation for these safety-related structures.27 Applicant subsequently developed such procedures.20 ~' It

)

(i is apparent, however, that, even without quality assurance procedures for this excavation, the applicant detected and took action to remedy f the overbreak. Moreover, the soundness of all areas excavated before the quality control procedures were instituted was verified by an engineering geologist who was present during all phases' of the exca-4 - The engineering geologist verified the soundness of all the vation. i qf !I H materials on which concrete was to be placed. In making this judgment, <g. l he rel'ied on professional knowledge and confirming photographs.28 The repair work, in addition, was subject to quality assurance procedures.30 While it was a quality control deficiency to have done the exca-vation without quality control procedures, we find that this has not led .i 25 Tr. 335, 838. 26 Tr. 957-58. l 27 Tr. 1051, 1273, 1279; CASE Ex. 15. 28 Tr. 1055. 29 s I Tr. 1047-48. g 30 Tr. 843. \\, =<; _ . m z._ m. - ~ __-

~ l '.... 1 -- Initial Decision: 10 '? i I I to a lasting deficiency. This potential problem has been negated by the detection and satisfactory repair of the excavation defects which should have' been detected by the quality control program. The possible, implications of management's failure to implement quality assurance procedures for the excavation activities is not being considered in this decision. I. l. B. Cracks in Concrete Allegations were raised that cracks were present in the basemat for the containment. CASE presented no witnesses who addressed the l allegations. However, CASE did introduce as exhibits a non-confonnance report (NCR) and a revision to it which refer to cracks "on 812' Base j Mat Containment #1".31 Using the pour nunters given in that same NCR and revision,32 1 witnesses for the applicants and the staff detennined that the cracks in 33 question were not in the basemat but were in the reactor shield wall. l. This is part of the internal concrete which is, in fact, separated from the twelve foot thick concrete basemat by a steel liner.34 The concrete I L r 31 CASE Exs. 8 and 9. 32 Tr. 1363. 33 Tr. 1011-13, 1363. 34 Tr. 850-54,1181. l. / ~ m

~ l Initial Decision: 11 i I i d pours in which the cracking occurred surround the reactor vessel. 5 The wall of which these pours are a part is steel reinforced.36 Its main func' tion is the provision of radiation shielding.37 t The witnesses uniformly concluded that these were shrinkage cracks, caused by the shrinkage of concrete as it cools.38 These are essen- } tially hairline cracks into which something as small as a pencil lead would not fit.39 Cracks of this type are not unusual or troublesome i.j. when found in large concrete pours.40 l The staff witnesses believed that the cracks were about two inches deep.41 The applicant's witnesses believed that the cracks went through the entire depth of the pour.42 In determining the structural adequacy [, of the shield wall, the assumption was made that the cracks went 'k entirely through the pour.43 In addition, expert testimony was given by the architect-en,gineering firm responsible for the design that a 35 Tr. 856, 859-66; App. Exs. 21 and 22, 36 Tr. 855. ~ 't 37 Tr. 865. t I 1 38 Tr. 867, 870-71, 1384. 39 Tr. 867, 1198. 40 Tr. 869. 41 Tr. 1374-75. 42 Tr. 960. 43 Tr. 1398. i o ,m.._ 4,, =

Initial Decision: 12 i t. construction joint could have been placed at approximately the location of the shrinkage crack, since the concrete pour was so large. The 1 femation of the shrinkage crack is said to serve much the same purpose as a construction joint at this location might have served.44 { 1 Unrebutted testimony was presented that these cracks in the con-crete did not impair the wall's capacity to perform its intended functions.45 Radiation shielding would not be affected.46 Nor would f the wall's ability to transfer vertical loads.47 The reinforcing' steel (rebar) in the wall is relied on to take tension loads." A potential problem that was examined is whether the crack would admit water that would rust the rebar. However,'there is unlikely to be 1 water reaching the reinforcing steel through the any problem with 4 cracks, causing rust and weakening the steel's load bearing capacity.49 3 'After a short time, the exposed surfaces of the cracks were grouted to present a smooth face for painting.50 The grout excludes water. l i 44 Tr. 882-83. 45 h Tr. 885, 1295. 40 Tr. 885-86, 1301. 47 Tr. 869, 1300. l 48 Tr. 866, 885. l. 49 Tr. 897-98, 1022, 1028. 50 Tr. 1205, 1313. l l l ..e--m._ ,upmn

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Initial 0:: cision: 13 s ( These cracks were identified through the quality assurance system, which illustrates that, at least in this regard, the system was func-l tioning at Comanche Peak. They have been analyzed and minor repairs made. Based on the record before us, we conclude that the repair was adequate and there is no safety problem associated with the cracks. C. Other Specific Allegations Raised in the Context of Contention 5 CASE presented several witnesses who made allegations about 'h construction deficiencies and deficiencies in the quality control system at Comanche Peak. However, CASE did not include these. matters in its proposed. findings even though the proposed findings were mandatory. See j;i Order (Proposed Findings of Fact; CASE Exhibits) at 3-4 (December 7, 1982).(unpublished). Because this Board is consequently left to specu-late about what CASE currently contends about these issues, its failure to file proposed findings constitutes abandonment of this portion of its case. See Southern California Edison Co. (San Onofre Nuclear Generating 'l Station, Units 2 and 3), ALAB-717, 17 NRC , slip op. at 48-49 (March 4 4,1983); Detroit Edison Co. (Enrico Fenni Atomic Power Plant, Unit 2), ALA8-709, 17 NRC (January 4, 1983); 10 C.F.R. I 2.754(b). I We therefore find that CASE has abandoned the allegations on which i it has not filed findings. We have, however, considered whether these allegations raise such' serious questions of public health and safety 'i that we should raise them as sua sponte issues. We discuss briefly why, based on the record, we either have decided not to raise 'these matters l ~ e.._ - -, - - + __ ~- - &. t -- -J.~. ?.. n., n _ -.u_-n-

Initial Orcision: 14 ) sua sconte or have indicated that further information is required before j we can reach that determination.51 1. Allegations by John Junior Gates CASE witness John Junior Gates was a carpenter who worked on the Comanche Peak site from Novembe[1976 to March 1979.57 Much of his I related t,o' the fact that work was done and, due to design concern i changes, had to be ripped out and redone. He related this to increased cost and icw worker morale. As Mr. Gates himself testified, rework and low worker morale does not affect the quality of the plant if ultimately h the work will only be approved when it is done correctly.53 Mr. Gates testified that he did not know whether in the end the work was done ,i correctly.54 The fact that rework was required, however, suggests that approval was not forthcoming unless the work met the specifications. The Board finds that this allegation does not by itself 'aise a serious r health or safety issue. i;t i l 51 Some of the allegations could relate to the question of the extent of management's commitment to quality control. That issue, of course, we have specifically left open in this decision. In addition, matters which relate to allegations made by CASE witnesses Mark Walsh and Jack Doyle or to issues raised by the, staff's Construction Appraisal Team report also remain open. 52 j Testimony of John' Junior Gates, CASE Ex. 651, at 5. 53 Tr. 2795, 2820-21. 54 Gates Testimony, CASE Ex. 651, at 21. ' ^ s., hd- ... e..- 4 s s .--.*f_- m--'-

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~ ^- x.. ^ .~ fnitial Decision: 15 t l Mr. Gates also alleged that at one point work on the steel liner for the Unit 1 containment was halted because the steel liner was four inches out of plumb.55 The applicant agreed that the liner had been out i of alignment, but applicant's witnesses testified that this had been corrected to within specified tolerances.56 Construction was stopped and stiffeners were added to the inside of the liner, bringing it back in line before continuing construction.57 This matter does not raise an 3 issue which the Board finds it should pursue sua sponte. { According to Mr. Gates certain water stops were improperly in-stalled.58 He mentioned that nails were put in the water stops, taching them to the foms. A water stop is a neoprene strip half of which is

i embedded in concrete on each side of a joint in a wall.

Since the nails are located at a distance away from the actual joint, applicant assured ^ the Board that there is no problem of leakage from this normal-procedure.59 The foms are removed, together with the nails after the concrete is poured. The applicant also testified that, as of the time of the hearing, no water seepage was occurring through the outside walls 55 Gates Testimony, CASE Ex. 651, at 37. ,l 56 l Tr. 2988, 2992. 57 l Tr. 2992-93, 2995-3000. 58 l Gates Testimony, CASE Ex. 651, at 37. 59 Tr. 2989-90. I u; -. mE ,..om e --e ---k.--am--- --mu -4 o m

Initial Decision: 16 4 of any safety-related building.60 For these reasons, we see no need to pursue these questions further. 'Mr. Gates also alleged that the concrete work at Comanche Peak was j " sloppy." He mentioned observing honeycombing, watery concrete, and materials left in the concrete.61 His knowledge, however, did not seem ..^ to extend to whether possible problems have been corrected.62 To illustrate the problems with the concrete work, Mr. Gates pointed to a photograph of the containment buildings.63 He stated, however, that the i j defects which he believed were shown by the photographs would not have any structural significance.64 This lack of structural significance was confirmed by Ralph McGrane, a professional engineer who appeared as a i witness for the applicant.65 The Board finds that Mr. Gates' -( allegations about concrete work do not contain any specific information which causes the Board to be concerned with the structural integrity of safety-re! t.ted concrete at Comanche Peak. Hence, the Board will not raise this, issue sua sponte. i ( 60 Tr. 2993. 61 Gates Testimony, CASE Ex. 651, at 24-26, 36. 62 See, e_. 3. Tr. 2842; Gates Testimony, CASE Ex. 651, at 26. 63 Gates Testimony, CASE Ex. 651, at 24; Board Ex. 4. 64 Tr. 2883. j 65 Tr. 2990-92. 9

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Initial Decision: 17 i i l 2. Allegations by Stanley G. Miles CASE witness Stanley G. Miles was employed at Comanche Peak from March 1977 to May 1982.66 Like Mr. Gates, he was concerned about low worker morale.07 Low morale alone, assuming that it could be adequately + i - defined and measured, does not raise health and safety concerns. Only if low morale causes defective work to be accepted as the final product would this cause us to question the safety of the plant. We have no reason to believe that defective work was accepted. If a specific instance is brought to our attention, we can, of course, address that i i specific instance. As a general matter, however, we have no reason to 1 raise a question about the possibility that low morale has, by itself, led to an unsafe plant. 'Mr. Miles also alleged that in one instance he was instructed to do work for which he had not been provided a blueprint.68 He alleged that 1 j - he had welded boom struts on rigs manufactured by Manitowoc, in violation of the conditions of Manitowoc's warranty.09 He alleged that I 66 Testimony of Stanley G. Miles, CASE Ex. 655, at 1. Ol H. at 49. 68 Miles Testimony. CASE Ex. 655, at 31. 09 Supplementary Testimony of Stanley G. Miles, CASE Ex. 657, at 4. -.x y* . m e*=-eim_=.e ~,-e ,wa - y_ =*r 2, .. - - ~ -,

• 9

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~" '~ ..._1..._____.'..... ( a panel was made to appear to have been anchored with Hilti bolts when it had not been.70 'None of these allegations would have safety significance for the plant. Neither Mr. Miles' work without a blueprint nor the false Hilti bolts, apparently anchoring the panel, occurred in a safety-related area.71 The Manitowoc rigs appaie'ntly are no longer even used on the Comanche Peak site.72 l Il. Mr. Miles also alleged that there were problems with the polar crane. He alleged that there were 3/8 inch gaps between each longitudinal section of the rails on which the polar crane ran. As the crane was operated, the rails could move, accumulating some of the gaps so that as much as five inches could be found in a single gap. When the .l polar crane wheel dropped into this gap, it would stop.73 However, Mr. Miles 'also testified that the problem had been corrected by the clips that he personally had installed. These clips were made of a weld-on piece, a bolt-on piece and the bolt itself.74 So there is no defect 4: j,l remaining that the Board might declare to be a sua sponte issue, i.! 6 70 Miles Testimony, CASE Ex. 655, at 26-27. 7I Id. at 31, 27. 72 Miles Supplementary Testimony, CASE Ex. 657, at 5-6. 73 i Tr. 2932. I' 74 Tr. 2978-79. f

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Initial Decision: 19 i-t I Also, in connection with the polar crane, Mr. Miles alleged that, contrary to the design documents " fingers" were cut off of several shim's to make them fit in their designated places.75 It appears that this did occur. However, the Board does not believe that it is a matter which the Board should pursue sua sconte because it appears that the staff and the applicant are addressing it. The staff issued a Notice of Violation in connection with the failure to inspect these shims.76 The a'pplicant has stated that all the shims in the polar crane girder support bracket assemblies will be removed and inspected. Shims which have clipped " fingers" will be evaluated by an engineer to detennine whether they are acceptable.77 3. Allegations by Cordella Marie Hamilton and Robert L. Hamilton / Mr. Robert L. and Mrs. Cordella Marie Hamilton appeared as a panel; r l both raised coni: erns about the quality assurance program for protective coatings. Mrs. Hamilton worked as a documentation clerk for the protec-tive coating quality assurance program.70 Mr. Hamilton was a quality i 75 Miles Testimony, CASE Ex. 655, at 17. 76 Staff Ex. 1488. 77 Testimony of John T. Merritt, Jr. Regarding Placement of Shims in Polar Crane Glider Support Bracket Assemblies, App. Ex.127, at 6. 70 Testimony of Cordella Marie Hamilton, witness for CASE, CASE Ex. l 652, at 1-2. '"4 h_ ..a - ~. - --y --m .m

~' z : -...... ( assurance supervisor for the protective coatings.79 Certain allegations in this area were made by Mr. Hamilton, certain allegations were made by Mrs.' Hamilton, and certain allegations were made by both. Mrs. Hamilton alleged that there were deficiencies in documentation for protective coating quality assurance. Specifically Mrs. Hamilton ~ alleged that some paperwork required a large number of revisions or was never corrected and that approximately fifteen quality assurance in-spection reports were lost.80 Mrs. Hamilton testified, however, that the problems were. identified through an audit and all necessary cor-rections were made.81 Reinspection was required because documentation was missing. Although Mrs. Hainilton did not believe the required reinspection took place, because the material was out in the field and 'I had been cut up,82 she indicated that the lost documentation had not indicated any deficiencies.83 Therefore, it appears that the documenta-f tion problems identified by Mrs. Hamilton either have been corrected or had no safety significance. In addition, it appears that the procedures f f 79 Testimony of Robert L. Hamilton, witness for CASE, CASE Ex. 653, at 1. 80 C. Hamilton Testimony, CASE Ex. 652, at 5, 9-10. 81 Id. at 8, 9, 14. It is not clear from Mrs. Hamilton's testimony 'l Wether there was an audit by the applicant or an inspection by the NRC. However, an NRC inspection report did note deficiencies in i j the program in 1981. See App. Ex. 44B. 82 C. Hamilton Testimony, CASE Ex. 652, at 15,18. 83 Id. at 18. 0 6 d, h,.M-- ,h %g,, y 4 m =m e,-- ,_,,..,, ewe; ,-e e=,, e ; y.7 w w

I governing paperwork for protective coating quality assurance were changed in July 1981.84 - Thus it would seem tha't the source of the alleged problems also has been addressed. The Board finds nothing to raise as a sua sconte issue arising from this allegation. f Mrs. Familton was also concerned about the specified method for i j determining " tack-free time" for paint, but she admitted that she did not know that the method being used was incorrect.85 She further alleged that there were problems with the calibration of instruments i used in paint inspections.86 These two allegations, made by a lay witness in very general tenns, are insufficient to indicate a deficiency or to serve as a basis for a sua sponte question by the Board. They ? appear to be in the nature of general questions rather than the identi-fication of specific problems which might require a serious inquiry. Mrs. Hamilton alleged that for a year quality control inspectors were directed not to write NCRs on work done by the paint department.07 Mr. Hamilton testified that he was directed to stop writing if not any, at least so many, NCRs.00 Neither the staff nor the applicant addressed these allegations directly. These allegations are, however, closely t 84 jd.at13. 85 Id. at 18-19. 86 Id. at 21-22. 87 . at 19. 88 R. Hamilton Testimony, CASE Ex. 653, at 22, 43, 53-54. ~< j>

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Initial Decision: 22 I ( related to the issue of management's comitment to the quality control program. As such, they remain open. In addition, Mr. Hamilton's allegation that craft personnel ha-99 rassed quality control inspectors is related to management's attitude ll towards the quality assurance program. If management permitted or ,i ll failed to discourage harassment of inspectors, that would, of course, ref. lect adversely on that attitude. The question of whether this has occurred remains open. Mr. Hamilton alleged that he and two other inspectors were fired for trying to do their inspection job.90 The given reason for his firing was that he refused to make an inspection under what he believed

l were unsafe working conditions.

However, he alleged that other indi-viduals who refused to make the same inspection under the same conditions were not fired.91 This allegation is relevant to the appli-

j cant's attitude towards the quality control program and will be dealt with in a subsequent decision.

j. Mr. Hamilton also alleged that his quality assurance supervision was not qualified.92 While this could also reflect on management's il 3; commitment to quality assurance, we note that Mr. Hamilton's concerns in 89 Id. at 36-38. t 90 H.at26. 91 Id. 4 9 ll M.at14,16. 1 N. -.~ - 7

Initial Decision: 23 this regard apparently related to his opinion that he was more knowl-edgeable than his supervisors in the area of procedures for quality assu'rance inspection of paint.93 He disagreed with changes in proce- ~ dures and objected to supervisors overruling quality control inspectors on specific inspection findings.94 We do not believe that an employee's disagreement with a decision made by his supervisor raises sufficient t questions about the supervisor's qualifications that the Board should raise supervisor qualifications as a sua sponte issue. However, there appears to be a gap in the record because neither the applicant or the staff has testified about whether the procedures which Mr. Hamilton criticized were, in fact, acceptable. (Specifically Mr. Hamilton has I questioned the lack of standards for determining near white blast for Il" surface preparation, the lack of a maximum roughness for steel substrate 4 surface, and procedural changes which allegedly reduced all painting inspection to adhesion testing.) If evidence introduced on this open i I item led us to conclude that there were significant faults in these f inspection procedures, that could lead us to question the qualifications of the supervisors who. approved them as well as the adequacy of the inspections performed following them. We note, however, that this does 1 93 H. a t 16, 39, 43. 94 g.at15,16,18-19,38. x me a ~ - - n w __. .n~ x- -:w...-.- _n. ,p. _. m ,,..n

Initial Decision: 24 l i / not indicate any present conclusion by the Board that problems exist in this regard. . 'It also appears that Mr. Hamilton disagreed with dispositions of, t NCRs. He disagreed with the repair method for minor defects in paint-ing.95 He questioned the disposition of an NCR which addressed contamination of a painted surface which was force-cured using Kelly heaters which smoked.90 Once again, these specific allegations were not addressed in the hearing by either the applicant or the staff. Since i the implications of a failure to provide adequate disposition of NCRs could be serious, we consider this an open item. In particular, we need i c to be able to evaluate the disposition of the NCR related to smoking i Kelly heaters and the procedures for repair of minor defects, as specified by Mr. Hamilton. There is one other allegation made by Mr. Hamilton which the Board is currently unable to evaluate. Mr. Hamilton alleged that a paint j applied by Westinghouse and not tested by the applicant's quality 4 95 Id. at 15. 1 Id. at 21-22. He also questioned the disposition o'f an NCR on I 90 i Wich he had documented what he believed was grease in paint. Id. l at 21. However, it was shown that the paint was ultimateTy returned to the vendor. Tr. 3502; App. Ex. 139. While Mr. l; Hamilton believed some of the paint had been used, he had applied 'j hold tags to the containers, the amount of paint missing was small (less than 2 containers), and some of the paint would have been used in an attempt to strain it and remove the foreign matter. Tr. i 3503-05. The Board is satisfied that this NCR was properly dispositioned and does not raise any question of the safe operation of the plant. r i i l g w&-. =. m ..s. __,_m-- t m _,_ - t-.ev. w - __..en.-----

Initial Decision: 25 ( control program could not pass an adhesion test.97 We need to determine whether this use of paint is safety-related and, if so, whether the paint will perform satisfactorily. Mr. Hamilton alleged that an audit of Carboline, a paint vendor, was a " white-wash."98 This audit followed two audits. which had found unsatisfactory conditions at the Carboline plant. Mr. Hamilton was not present for the audit which he felt was a white-wash and based his charges on a " gut feeling."99 He admitted that the sequence of events concerning Carboline would appear to illustrate the satisfactory perfor-mance of the quality assurance program in resolving unsatisfactory conditions.100 There is no evidence that the, audits of Carboline i j demonstrate anything other than appropriate functioning of the quality I..j control program. Mr. Hamilton's gut feeling is not sufficient to induce us to inquire further. Another of Mr. Hamilton's allegations addressed undocumented removal of cable trays for which quality assurance documentation had been completed.101 Documentation problems have been noted b9 the NRC 6 Construction Appraisal Team (CAT) and were addressed in a hearing l! I 97 R. Hamilton Testimony, CASE Ex. 653, at 55., 98 H.at24-25. 99 id,.at47-48;Tr.3522. d j 100 R. Hamilton Testimony, CASE Ex. 653, at 54. t 101 . at 55. ~< w_ m._,

Initial Decision: 26 i subsequent to the parties' filing of proposed findings of fact. This allegation will be considered at the time the CAT findings are addressed. Other allegations made by Mr. Hamilton do not require that the j Board raise sua sponte issue's because they have not been sufficiently related to deficiencies in the quality assurance program; consequently, ( the Board does not believe that, if true, they would raise serious health or safety issues. These include allegations that the only quality control vault ceiling is fire-proof,102 that it takes some time to retrieve records from the quality assurance vault,103 that quality control inspectors did not observe paint being applied to buildings outside the containment,104 and that he was never instructed to take I greater care because the project was a nuclear plant.105 4. Allegations by Darlene K. Stiner and Henry A. Stiner Darlere K. Stiner and Henry A. Stiner appeared as a panel providing direct testimony for CASE. Mr. Stiner had worked at Comanche Peak as a 1 L ~ i! 102 Id* 103 Id. 104 l Id. at 59. 105 _Id. at 65. l h. c - ~ _ _,, - - - - -.

Initial Decision: 27 I welder fro.n November 1979 to December 1980 and from June to July 1981.106 Mrs. Stiner began working at Comanche Peak in August 1977 and was employed there at the time she testified.107 She had been employed as a welder and was a quality assurance inspector at the time she testified.108 In his direct testimony Mr. Stiner indicated that he had a criminal record.109 The applicant's counsel brought out on cross-examination that he had multiple convictions and had not revealed all of them on his second application for employment at Comanche Peak.110 He had not indicated any criminal record on his first application for employ-ment.111 The Board has considered this information in detennining the weight it gives Mr. Stiner's testimony. The Board notes that, almost without exception, Mr. Stiner's allegations are duplicated by other witnesses. l Also, the Board is not aware of any motive that Mr. Stiner would have to 106 CASE Ex. 666A. i 107 Testimony of Darlene K. Stiner Witness for Intervenor CASE ji (Citizens Association for Sound Energy), CASE Ex. 667, at 3 (Tr. 4127). 108 l . at 3,5 (Tr. 4127, 4129). 109 Testimony of Henry A. Stiner Witness for Intervenor CASE (Citizens !I' Association for Sound Energy), CASE Ex. 666, at 47-48 (Tr. 4249-50). i' 110 Tr. 4488-89; App. Ex. 146. 111 Tr. 4484, App. Ex. 145. l' ,i l Li ~.- + . ~.. -- m - y

Initial Decision:.28 I make allegations in this proceeding which he did not believe were true. If anything, the fact that Mr. Stiner's wife was employed at Comanche l Peak' at the time he testified would seem to provide him with a motive q not to make allegations against the applicant. i 1 i The one allegation which was made by Mr. Stiner and by no one else, wasthathewasterminatedbecause$eshowedaqualityassuranceinspec-tor a gouge in a pipe.112 This allegation was not elaborated on to any } q,5 significant extent by any of the parties. However, it is related to the question of management's comitment to quality control and, as such, it i will be considered in a later decision. j' Another concern raised by Mr. Stiner was that a hole was created in concrete in the Safeguards Building when he removed a partially in-l' I stalled Hilti bolt.113 Mrs. Stiner also observed the hole.114 Mr. Stiner alleged that repair of the hole was not subject to proper quality 'l!' assurance control.115 The staff has investigated this allegation and reported its findings in Investigation Report 81-12.116 The inves-tigation concluded t.ha t this allegation did not raise any safety I i 112 f H. Stiner Testimony, CASE Ex. 666, at 34, 40 (Tr. 4236, 4242). 113 _Id. at 25 (Tr. 4227). 114

0. Stiner Testimony, CASE Ex. 667, at 40 (Tr. 4161).

115 H. Stiner Testimony, CASE Ex. 666, at 25 (Tr. 4227). 'i 116 Staff Ex. 178. l t ~ m

... ~... Initial Decision: 29 ~ i / questions.117 Because we have no reascn to doubt the staff's evaluation 1 i of the significance of a hole in the wall of the Safeguards Building, we will'not declare this a sua sponte issue.118 'f Mr. Stiner also alleged that he was not told that greater care should be taken because the facility was a nuclear plant.II9 This j allegation relates to the overall adequacy of the quality assurance. program, but it is not by itself a cause for concern about the safety of .j the plant. I An allegation that was made by both Mr. and Mrs. Stiner was that ~ pieces of scrap iron were added to hangers or supports in the field.120 At least one of these allegations would appear to relate to Class 5 ("V") hangers,121 some of which may not be subject to the quality assur-ance system.122 As to a Class III hanger mentioned by Mr. Stiner, he said that the f~ heat number was added to a shim that had been made out of a scrap 1 t 117 Id. at 9.

I j

118 We are concerned that the staff did not provide a reasoned ~ explanation for its conclusion, but we do not see any reason to inquire further about this particular defect. 119 1 H. Stiner Testimony, CASE Ex. 666, at 32 (T*r. at 4234). 120 Id. at 41-42 (Tr. 4242-43); D. Stiner Testimony, CASE Ex. 667, at l TT-48 (Tr. 4171-72). 121 l H. Stiner Testimony, CASE Ex. 666, at 42 (Tr. 4244). 122 Tr. 4082, 4565. _ -,- m..._. o r- % _ m, - .m. m., -n .mn N&m- +-

Initial Decision: 30 I i l metal.123 According to rebuttal testimony by the applicant, use of new i j material is not required and traceability through heat numbers is only { required for structural members used in component supports.124 The shim t is not a structural member. As far as Mrs. Stiner's allegation that she j was directed to weld a piece of angle iron onto a small Class III i support, she admitted that she does not know what happened to the support once it was taken from her. She did not known whether it was _i installed or scrapped.125 None of these specific instances appears to be a problem. There-fare, the Board does not believe the Stiners' allegations in this regard raise serious health and safety questions about the plant which require i the Board to raise this as a sua sponte issue. k The Stiners made numerous allegations about welding practices at l Comanche Peak. They alleged that weave beading occurred although it was f prohibited on site.126 Weave beading involves welding using transverse oscilla~tions of the electrode.127 They further alleged that if weave I beading welding violations occurrad, the weld was improperly repaired. I s 123 H. Stiner Testimony, CASE Ex. 666, at 41-42 (Tr. 4242-43). J 124 Applicant cited ASME Section III or ANSI B-31.1 as the applicable code sections. See Tr. 4628-29. i 125 D. Stiner Testimony, CASE Ex. 667, at 47-48 (Tr. 4171-72). i 126 Id.. at 23-4, 28 (Tr. 4147-48, 4152). 127 Tr. 4086. u

j i Initial Decision: 31 i' i \\ i i i { Specifically, they alleged that repair of weave-beaded welds requires the grinding out and rewelding of the entire weld 128 but that the j practice of welders was to grind the surface of the weld which showed transverse oscillations and make only a surface welding pass.129 ,l The applicant presented rebuttal testimony that established that j only welding which involved significant transverse oscillation was prohibited as weave beading. Brown and Root, the firm performing i i construction at Comanche Peak, defines this to mean oscillations greater than four times the diameter of the weld rod ' used.130 Because some i I transverse oscillation is pennitted at Comanche Peak as being acceptable under Section IX of the code of the American Society of Mechanical Engineers (ASME), some of the welds which concerned the Stiners may have l* j;, been acceptable. The NRC did an investigation of the allegations of weave beading. A I i i visual inspection of an area in which this weave beading had allegedly occurred did not disclose weave welds.131 However, it would not have d. 120 D'. Stiner Testimony, CASE Ex. 667, at 26 (Tr. 4150). l 129 Id. at 25 (Tr. 4149); H. Stiner Testimony, CASE Ex. 666, at 9-10 ..l TTr. 4211-12); Tr. 4357. ~ j 130 Rebuttal Testimony of C. Thomas Brandt, Ronald G. Tolson, Gordon R. Purdy, Raymond J. Vurpillat and Randall D. Smith Regarding Quality l Assurance / Quality Control, App. Ex. 141, at 30 (Tr. 4685); Tr. 4412, 4420, 4635-6. 131 Staff Ex. 178 at 5. 4 e t'

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i, 3.-_.._. Initial Decision: 32 I done so if the welds had been reworked as described by the Stiners.132 i,l Conversations by the investigator with five welders revealed that three had 'seen weave welds but that those welds had been corrected.133 The applicant presented testimony that the repair method described - by the Stiners did not violate any procedures.134 The witness testified I that by the time the weld has been ground down, it no longer exceeds the 1 'I allowable diameter.135

However, we are unable to accept this explanation because we fail to understand the engineering principles involved here.

In particular, we do not understand the configuration of the joints in which weave beading occurred, where the grinding takes place or how the grinding cures the underlying weakness in' the joint due to excess transverse oscillation during welding. Hence, we consider i this to be an open item. The Stiness alleged that " plug welds" were used to fill improperly i-placed bolt holes and that this was not a permissible procedure.136 The applicant's witnesses testified that this procedure was permitted for it l}{ filling misdrilled holes at Comanche Peak and that this work requires a l; f ,1 t 132 Tr. 4599. 133 Staff Ex. 178 at 5. 134 Tr. 4650-51. l 135 ~Id. t j ? 136 D. Stiner Testimony, CASE Ex. 667, at 30 (Tr. 4134); H. Stiner Testimony, CASE Ex. 666, at 43-44 (Tr. 4219-20). -4 9 b--

fnitial Decision: 33 j i J f final visual inspection by quality assurance.137 An NRC investigation of this allegation found that plug welding occurred and that quality control inspectors were aware that they were required to inspect it.138 1 j Mr. Stiner, however, specifically stated that these welds are being made without quality assurance inspecting them.139 Neither the applicant nor the staff appear to have addressed the question of whether such welds 1i are being made and not being inspected. Nor have we been able to find the ASME code provisions that may allow this practice. 'Hence, the allegation is an open item. f Mr. Stiner alleged that downhill welding sometimes occurred, in .) violation of site procedures.140 The applicant's rebuttal panel t testified that although Brown and Root procedures may have prohibited j 'I ~i downhill welding at Comanche Peak, it is allowed by ASME Section 9 for root and cover pass and is allowed by the American Welding Society (AWS) to repair undercut.141 The record does not specify whether the Stiner allegations were limited to root and cover pass welds. Consequently, 1 1 e O l 137 Applicant's Rebuttal Panel Testimony, App. Ex. 141, at 36 (Tr. ) 4691). 138 Staff Ex. 178 at 6. 139 H. Stiner Testimoriy, CASE Ex. 666, at 19 (Tr. 4221). 140 Id. at 44-45 (Tr. 4246-47). 141 Tr. 4601-02. f N. -=

~.. _ _. _ ~ i I I the record explanation of the allegation is inccmplete and this is an 1 i open item. ' Both of the Stiners alleged that there were problems with the control of welding rods. Mrs. Stiner stated that, while these rods were supposedly controlled and accounted for through the use of stubs that assigned rods to a particular welder for a particular job, she had d: 3 discovered rods which had been abandoned or thrown out without the control system having identified the problem.142 She also stated that q there were instances when welders claimed to have used more welding rods than would be needed for a particular job.143 Thus, welders could have claimed to have used rods, kept them, and later used them on a weld for q which they had not been issued and might be inappropriate. Mr. Stiner testified that he had performed welding using rods which had been issued to other welders.14 1 The applicant, in rebuttal to the Stiners' allegation, presented testimony that weld rods were controlled to assure that safety-related weld rods of a proper type were used for a specific application.145 The panel stated that some NCRs had been written when these procedures were i t-1 142

0. Stiner Testimony, CASE Ex. 667, at 41 (Tr. 4165).

!I 143 .Ld. 144 H. Stiner Testimony, CASE Ex. 666, at 19 (Tr. 4221). 145 Ap licant's Rebuttal panel Testimony, App. Ex. 141, at 34 (Tr. 4689). o '~ .I ( ..=... + - - =----:.-- - ._.,n-

..~ ~ Initial Decision: 35 i

i

{ !} not followed.146 (Mrs. Stiner herself identified one such NCR that she had written.)147 ' This testimony is not sufficient to resolve the issue. The fact I that NCRs have been written o,n uncontrolled weld rods does not refute a charge that the control system for these rods, while present, is less i' than perfect. Neither the sta d ~nor the applicant has presented evidence that the system is so effective that we may conclude that i almost all (or all) of the breaches are corrected by the quality control system. This is an open item. There appears to be no way to clarify i the scope of this problem without a field investigation. It was also alleged that it was a common practice 'for welders to leave unplugged for prolonged periods the containers which were intended -- ( to keep weld rods heated.148 The purpose of keeping the rods heated is to prevent the welding rods from absorbing moisture.149 If the rods do }e.! absorb moisture, the moisture will escape as steam during the welding. j ', This will cause surface porosity.150 (Porosity is holes or voids made ji t. }! Ll 146 Id.. at 35 (Tr. 4690). 147 CASE Ex. 6675. b I40 D. Stiner Testimony, CASE Ex. 667, at 39 (Tr. 4163). 149 Tr. 4597. lt I 150 Tr. 4302, 4597. -_ m --_b --L m n-. , ~ % _ _ 3-w ^_;;.; -. n u

. ~. - 7'_...__....._. Initial Decision: 36 l I i in a weld by escaping steam.)151 Porosity is a visual inspection criterion for welds under the American Welding Soci.ety code.152 gf unheated rods cause welds with porosity, the weld must be inspected and j repaired.153 This will solve the problem caused by the unheated rod. Thus, this allegation involves construction practices rather than quality assurance and it involves practices which would not affect the 1 safe operation of the facility. The Board asked the applicant how welding in safety-related buildings would be verified prior to operation of the nuclear plant. In describing the welding verification process, the applicant stated that all Class I, II, III and V supports in safety-related areas will be examined on a case-by-case basis prior to turnover to the operations M group.154 The process includes looking for evidence in the in- ~~ spection record that there was a final visual inspection and other l[ inspections that were required to be performed on all pipe and equipment .1; l l l Ji ! 151 Tr. 4631. t 152 l j Tr. 4632. Porosity is not a criterion for inspections perfonned to Idd the ASME code. j 153 Applicant's Rebuttal Panel Testimony, App. Ex. 141, at 35 (Tr. ~ 4690); Tr. 4597. 154 Tr. 4646-48. l< i l l !) l

7-. 7 -- - Initial Decision: 37 t supports.155 The Board concludes that this voluntary action of the applicant is important to assure the integrity of thes.e supports.

i

'The Stiners also alleged that adequate' quality control was not) maintained over torquing of Hilti bolts. The quality assurance inspec-I v<

l tor is supposed to observe the torquing of Hilti bolts and to apply a

~ material known as Torque Seal aYter the proper torquing has occurred. (Inspection is not required for 100 percent of Hilti bolt torquing, but that is the goal of the program.) The Stiners alleged that quality control inspectors performing aG. .i ~ Hilti bolt inspection would not always stand where they could observe the actual torquing;156 that Torque Seal, whose handling sas supposed to . 1 /( n be controlled, was improperly in the hands of craft workers;157 and that ... ( Torque Seal had been found on Hilti bolts that had not been properly ( \\ torqued.158 The applicant's response in this area was that it was not

1li necessary for the inspector to observe the actual torquing of a Hilti,

if the inspector checks the torque wrench for a proper setting, k bolt i \\ \\ 1 j 155 Id. 3 156 D. Stiner Testimony, CASE Ex. 667, at 34 (Tr. 4158); H. Stiner Testimony, CASE Ex. 666, at 23 (Tr. 4225); Tr. 4299-4300. 157 ] D. Stiner Testimony, CASE Ex. 667, at.31 (Tr. 4155). 158 ,Id at 36 (Tr. 4160); CASE Ex. 667R. 1 1 ' gesp --.m_. _me

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~ ~ ~; ~~~ ~ ~ Initial 02 cision: 38 I r .f hears the click indicating that the bolt has been torqued, sees the' ,l craft person doing the torquing, and has no room to be in position to see 'the torquing indicator on the wrench.159 The applicant's panel \\ -x~ admitted that it was likely that the possession of Torque Seal was not i' ., / i ' I entirely limited to quality assurance personnel, in spite of attempts to control it.160 The applicant argued that this was not a cause for concern, however, because final verification of Hilti bolt torquing is dependent on a review of quality assurance inspection records.161 The Board finds that there is no problem with the nature of th quality assurance inspections performed on Hilti bolts.162 The Board is concerned, however, that Mrs. Stiner understood her instructions to be f! that she should assume that all Torque Seal had been applied by quality Y .{ assurance and that she should sign her inspections on that basis.163 gf t quality control inspectors signed inspections because they found Hilti i i A. 1l bolts covered with Torque Seal, the paper review of inspections would V N j not reveal the fact that quality assurance had not actually checked i ; i! c ] 159 Applicant's Rebuttal Panel Testimony, App. Ex.141, at 32-33 (Tr. 4687-88); Tr. 4537-9. q 160 Tr. 4534, 4536. a l 161 Applicant's Rebuttal Panel Testimony, App. Ex. 141, at 33 (Tr. 4688); Tr. 4541-42, 45-44.

p.

162 .] See also Staff Ex. 178 at 7-8. 163 i Tr. 4085. O --_y -n...~.- w, =_n_ m- ~, - .y

Initial Decision: 39 t i \\ 'j ( \\ it i whether the Hilti bolt had been torqued. Consequently, this is an open f.'.\\ There needs to be further evidence, based on field investigationy\\.\\ t item. e-1 r concerning whether quality control inspectors considered the presence of 1 Torque Seal to be so definitive that they did not check quality j { assurance records further. /' Mrs. Stiner alleged that an NCR which she had written on a burned bus box adjacent to the polar crane rail and resulting in gouges in the i polar crane rail was unfairly voided.164 The disposition of the NCR was that it was voided because its subject was not an item covere'd by the quality assurance program.165 There is no evidence that the bus box in question is safety-related, so there is no reason for us to declare a sua sponte issue. ) i Mrs. Stiner made several other allegations to which the staff and applicant have not responded. Some of them are reasonably specific and i I are open questions. These include her allegations.that Hanger

1. SW-1-102-106-Y33K is in a safety-related area and is severely mis-I matched;166 that a craft person was involved in performing quality

) assurance liquid penetrant inspections on the fuel pool liner;107 and i 164 D. Stiner Testimony, CASE Ex. 667, at 53 (Tr. 4177); Tr. 4073, 4102. 165 . at 54 (Tr. 4178). 100 H. at 45 (Tr. 4169). 167 M.at49(Tr.4173).

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i .7_,.__..__._._. Initial Decision: 40 ] \\ { j that ineffective action was taken when she identified numerous problems i on a hanger previously approved by quality assurance.168 Two allegations made by Mrs. Stiner may simply be dismissed without further consideration. She alleged that there is no traceability of materials until quality assurance becomes involved.169 Applicant argues I that traceability is only required for the quality assurance program and that it need not be applied to materials that are not covered by that program. In this posture, there is some ambiguity in the record, but we l conclude that there is only one logical explanation for that ambiguity. i Theoretically, it is possible. that items that had not previously been traced could be added to the quality assurance system, where they would - then become traceable. However, we exclude this inference because a rd testimony that items in the program are traceable means that their entire pedigree must be known. Those items could not previously have been untraceable. On this basis, we accept applicant's explanation.170 ,i - ij The second allegation is that material from a scrap bin could be I utilized on the site and would lack traceability.171 Since there is no t l i i 168 H.at56-57(Tr.4180-81). 169 . at 46 (Tr. 4170). 170 j* A party with information that our resolution of this issue is factually in error would have an obligation to correct the record, even if our finding were. favorable to its interests. t l 171 H.at57(Tr.4181). i w. ,..,-.:w___.:......

Initial Decision: 41 t 'l / .i .i allegation this has happened, it raises no issue about the quality of } actual construction at Comanche Peak. c l . Mrs. Stiner alleged that she was unqualified for a quality assur-4 ll ance position which she held.172 The CAT report discusses inspector ] qualification and this allegation will be considered in connection with l our decision on matters in the CAT report. Mrs. Stiner alleged that management at Comanche Peak had harassed her because she would be testifying for the intervenors in this proceed-3 ing.173 This is related to a matter considered in the CAT report and reflects on. management's commitment to its quality assurance program. 1 l It will be evaluated in a subsequent decision. it h ~7 5. Allegations by Charles A. Atchison Charles A. Atchison was employed by Brown & Root to work on the Comanche Peak site from February 27, 1979 to April 12, 1982.174 He

I CASE.

Among his many testified in these proceedings on behalf of allegations, Mr. Atchison claimed that he was improperly fired for l performing inspections.175 This allegation has been covered by a I f l I 172 H.at8(Tr.4132). 173 _Id. at 63-72 (Tr. 4187-96). 174 Testimony of Charles A. Atchison, Witness for Intervenor CASE, Case Ex. 650, at 5-7. 175 See,e.g.,jd..at53-54. l w., - _u

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_ ; _ ~.. _ Initial Decision: 42 1 l previous Memorandum and Order of this Board.176 That decision found l that Mr. Atchison was improperly fired. i ' The allegations not covered in either of our decisions include the f following:

1) problems with welding on Chicago Bridge and Iron pipe

[ whip restraints and moment restraints;177

2) problems with welding on NPSI pipe whip restraints;178 3) uncertified employees perfomed liquid penetrant testing;I79

4) unstated management direction to overlook f,

f' problems;180

5) and pressure to approve an audit of Tennessee Wall, Tube and Metal.101 These appear to be open issues.

Issue number 4 is j the subject of two ongoing investigations, which may also cover number i 5. Other Atchison allegations are vague, unrelated to the quality ij -- assurance program, or are speculative. These allegations, which are not treated in detail in this opinion, are that

1) the quality control f

t '.l 176 LBP-83-34, 18 NRC , July 6, 1983. 177 Atchison Testimony, Case Ex. 650, at 23-24, 40-41; Supplementary !j, Testimony of Charles A. Atchison Witness for Intervenor CASE ~ (Citizens Association for Sound Energy), CASE Ex. 656, at 2-3, 5-6. .j 178 Atchison Testimony, CASE Ex. 650, at 33. 179 Id_. at 51. 180 Id. at 58. 181 Atchison Supp. Testimony, CASE Ex. 656, at 2. I i i f l b ,__-..._h,___J. --m_ __.m

Initial Decision: 43 i l i I 1 vault may not be fireproof;182 2) there is low morale among workers;183

3) he was not instructed to use special care because he was working on a nuclesr facility;IO4 4) Japanese steel was being used on site;185 5) j pictures he had found in a desk on site showed a void at an unspecified f

location in Reactor Building 1;186

6) an individual, employed as a contractor's quality assurance manager, ordered for Brown and Root;187

~^ ~ 3 and 7) engineering permitted a type of welding by NPSI not authorized by procedures.108 Among Mr. Atchison's more substantial concerns is the allegation that there were problems with getting component modification cards to the document control center and incorporating them into appropriate document revisions.189 This allegation is related to matters discussed ~ --i in the CAT report and will be discussed later in that context. Mr. Atchison alleged that A490 bolts were being broken and that after tests were run to establish torque values for the bolts, the new i j l* i' I ,l 182 Atchison Testimony, CASE Ex. 650, at 34.

i 183

_Id. at 49, 64. 184 \\ Id. at 67. ~ 185 Atchison Supp. Testimony, CASE Ex. 656, at 1-2. 186 Id. at 7. 4 187

i

_Id. at 9. i 188 Id. at B. \\' 189 ' i. Atchison Testimony, CASE Ex. 650, at 35-36. ' l. ~: e

v -

. _. ~ Initial Decision: 44 I -l torq'ue values were not incorporated into site procedures.190 Neither .l the applicant nor the staff has responded to the allegation. It is an open item. Another allegation made by Mr. Atchison was that he had observed a welder " quenching" a weld directly, in violation of site procedures.191 4 It appears from Mr. Atchison's testimony that he wrote an NCR on this matter and that the project engineer determined that, while the j quenching violated site procedures, it did not affect the qua'lity of the weld.192

However, we have no understanding of the reason for prohibiting the quenching of welds or why this particular weld was found

.l-l to be acceptable. Mr. Atchison alleged that a flammable lubricant was used to assist in pulling cable through electrical conduit.193 The lubricant was tested in a laboratory and found to be satisfactory. Without having been present for the laboratory test, Mr. Atchison questioned whether it reasonably approximated conditions in the field.194 In essence, Mr. Atchison gave no reason for questioning the accuracy of the results of

I I

190 Id. at 29-31. f ,I_d at 50-51. 191 j4 192 Id. at 50. I 193 Id. at 55. 194 _I,,d_. a t 55 -56. l 0 ~. = - - ~.:r k. --.n;__ s

x___,

Initial Dacision: 45 I the laboratory tests. We find nothing in this allegation which we i should pursue sua sconte. ' r. Atchison also alleged that there were too few quality control M inspectors to perform the quality assurance work at Comanche Peak.195 Applicant provided rebuttal testimony indicating that during the time 1 Mr. Atchison was employed at Coma he Peak, the ratio of quality control 'l i inspectors to craft personnel was within the average for the indus-l} try.196 In addition, the applicant's witnesses testified that if the ratio had been less favorable it could have influenced how rapidly l1 required inspections were performed, but would not have affected whether 7 they were performed. Accepting the applicant's testimony that the inspections will be performed regardless of the number of inspectors, l --( Mr. Atchison's allegation does not by itself raise an important issue 2 l about the number of inspectors. We do not. decide whether the parallel irsue raised by the CAT inspectors is meritorious. Mr. Atchison alleged that hundreds of flange bolt-up joints had not

{

been submitted to quality assurance for final inspection. Thus, he alleged the start-up group would repeatedly disassemble and reassemble t }' 195 Id. at 57. 196 Applicant's Rebuttal Panel Testimony, App. Ex. 141, at 38-39 (Tr. 4693-94). 197 J_d. - - ~ ,n

Initial Decision: 46 4 i'I l the joints.198 Certainly, if the units are to be disassembled, this should be done before the final quality assurance approval. It does not appear that Mri Atchison is alleging that these joints will never be j submitted for final quality assurance approval. They appear to be just f' one additional item left for inspection during the final walk-down at t the end of the project.

j Mr. Atchison alleged that he had observed the " cold springing" of two lines from reactor coolant pump compartment number three.199 7n i

rebuttal, the applicant indicated that the cold sprung pipe was part of the component cooling water system, that an NCR had been written on it, ff and that repair work had been required.200 Mr. Atchison was reasonably specific about the lines he. alleged had been cold sprung. The applicant f did not indicate how they determined that the lines to which he referred ij were not part of the reactor cooling system. They may well be correct l ll in their conclusion; however, there is an important gap in our record l that needs to be filled. / Mr. Atchison's final allegation was that minimum wall thickness violations had occurred in piping.201 He testified that an NCR had been {q 4 1. j-198 Atchison Testimony, CASE Ex. 650, at 62. I; 199 Id. at 63. 200 Applicant's Rebuttal Panel Testimony at 36-37 (Tr. 4691-92). l 201 Atchison Testimony, CASE Ex. 650, at 63. g n wn ~ - -, - g - ym,-. -.-, y ,y. ~

~ Initial Oscision: 47 ( { written on this matter and had led to two backfit programs.202 As far l as he knew the NCR had not been closed.203 Since an NCR had been j wri.tten on the problem and there are controls requiring that there be an appropriate disposition, we find that this allegation demonstrates tne 3 l correct working of the quality assurance program and does not present an l I allegation that we should pursue sua sponte. .t. 4 6. Miscellaneous Allegations' a) Lobbin Report Mr. B. R. Clements, Vice President, Nuc. lear of TUGC0 comissioned a management study by Mr. Frederick B. Lobbin, to review the effectiveness l of management controls within the quality assurance organization.204 This review was entirely voluntary on TUGCO's part.205 It was a quick-and-dirty effort to identify problems that Mr. Clements might ll llC ii i ,j 202 Id. at 63-64. cl i 203 ~ ]j Id. at 64. 204 See App. Ex. 48; Testimony of B. R. Clements Regarding Reviews of Management Control Program and Activities of Texas Utilities Company Quality Assurance Organization, App. Ex. 118, at 2-3; Testimony of Frederick B. Lobbin Regarding Review of Management Control Program and Activities of Texas Utilities Generating Company QA Organization, App. Ex.119, at 2; Clements, Tr. 2156; Lobbin Tr. 2163-64. 205 B. R. Clements Testimony, App. Ex. 118, at 3; F. Lobbin Testimony { App. Ex. 119, at 2. See also 10 CFR Part 50, Appendix B which does not require any such management study. i y .y 41 J_-

.~ -~ Initial Decision: 48 I ~! follow up if he thought additional effort was warranted. Clements at

1 3-5.

Mr. Lobbin testified that he sometimes overstated his conclusions .j in order to assure that they would be attended to.206 Despite this method of exaggeration, each of his findings was evaluated by applicant in a response document.207 ' [. We conclude that as a result of the nature of Mr. Lobbin's study, his individual findings are entitled to little weight in this .1' ,} j proceeding. This conclusion does not prejudice the right of a party to use his findings as cumulative evidence, together with other direct - t-f. evidence, of positive or negative findings concerning the quality 1 assurance program. t , 7 b) The Number of NCRs CASE introduced a large number of documents that demonstrated the existence of construction deficiencies at Comanche Peak.208

However, there were no witnesses that testified that the number of deficiencies was abnormal.

Indeed, the staff's resident inspector, Mr. Taylor, .l testified without contradiction that the number of NCRs indicates only f l 206 Lobbin, Tr. 2170. 207 App. Ex. 49. 208 See, e.g., CASE Ex. 305-570. g. ~ ..-..-e.,--.~.. L4 . QF h gy M'" *

• Initial Dacision: 49

( that the quality assurance program is working.209 In addition, the j staff examined a sample of NCRs to detect trends indicative of problems, f j and they have concluded that there are no serious problems revealed by the logs of NCRs.210 Staff examination of corrective actions taken pursuant to NCRs also resulted in a positive evaluation.211 We find no evidence that the numEer of NCRs and of other deficiency reports was in any way excessive for a project of this size. To the contrary, the existence of these reports is consistent with the Comission's quality assurance requirements. ( i s 8 ,j~ l 4 ,t 209 Taylor, Tr. 1712, 1730-31. 210 Stewart, Tr. 1282, 1285; Crossman, Tr. 3021. 211 Crossman, Tr. 3022. l I .4 e *, . -~ 4 ;,~ w: 3 ~3

Initial Decision: 50 [ II. Contention 22 Contention 22 states: Applicants have failed to comoly with 10 CFR part 50, Appendix E, regarding emergency planning for the following reasons: (a) The FSAR does not identify state or regional authorities responsible for emergency planning or who have special i qualifications for dealing with emergencies. (b) No agreements have been reached with local and state ( officials and agencies for the early warning and evac-J uation of the public, including the identification of the 1; principal officials by titles and agencies. ? (c) There is no description of the arrangements for services of physicians and other medical personnel qualified to handle radiation emergencies and arrangements for the transportation of injured or contaminated individuals beyond the site boundary. (d) There are no adequate plans for testing by periodic drills of emergency plans and provisions for participa-4. tion in the drills by persons whose assistance may be l,n needed, other than employees of the Applicant. ii: (e) There is no provision for medical facilities in the

,j immediate vicinity of the site, which includes Glen Rose.

ti L; ' (f) Thereisnoprovisionforemergencyplang{ggforGlen i. Rose or the Dallas / Fort Worth metroplex. !J CASE did not address this contention at all in its proposed find-I i l! ings of fact. In recent hearings CASE has had very few questions for i i ]l 212 Order Subsequent to the prehearing Conference of April 30, 1980, slip op. at 11 (June 16, 1980) (unpublished). i l '4 wh-' m- --c-m .~. m m . -, x v~e. _ . _.. e - -

... _ 2 u._ Initial Decision: 51 ( witnesses on emerger.cy planning.213 CASE's failure to file proposed findings on the emergency planning contention when directed to do so constitutes abandonment of the contention.214 CASE subsequently has failed to pursue the contention vigorously, confirming the wisdom of [ declaring this to be an abandonment. l, The development of emergency plans is an evolutionary process. In i May 1983, the staff introduced into the record an interim finding by the M Federal' Emergency Management Agency (FEMA) that there is reasonable assurance that the off-site protection of public health and safety is -{ adequate.215 The documentation attached to this interim finding makes it apparent, however, that the state and county emergency plans do have I deficiencies. 'l lhd The interim finding is based upon a review of the paper plans.216 The ffnding is in the nature of a progress report.217 It indicates that ~

.i if all the commitments made in the emergency plans are carried out, iil il

!j l q 'a 213 Tr. 7286, 7480-81. 214 10 C.F.R. I 2.754(b). 215 Memorandum from Lee M. Thcmas to William Dircks (September 29, p. 1982), ff. Tr. 7414. j 216 Tr. 7417-18. 217 Tr. 7456. w"- - r-,,_---."--- c Alwo-:- -~_ y.- z w w -n ~** S- - A-......,,m

Initial 02 cision: 52 ( there is reasonable assurance that the plans will provide adequate protection for the public.218 l At this stage, it is too early to determine whether all those commitments will be fulfilled. The Board is not satisfied that the i plans as presently constituted are adequate. It remains concerned about j these promises. Since the evolutionary process is not yet complete (there must, for example, be a drill or exercise),219 the Board does not believe it should raise any issues sua sponte at this tire.

However, the Board will continue to observe the development of the emergency plans and may raise issues sua sponte later if the commitments are not met or the deficiencies are not rectified.

Our Order of June 27, 1983, y l elaborates further about the extent of our concern about this issue. -d 1 III. Board Questions In addition to the contentions, the Board had posed four questions. .j ! These questions were posed for the purpose of obtaining information so f t i ll ' that the Board cou'Id determine whether a serious health or safety issue 't l existed which the Board should raise sua sponte. This decision resolves Ll i those aspects of these Board questions that were raised during hearings j; occurring prior to March 1983. .l 218 Tr. 7452-4. 219 Tr. 7416-17, 7441-3, 7481. ~~ 4, c.---- _ w --,

i _ _ _.s Initial Decision: 53 l 1 ( A. Board Question 1 In Board Question 1 the applicant and staff were asked to "[d]e-scribe in detail the planned method for handling any hydrogen gas in the CPSES containment structure."220 The potential source of hydrogen gas ] in the containment structure would be a loss-of-Coolant Accident (LOCA). Combustible gases, principally hydrogen, would be generated inside the .j containment during a LOCA by: a zirconium-water reaction, release of /! free hydrogen from the primary coolant system, radiolysis of water, or corrosion of susceptible construction materials in containment.221 As it has previously indicated on the record,222 the Board is satisfied that any hydrogen generated within the containment structure can be satisfactorily handled. The Board relies on the large centainment j structure at Comanche Peak, the redundancy of electrical recombiners I j provided, and the requirement by the staff that the recomendations for l! operator training found in the THI-2 Short Tenn Lessons Learned report ~ will be implemented prior to issuance of the operating licenses.223 l, I i 220 Order Subsequent to the Prehearing Conference of April 30, 1980, j slip op at 4 (June 16, 1980) (unpublished). j 221 Final Safety Analy' sis Report (FSAR) il 6.2.5 & 6.2.5A. f Tr. 693, 731. 222 223 See Board Ex. 1; Tr. 730-31. i i ~.m vm - < -- - ~ 7-, % m - _ _ _ _ _ _ _ _ _ _ _ _ _

c. Initial Decision: 54 y ~ ( B. Board Question 2. Board Question 2 states: Applicant and staff should describe in detail the operating quality assurance program for CpSES. A description of the provisions for conduct of quality control audits should be provided, including a description of how reagr operations and reactor operator training will be audited, t The applicant and the staff provided the Board with extensive informa-i tion on the structure and purpose of the quality assurance program for operations at Comanche Peak.225 The Board is convinced that if the operational quality control program is instituted as described, it will function adequately. The Board notes that while specific implementing ~ procedures were not provided to the Board,,the staff will review them l, before it will issue the ' licenses.226 The staff will also audit implementation of the operational quality. assurance program.227 In light of the comitments made by the applicant and the staff, the Board l' i! 224 Order Subsequent to the Prehearing Conference of April 30, 1980, -l { slip op. at 5 (unpublished). 225 Testimony of B. R. Clements Regarding Management Commitment to Quality Assurance, App. Ex. 8; Testimony of David N. Chapman Regarding the Operating Quality Assurance program for Comanche Peak, App. Ex. 9; Testimony of R. A. Jones Regarding Commitment of On-Site Management to Quality Assurance, App. Ex.10; Testimony of 1 Antonio Vega Regarding Provisions for Con ~ duct of QA Audits and j' Reactor Operator Training, App. Ex.12; staff Testimony of John G. 1 Spraul Regarding Operating Quality Assurance (Board Questien No.

{

2), Staff Ex. 5; App. Ex.11; Tr. 506-662. 226 Tr. 657-58, 662. l

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227 Tr. 656. l I ~- L- .a,.

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Initial Decision: 55 (

{

is satisfied at this time that the Board need not pursue this matter further by raising it as a separate, sua sconte issue.228 l! C. Board Question 3 Board Question 3 asked the applicant and the staff to describe the status of the resolution of Safety Issue TAP-9 (Anticipated Transient 1 J Without Scram or ATWS) as it relates to Comanche, Peak. The staff answered the Board question with three affidavits.229 The staff noted that the Consnission has issued a notice of ruiemaking on ATWS.230 Prior s !l to operation of Comanche Peak, the applicant will be required to develop

3 j

emergency procedures and to train its operators to recognize and cope with an ATWS event. The staff indicated that the scram systems were '"( redundant and highly reliable and that in view of. favorable operating experience, Comanche Peak could be operated without undue risk during 3 .i the period pending the final ATWS rule. J.

j With respect to the favorable experience, Mr. Pyatt stated that, i

l "There have been roughly one thousand reactor years of experience l.

accumulated in foreign and domestic coninercial light-water-cooled 228 This conclusion would not prevent us from renewing our concern about operations quality assurance should we ascertain that there have been substantial deficiencies in aspects of the construction quality assurance program that also are present during operations. 229 l Affidavits by David W. Pyatt, James W. Clifford and Marvin W. !j Hodges, dated May 5, 1982. (Bd. Ex. 3). 230 46 Fed. Reg. 57521 (1981). i a J_ ~

) Initial Decision: 56 j ( reactors without an ATWS accident." He chose his words carefully to avoid having to mention that there had been at least one ATWS event. . Although the Board had considered the staff's response on ATWS to be satisfactory,231 we note that on February 25, 1983, a potentially serious ATWS event occurred at Salem Unit 1 as a result of a failure of l redundant reactor trip breakers.232 After that event, we asked whether the reactor trip breakers at Comanche Peak were similar to the Salem breakers and whether there would be any new requirements prior to operation. The staff has infonned us that a task force has been formed to look into the generic implications of the Salem event and that final 3 l actions in response to the event are still under consideration. Consequently, we are not yet satisfied concerning the need to declare L ATWS to be a sua sponte issue. ,i t D. Baron Injection Tank l When the Emergency Core Cooling System at Comanche Peak is 233 activitated, high-head pumps take borated water from the refueling water storage tank and inject that water into the reactor cooling system. The original design called for the insertion of a Baron i I j 231 Tr. 693. l 232 Memorandum from Darrell G. Eisenhut to Chairman Palladino, et al. 1 (March 3,1983) (Board Notification 83 Failure of ReactoTT7Tp l Breakers to Open in Trip Signal). 233 Pumps designed to inject into the primary cooling system when it is fully pressurized. i - ; ___ G m==wo:e - c e~~ _-3~3; _ -, ~. ~.. m m - u~ __-w-

_. ~. _. Initial Decision: 57 I ( Injection Tank (BIT) between the high-head pumps and the reactor cooling system. The Applicant proposes to omit the BIT.234 The Board inquired int.o the appropriateness of the deletion.235 The concentration of Baron in the refueling water storage tank is 0.2%; the concentration of baron in the BIT would have been 12%. Such a f high concentration of baron would require that the tank and all lines i i and valves of the BIT be kept at high temperature to prevent the precipitation of baron crystals in the BIT and the consequent plugging i of valves and lines which connect the high pressure injection system i j (HPIS) to the primary coolant system, a potential hazard to the operation of the ECCS system dur,ing a transient. Representatives from Westinghouse have informed the staff that the ~~ BIT was included in the original design for the sole purpose of mitigating the consequences of a steam-line break accident. They have l made an analysis of a worst case scenario, a large steam-line break when ) the reactor is just critical, at zero power ed at operating temperature. In this scenario, the secondary system would rapidly !lI depressurize causing rapid cooling of the primary system, an increase in reactivity above critical (due to more optimal moderation at the reduced 'I i 234 The description of the BIT system and its effectiveness in reducing I the transient following a steam-line break is taken chiefly from the affidavit of staff witness Samy Diab (following Tr. 781) and i the attached "Sumary of Meeting on Comanche Peak Design Change and Responses to RSB Questions." 235 Order of April 2,1982 at 2-3. 1 _.~.n n - - m - ~~ w, ~, . Q ', ' a-% ,, _ _ _ ~ _ _ S n

Initial Decision: 58 ~ I temperature) and a return to power production in the core. The reduction in primary coolant temperature and pressure would trip the saf,ety infection signal and initiate the pumping of borated water into. 236 the core. If there is a BIT, the power would peak at about 15% of i. full power and then gradually decrease as the boron reduces the reactivity. Without the BIT the power would peak at about 20% of full l, power and persist somewhat longer.237 [. Applicant, with support from the staff, has argued that this ie increase in power without the BIT is not significant. It relies on }.;. Westinghouse calculations that show that the DNBR (departure from i nucleate boiling ratio) would remain above 2.5, indicating a wide margin of safety before the coolant would reach a " film" condition which would -k interfere with the safe removal of heat from the core. Although applicant and staff support deletion of the BIT in order to reduce the risk that baron crystals might interfere with ECCS l, operation, neither identified an instance where the ECCS had been compromised due to precipitation of baron.238 The witnesses stated, l i 11 I il l 236 The main steam-line break incident is analy:ed using a conservative j assumption that the control rods for the most reactive section of the core do not insert. Most of the power'that is generated comes from this one section. 237 Figures 3-4 and 3-5 in Attachment I to the Testimony of S. Diab, following Tr. 781. 238 Tr. 746; Tr. 783. l l m .~

Initial Decision: 59 i (' generally, that there have been operational problems with the BIT.239 i

However, the Board is independently aware that there have been f

boron-crystallization events of sufficient seriousness to be called. precursors to potentially severe core damage accidents.240 We agree with applicant and staff that on balance the Comanche peak Station apparently would be safe without the BIT. However, we are concerned that the NRC staff has relied entirely on the Westinghouse analyses.241 We recognize that the matter also has been brought to the 1: I attention of the Advisory Comittee or Reactor Safeguards (ACRS); 'I however, the staff's reliance on the Westinghouse analyses was not 242 brought out and is a matter of concern to us. We request the staff to bring this matter to the attention of the. ACRS once again, clearly j ~ '(N indicating that the staff relies entirely on Westinghouse analyses of n DN8R. I i 'l, s, !l 239 Tr. 778 and 782, staff; Tr. 746, applicant. 240 ) Science Applications Incorporated and Oak Ridge National i Laboratory, NUREG CR-2497, Precursors to potential Severe Core j-Damage Accidents: 1969-1979 (A Status Report) (June 1982) at j Appendix C, pp. C,9 to C-10. { 241 Tr. 782. 242 Advisory Comittee on Reactor Safeguards, 259th Annual Meeting (November 13, 1981), Tr. 360-361, 9. See also staff presentations i of November 11 and 13, found appended to the ACRS transcript. e' b --~~-.+s .. w.sa i w - 1---1w. . ng e_ _ - ..y_.--

p.. Initial Decision: 60 I j 0RDER j For all the foregoing reasons and based on consideration of the entire record in this matter, it is this 29th day of July 1983 i I ORDERED: ? 1. This is a proposed decision. 2. Pursuant to the Board's authority to require the filing of i; Findings of Fact, objections to this decision are waived unless they are l filed in compliance with the format requirements prescribed near the >6 beginning of the accompanying memorandum. 3. Objections to this decision must be received within 22 days of I { ~\\- issuance of this Order. 4. Replies to objections must meet the same specificity l ll l requirements applicable to the objections themselves. In particular, j' they must clearly state what they are replying to and provide a i reasoned, documented discussion that responds directly. l 5. Replies.must be filed within ten days of receipt of the 1 l objection being replied to. I i ~ I 1 i l ) i (} ( u n-- .,. ~ _ - n

3 Initial Decision: 61 '.Y f. N( FOR THE ATOMIC SAFETY AND LICENSING BOARD i b1h ' j Peter 8. Bloch, Chainnan ADMINISTRATIVE JUDGE Walter H. Jor g t I ADMINISTRATIVE JUDGE l Kenneth A. McCollom ADMINISTRATIVE JUDGE Bethesda, Maryland ,i " i i j 1 J .i l i t l f 4 g . ~ ' - M mmme e P 6.]

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. ~. _ T (- l j < t c.. t,. 1 p p APR 161962 I I MEMORANDUM FOR: Roger Fortuna, Chief. Investigation Branch, EI Office of Inspection and Enforcement HQ FRON: John T. Collins, Regional Administrator, RIV

SUBJECT:

ALLEGED COVERUP OF WELD DEFECTS AT CPSES (Dockets: 50-445/81-12; 50-446/81-12) 1 In accordance with memoranda from V. Stallo and D. Thompson, dated April 21, 1980, and July 15,1980, respectively, reganting reports of investigation, i a copy of the subject investigation report is being forwarded to you for your information. Enclosed in a sealed envelope attached to the investigation report is j supporting docismentation concerning this matter. crista 1 518 D ,J.ct: L Collins John T. Collins Regional Administrator i

Enclosures:

I As stated 1 1 J i i V C) h esswC F01A-85-59 ....E .0/EIS. ... rap. RIV..... " " > D .1 - ic..R - ..EJohns .j...k.11. ins..... om>i....4./ /82........ 4../..d.../. 8 2.... . 4../..U /. 82........ 4../.l.y. 82 ........'..8....

1 hl U In Reply Refer To: Dockets: 50-445/81-12 50-446/81-12 M i E !552 i Texas Utilities Generating C==ay ATTN: Mr. R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen: This refers to the investigation conducted by Messrs. D. D. Driskill and R. K. Herr of our staff during the period August 7, October 8-24, and November 23-24, 1981, of activities authorized by NRC Construction Permits CPPR-126 and CPPR-127 for the Comanche Peak, Units 1 and 2. i Areas examined during the investigation and our findings are discussed in the enclosed investigation report. Within these areas, the investigation consisted of selective exasrination of procedures and representative records, interviews with personnel, and observations by the investigators. l-Within the scope of this investigation, we found no instance where you failed to meet NRC requirements. l In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Doctaient Room. We understand that you will conduct an expedited review of this investigation and will notify us by telephone when we will be able to release this report.to the NRC Public i l Document Room and the ASLB. e i lf r th a (10 C Q a 90 yv um s'*:t>- El. #1.....EIf..,,,.,,j.. 0/.EI.S....jd.,, RPB1 f,v ', ORR$ EP., y,7, RIV 'x j -*> 000.t sk.i..1.d c. O.......EJohnson ..G.LMads en. JGagliardo. h 1 i ns..,, 1 4 4/ll/82 4/@/82 ....../..d 82....,. 4 /.*... /. 8 2 ... /l6.../ 82

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Texas Utilities Generating Company ' i Should you have any questions, concerming this investigation, we will be j pleased to discuss them with you. Sincerely, G. L. Madsen, Chief l . Reactor Project Branch 1

Enclosure:

bec to DMB (IE01): Appendix - NRC Investigation Report BC 50-446/81-12; 50-446/81-12 pM AE00 cc: ELD Texas Utilities Generating Company IE FILE ATTH: Mr. R. C. Schmidt IE/DRRRI/RPRIB Project Manager NRR/DHFS i 2001 Bryan Tower NRR/DHFS/0LB } Dallas, Texas 75201 NRR/0LB/RAB i RES LDPR POR:HQ NSIC NTIS .j bec distrib. by RIV: R. TAYLOR, RRI j D. KELLEY, RRI T. WESTERMAN, RPS-A INFO SYSTEMS a t ; RPB2 TPB J. T. COLLINS, RA E. H. JOHNSON, 0/IES C. WISNER, PA0 TEXAS STATE DEPT. HEALTH RESOURCES Juanita Ellis David Preister Richard Fouke t 1 i iI f. .. n ..?s u t_ m, mm _ y,a* - ,c ;ef

~ t a .a U. S. NUCLEAR REGULATORY COMMISSION REGION IV Investigation No. 50-445/81-12 50-446/81-12 Docket Nos. 50-445; 50-446 li Licensee: Texas Utilities Generating Company Facility: Comanche Peak, Units 1 and 2 Investigation at: Glen Rose, Sommerve11 County, Texas Investigation conducted: August 7, October 8-24, and November 23-24, 1981 i Investigators: L> o.O d f-D 0. ). Driskill, Investigator, Investigation anc Date Enforcement Staff Ja [ d d C c.V .R. K. Herr, Investigat37, Investigation and Date 1 Enforcement Staff l8L Reviewed by: =c E., H. Johnson, Director, Investigation anc Date Enforcement Staff d 4 l'/i - l Approved by: L J. E. Gagliardo, Director, Division of Resident, Date \\Reactor Project and Engineering Program i t i Ilt i l1 l' 7ptess g > Il a ,...U.l . _1

i- ~. 0 2 Summary: l Investigation conducted on August 7, October 8-24, and November 23-24, 1981 (Report No. 50-445/81-12; 50-446/81-12) Areas Investicated: Individual A alleged that improper welding techniques have been utilized on some Ripe supp.ons.;_. holes drilled,.i.n..vatiguyomppone ts_3re i.llegally_ plugged To_me hilti bolt QC inspector 5'do_not conduct tEeir inspections in complian~ce with procedures; there is a lack of control on a substance called Torque Seal (used_ to e.nsure integetty of,,insp_ected_hilti boltsl;'and a hole in a concrete floor was { m } not reported to QC, as required, prior to' repair. This investigation involved j. 48 hours by two NRC investigators and one NRC inspector. Results: Investigation of the above identified allegations disclosed no evidence support-ing the claims of Individual A. Numerous interviews of Comanche Peak Steam Electric Station (CPSES) construction personnel, Quality Control Inspectors and Craft Supervisors were conducted, in addition to inspections of areas identified by Individual A which allegedly exemplified his claims. I i t i ~f f

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1. L INTRODUCTION Comanche Peak Steam Electric Station, Units 1 and 2, are under construction in Sommerve11 County, Texas, near the town of Glen Rose, Texas.

Texas Utilites Generating Company (TUGCO) is the construction permit holder with Brown and Root, Inc. (B&R), as the constructor and Gibbs and Hill, Inc. (G&H), as the architect / engineer. REASON FOR INVESTIGATION i On July 21, 1981, Individual A telephonically contacted the NRC Region IV Duty Officer and alleged that defects existed in pipe welds within the CPSES Auxiliary Building. Individual A provided no additional information and agreed to be personally interviewed at a later date. SUM 4ARY OF FACTS On August 7, 1981, Individual A was interviewed regarding his concerns related to construction activities at CPSES. During this interview, Individual A 'l identified the following matters as having potential, adverse safety signifi-cance: I 1. Numerous pipe supports were fabricated utilizing weave welds which are prohibited by procedure. 1 2. Holes are drilled in various types of pipe supports, cable tray supports, and plates, and when, on occasion, they are found to have been drilled in 2 'h the incorrect location the holes are filled, utilizing illegal plug welds. \\ [There is a lack of control on a product called Torque Seal which is used o 3. d hilti bolts, once they have been inspected and torqued. This lack of t l control could lead to a questionable integrity of bolts which are marked with this substance. 4 Some hilti bolt QC inspectors do not properly ensure that bolts are correctly installed and torqued prior to documenting their satisfactory installation. j 5. In November 1980, the removal of a hilti bolt from the floor in the Safeguards Building resulted in a cone-shaped section of concrete being removed, which extended through the floor. A general foreman and foreman [ had this hole repaired without properly notifying the QC Department. l !!l, 6. Welders are not keeping their rod cans plugged in during the work day. Ii (During a subsequent interview of Individual A, he withorew this allega-t tion, stating the problem had been corrected.) l l M' >hm T* -,$^**

[, .4 1. Persons Contacted Licensee Emolayees "R. G. Tolson,. Site QA Supervisor, TUGC0

• J'. T. Merritt, Manager, E&C, TUCI Other Persons Contacted Individual A through L
• Denotes those attending exit interview.

~ Investication of A11ecation 2. Allecation No. 1 1 Numerous pipe supports were fabricated utilizing weave welds which are prohibited by procedure. Investicative Findinas l On September 8, 1981, Individual A was interviewed concerning his safety-related concerns at CPSES. Individual A stated that during his employment I at CPSES he became personally aware that numerous pipe hangers were fabri-l cated utilizing weave welds. He stated that according to his instructions i weave welding was not to be used on any jobs at the site. He additionally stated that weave welding violates written procedures at the site, as he knew them. Individual A stated that numerous examples of weave welds on pipe supports could be found in the South Yard Tunnel, unless the surfaces of the welds were ground off and the welds were capped. Individual A identified three welders whom he believed would corroborate this allega-tion. Individual A went on to state that " literally thousands of weaved 1 welds exist at various locations on the site." He stated that many of .i these have had the top surface of the weld ground off and the weld capped L with the required stringer bead. Individual A identified five additional welders whom he stated were aware of this practice. l l-l il h J li i l l M &gm y V -.g~- WS M*'W- HQ*

s .5 Interview of B&R Emoloyees i' On September 8, 1981, Indiviudal B, a Brown & Root Mechanical Quality Control inspector at CPSES, was interviewed regarding the use of weave i welds. Indiviudual B stated that weave welds are used in some instances. Individual B recalled one occasion when a weave weld was noted during an inspection, but was subsequently ground out and correctly rewelded when the appropriate supervisors were notified. On September 16, 1981, Individuals C, D, E, F, and G, all B&R structural welders at CPSES, were interviewed regarding the use of weave welds. All agreed that weave welds were not authorized for use at CPSES. Individuals C, 0, and E stated that they had observed inexperienced welders use weave welds, but that these were always ground out and rewelded when identified. Individuals F and G stated they had never observed weave welds used'at i CPSES. None of these individuals were aware of the existence of any components containing weave welds which were not properly repaired. Examination of South Yard Tunnel Hancer Welds On September 10, 1981, a visual examination of the numerous pipe supports (hangers) located in the CPSES, Unit 1, South Yard Tunnel, was conducted by Mr. Robert Taylor, Resident Reactor Inspector, NRC, and reporting investi-4 1 gator. It was noted that all of the welds on these pipe supports were the approved stringer bead type welds and no evidence of griding was i apparent. Some welds did exhibit evidence of grinding, but had weld l characteristics which made it possible to identify the type weld used as l stringer beads. No evidence was found to support Individual A's contention that weave welds were used in the South Yard Tunnel. 1 A11ecation No. 2 l Holes are drilled in various types of pipe supports, cable tray supports, i and plates, and when, on occasion, they are found to have been drilled in j the incorrect location the holes are filled, utilizing an illegal plug weld.

l Investicative Findings N

On September 8, 1981, Individual A was interviewed and stated that holes l are drilled in pipe supports and cable tray supports to facilitate i bolting them in place. He stated that on occasions when it is determined it that the hole was drilled in the incorrect location, it is filled in utilizing a plug weld. He stated that another hole is then drilled in the proper location on the component. Individual A stated he had the }} impression, based on comments made by various supervisors, that plug j! welding was not an accepted practice. I G -h. % '.6 m -W*-

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.6 f Interview of B&R Welders On September 16, 1981, Individuals C, D, and E were interviewed regarding th'e use of plug welds. Each stated they had repaired holes in various components utilizing plug welds on numerous occasions in the past. Individuals F and G stated they had never been involved with plug welding, and both stated they understood that it was not an authorized practice. ~ Interview of B&R Oc Insoectors t On September 15, 1981, Individuals H, I, and J were interviewed and stated they have observed on numerous occasions holes which were filled, utilizing the plug welding technique. Each stated a QC inspector is required to inspect the plug welding of holes (which is authorized by procedure) to ? determine if it is done properly and does not violate hilti bolt spacing j requirements. A11ecation No. 3 There is a lack of control on a product called Torque Seal which is used on hilti bolts once they have been torqued and inspected. This lack of i control could lead to the questionable integrity of bolts which are marked I with this substance. I Investicative Findinos ! n September 8, 1981, Individual A stated he was concerned about the O f ,,*i/, lack of control of the product Torque Seal, which he stated was placed on a hilti bolt after it had been torqued and verified by a QC inspector. b '/, He stated QC personnel are the only personnel onsite who are authorized -{/[h) s. to possess Torque Seal, and possession of Torque Seal by a member of craft is grounds for dismissal. He stated that he knows that Torque Seal is c, occasionally in the possession of craft personnel and that they use it. r p ,J / Individual A explained that when a hanger is installed using hilti bolts, 1 k a hilti bolt QC inspector is supposed to verify the torquing procedures. ( f He stated this inspector will then put Torque Seal on the bolt to ensure 4 that it is not moved. He stated that when a QC inspector does a final QC inspection, at a later date, of the hanger he/she checks to ensure that l the Torque Seal has not been broken. Individual A stated he is personally aware that anchor bolts are moved and craft personnel replace the Torque Seal, therefore, the integrity of the Torque Seal is questionable. Indi- ) vidual A was unable to identify any personnel at CPSES who could provide additional information concerning this allegation. I Interviews of B&R Quality Control Insoector On September 15, 1981, Individuals H, I, J, and K were interviewed regarcing the misuse of Torque Seal. Individual H stated he has never observed any problem relating to the use of Torque Seal, nor has he ever observed craft l .m.

~ 5 .?. s. TEXAS UTILITIES GENERATING COMPANY S003 3RYAN ' rower DAL1.A& TEXA8 78901 1 4 m L/ October 21, 1980 Js n.a.e4=v ""li.". :".*".t"."' TXX-3223 Mr. W. C. Seidle, Chief Reactor Construction & Engineering Support Branch U. S. Nuclear Regulatory Commission Office of Inspection & Enforcement 611 Ryan Plaza Dr., Suite 1000 Docket Nos. 50-445 Arlington, Texas 76012 50-446 i COMANCHE PEAK STEAM ELECTRIC STATION HILTI-KWIK CONCRETE ANCHOR BOLTS FILE NO: 10110

Dear Mr. Seidle:

\\ In accordance wit 10 CFR 50.55(e) e are submitting the attached interim report concerning' deficiencies i he installation of Hilti-Kwik concrete anchor bolts. We previously reported this item verbally to your Mr. R. G. Taylor on October 1,1980. We anticipate submitting a follow-up report on the planned investigations by January 23, 1981. If we can provide any additional infomation, please advise. Very truly yours, R. . Gary RJG:dk Attachment cc: NRC Region IV - (0 + 1 copy) Director, Inspection & Enforcement - (15 copies) c/o Distribution Services Branch, DOC, ADM. U. S. Nuclear Regulatory Comission Washington, D. C. 20555 gwwas - F31A-85,9-0 I 0

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(, i s ! f 1 TXX-3223 Attachment October 21, 1980 4 ] HILTI-KWIK CONCRETE ANCHOR BOLTS j Description of Deficiency ] As a result of an allegation to Construction management, Ultrasonic Testing (UT) of a number of Hilti-Kwik concrete anchor bolts was accomplished to o' ' verify that the installed lengths were consistent with a code letter designation-on the end of the bolt and in compliance with design require-mentsc.-Two of approximately 50 anchor bolts examined were indicated by UT as being shorter than required. Both anchor bolts were associated with a single conduit support in the Unit 1 Cable Spread Room. One anchor bolt, which was removed by jacking, was found to have been field modified in an unauthorized manner to shorten the anchor bolt by cutting the mandrel and a grinding a new mandrel into the shorter shaft. The second anchor bolt was checked for torque prior to removal by jacking, and the required torque value 4 could not be obtained. This second anchor bolt had been. modified in a manner similar to the first. px.s.' :. i Analysis of Safety Implications i' l After evaluating the previously described Hilti-Kwik anchor bolts, it has been determined that due to the unauthorized and uncontrolled modification 6 jj of these items sufficient data is not available to detemine their structural li capabilities. Therefore, it has been concluded that this deficiency could l' have adversely affected safe plant operation due to the conduit supports' inability to sustain the design loading conditions. 9 l-Corrective Actions d i! The following actions have been or will be taken to correct this deficiency { and to ascertain whether additional problems exist with these items: l A. The identified Hilti-Kwik anchor bolts have been removed. The support will be reworked and/or l relocated to meet the required design criteria. l }, l b B. To evaluate the extent of the identified deficiency, p a detailed program of ultrasonic testing is planned. The initial phase will include a sampling program l ;' of Hilti-Kwik anchor bolts installed by functional craft disciplines and subcontractors (e.g., mechanical, l' electrical, instrumentation, HVAC, etc.). The results C of these examinations will be used to detemine any I future actions required to provide assurance of the integrity of installed expansion anchors. I. yew W ~. - -- w m m * - +++, ,t-..p. ,y;

n n TXX-3223 October 21, 1980 The test program is currently under development, and implementing instructions are being prepared. We anticipate beginning the evaluation on or before November 3,1980, with scheduled completion i by December 31, 1980. C. It has been emphasized to craft personnel by Construction management that any unauthorized modification or alteration to permanent plant components will result in disciplinary action. l 4 h 4 4 . I . 1 I 4 i I 4 0 l 9 -- w- . ~.

4 3 3. Action on Licensee Identified Desion/ Construction Deficiencies ] (Closed) Use Of Architectural Concrete on Floor Slabs. j This item was initially reported to NRC Region IV by telephone on j July 2, 1980, with formal notification by a licensee's letter dated i August 8, 1980. In summary, the issue reported dealt with the I 4 possibility that the design embedment of concrete anchor bolts might not have been achieved when embedded through a two-inch layer of concrete placed on top of the structural floor slabs placed to obtain a smooth finish for the floor. The topping could not be considered a strength member since there was no assurance of its being bonded to the structural concrete. The anchor bolts attach various pipe, conduit, and instrument support devices to the building structure. The licensee's engineering forces conducted an extensive survey of all devices then installed to determine if the anchor bolt embedments were adequate to meet design loads. Due to the low loadings involved and the anchor bolts specified, the engineer was able to establish that the anchor bolts used for conduit supports were adequate on a generic basis. The pipe and instrument support devices were analyzed on a case basis and where the design requirement was not satisfied, were either corrected immediately 'or were identified for later rework.- The licensee's status reports indicate that the relatively few instrument supports requiring rework have been reworked. One hundred forty-two pipe supports that require rework which has not yet been done, have been documented on Nonconformance Report M-81-01667. These supports will be reworked as directed by the Nonconformance Report with the repairs documented on an individual basis by Quality Control personnel. Ongoing work accomplished after the date of the original notification was controlled by both procedural means and by engineering changes to drawings. The procedural mechanism involved posting large signs in all areas where the topping had been applied. I The installation procedures provided that where topping was so posted but not so noted on the device drawing, the craft would add two inches to the embedment length noted by the drawing and use the appropriate i length anchor bolt to achieve that figure. The Quality Control pro-1 cedures were revised and implemented to verffy these actions. The RRI, i through random observations of installation activities during the past 1j year and a half, is satisfied that the procedural controls have been 4 effective. Although a substantial amount of rework remains to be accomplished as documented by the above referenced Nonconformance Report, l, which will be monitored on a routine basis by the RRI, the licensee has fulfilled his responsibilities under 10 CFR 50.55(e) and this item is

}

considered closed. 4 (Closed) Uncontrolled Modification Of Concrete Anchor Bolts On September 29, 1980, a craft electrician contacted Brown & Root Construction Management with an allegation, and confession, that he and EY N* __J " ' * * - [ m. yy "*M

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~. 4 other electricians had on occasion modified concrete anchor bolts by j cutting off the wedge area and grinding on a new area in some cases or j cutting off the second wedge section where the super strength bolts were utilized to make easier the embedment of the bolt. The RRI was informally notified of the situation on October 1, 1980. The licensee followed up L1 with~a formal 50.55(e) notification on October 21, 1980, and with various

I letters culminating in a final report on the matter dated December 4,1981.

I The licensee investigated this matter with a twofold approach. Approach 1 dealing narrowly with the allegation was to identify where the particular crew had installed anchor bolts during their entire period of employment since the alleger said that the modifications were done under l1 the direction, or with the full knowledge of his foreman. Based on information provided by the alleger, the initial investigation. located three anchor bolts that had been modified. One had had the wedge cut off and crudely ground back in, one simply would not accept load from applied

1 torque, and one had been glued into place.

The licensee began a more 'i extensive examination such that all of the bolts installed by the suspect j crew were ultimately tested. Approach 2 involved a much expanded testing program to develop information as to whether other craft and/or other electrical crews had also made such unauthorized modifications to the bolts. The details of the testing program are included in the licensee's final report referenced above. The RRI reviewed the sampling basis for the investigation and found it to be statistically sound. The RRI observed portions of the initial ultrasonic calibration program and reviewed the procedures for performing the testing. The RRI observed at Various times the actual field testing activities and verified that the .l work was being done carefully by well qualified personnel. Based on the 'j licensee's report and the RRI's observations, this matter is considered closed. (Closed) Failure of Westinghouse Motor Operated Valves to Fully Close at Maximum Design Pressure. On. February 17, 1981, the licensee informally notified the RRI that notification had been received from Westinghouse that a group of 3-and 4-inch motor operated valves had been identified that might not close fully at full design level pressure. The licensee followed up with a formal 50.55(e) notification dated March 13, 1981. The NRC 1 staff also was aware of the problem which caused the issuance of IE Bul-l letin 81-02. The overall scope of the problem expanded over the past several mont to include valves up to 18 inches as well as the original sizes reported. I Ouring October 1981, Westinghouse personnel supervised modifications to the stipulated valves in accordance with Westinghcuse provided engineering information. These modifications were fully witnessed and documented by site Quality Assurance personnel. The only work remaining to be accomplished is full electrical actuation under system operating conditions which cannot be done at this time. This fact has been documented by Nonconformance Report M-2679 and appropriate arrangements made with the licensee's station l 1 ,y .-ms s amm . a r *WJwa L* "

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- ;c - u. TEXAS UTILITIES GENERATING CO.illnNY g6-/g December 4, 1981 TXX-3442 i ..;.c.., ";;' Cl **2.*l:' a I 1 i j Mr. G. L. Madsen, Chief Reactor Projects Branch l U. S. Nuclear Regulatory Commission Office of Inspection & Enforcement 1 Docket Nos. 50 445 611 Ryan Plaza Dr., Suite 1000 j Arlington, Texas 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION i HILTI-KWIK CONCRETE ANCHOR BOLTS FILE NO: 10110 i

Dear Mr. Madsen:

In accordance with 10 CFR 50.55(e), we are submitting this final report concerning identified deficiencies as a result of unauthorized field modifications of Hilti-Kwik concrete anchor bolts during the installation This matter was verbally reported to your Mr. R. G. Taylor on j process. October 1,1980. Interim reports have been forwarded by our letters TXX-3223, TXX-3266 and TXX-3303 dated October 21, 1980; January 12, 1981; and April 15, 1981, respectively. The attached report includes information previously submitted and defines l the corrective actions that were developed as part of our overall evaluation. Records supporting our study and the conclusions reached are available at the CPSES site for your inspector's revf ew. j If we can provide any additional information, please advise. 1 Very truly yours, I 'I k R. . Gary RJG:kh ,*l Attachments ,cc: NRC Region IV - (0 + 1 copy) Director, inspection & Enforcement - (15 copies) c/o Distribution Services Branch, 00C, ADH. U. S. Nuclear Regulatory Commission I Washington, D. C. 20555 }~f-h

December 4,1981 l TXX-3442 i i ATTACHMENT A HILTI-KWIK CONCRETE ANCHOR BOLTS

SUMMARY

Description of Deficiency i As a result of an allegation to Construction management, Ultrasonic I Testing (UT) of a number of Hilti-Kwik and Super-Kwik concrete expansion anchor bolts was accomplished to verify that the installed lengths were consistent with a code letter designation on the end of the bolt and in compliance with design requirements. Two of approximately 50 expansion anchor bolts examined were indicated by UT as being shorter than required. Both anchor bolts were associated with a single conduit support in the Unit 1 Cable Spread Room. One expansion anchor bolt, which was removed by l Jacking, was found to have been field modified in an unauthorized manner to shorten the bolt by cutting off the mandrel and grinding a new mandrel into the shorter shaft. The second expansion anchor bolt was checked for torque prior to removal by jacking, and the required torque value could j not be obtair.ed. This second expansion anchor bolt had been modified in a, i manner similar to the first. i More extensive UT examinations, as described subsequently, also indicated i a condition in HVAC support installation wherein the second or lower mandrel of three Hilti Super-Kwik expansion anchor bolts (out of a sample of 818 bolts UT examined) had been removed. Torquing or load data was r.ot obtained for these three bolts. g Analysis of Safety Implications Af ter evaluating the previously described Hilti-Kwik expansion anchor l bolts, it has been determined that due to the unauthorized and I uncontrolled modification of these items, sufficient data is not available to determine their structural capabilities. Therefore, it has been 'l concluded that this deficiency could have adversely affected safe plant operation due to the inability of the supports to sustain the design loading conditions. Corrective Actions The following actions have been taken to correct this deficiency and to evaluate the overall integrity of installed Hilti-Kwik concrete expansion anchors. t' L c_. .2.. ,,.__....-.___r..

December 4, 1981 TXX-3442 l i 1. It has been emphasized to craf t personnel by Construction f management that any unauthorized modification or alteration to j pennanent plant components will result in disciplinary action, j 2. Field modified Hilti-Kwik bolts have either been removed from or i cut of f and driven into the concrete structures (and patched per existing CPSES procedures). Supports involved have been reworked as required to conply with design requirements. 3. To evaluate the extent of the identified deficiencies and to correct problems, an extensive program of ultrasonic testing was } planned and implemented by the site Ouality Engineering Group. The initial phase of the program consisted of a statistically sound sampling program of Hilti-Kwik expansion anchor bolts installed by functional craf t disciplines and subcontractors (mechanical, electrical, instnamentation, fire protection and HVAC). This initial phase indicated adequate confidence that d the mechanical, instrumentation and fire protection Hilti-Xwik bolt installations had not been modified in an unauthorized manner. The second phase of the program included further examination of HVAC and electrical installations. Details and results of the UT examinations are provided in Attachment B. Based on all of the examinations performed, we are confident that the i j support installations involved are now in full compliance with design and installation requirements and will perform their intended function. i 9 i i i a, .--~-,.m__._aw.. ,,,,---,~.~c_-

.. ~. December 4,1981 TXX-3442 l ATTACHMENT B 1 HILTI-KWIK CONCRETE ANCHOR BOLTS j DETAILS AND RESULTS OF UT EXAMINATIONS 'l Proce&re Development and Implementation Ultrasonic Examination of Hilti bolts was developed to determine the length and type of Hilti bolt after they were installed. Actual Hilti bolts were used as reference blocks to produce an acoustical equivalent to the installed bolts. One bolt of each size and type (Kwik or Super-Kwik) was drawn from existing stock and collectively these bolts were used to develop the testing procedure used in the field. The UT instrument used is capable of generating frequencies of at least 1 Mz through 5 mz and provided linear presentation (within + 5 percent) for at least 75 percent of the screen height (sweep line to tiip of screen). Glycerin was used as a couplant and the transducer was in the frequency range of 1 MHz through 10 MHz. To ensure that the ultrasonic presentation seen by the inspector on the CRT screen during inspections } would be interpreted correctly, a predetermined screen pattern for each bolt type and length (as developed frcn the reference block graph) was produced and provided to the inspector in figure fom. All testing was performed under close supervision of the Quality Engineering Group. All inspectors perfoming UT of Hilti bolts are senior in status, having obtained at least their Level II certification in j accordance with the criteria established in SNT-TC-1A. All inspectors I have a minimum of 7 years proficiency as a Level II Technician in UT NDE. t Additionally, all inspectors were given additional instruction in testing i of Hilti bolts by ultrasonic means by the CPSES Site Level !!! NDE Supervi sor. Initial Sampling Program Using the Esti.mation Theory *III, a number was derived, such that, if the .0 observed data corresponded to the predicted results we could be 09.5 percent confident that 99 percent of the bolts have the correct embedment and configuration. The required number of bolts to be inspected was 782 l with 1 or fewer bolts failing inspection. 1 The equation is shown as follows:

• (1) Probability and Statistics for Engineers and Scientists, R. E. Walpole and R. H. Myers, Macmillan Company, N.Y., (31972, pages 201-205.

-.......~ ~ ~ ' - - December 4,1981 i TXX-3442 l Lg= 2.810 (normal random variable) i or = .005 (1-m = confidence coefficient) $= .01 (estimate of proportion unacceptable) 8 ( .99 (estimate of proportion acceptable) = } e= .01 (maximum error) i n. (Z.v2Y s 2 (2.810)2 (.99) (.01) = v 782 (.01)2 e Test results were analyzed using the standard test of hypothesis where the null hypothesis (H ) is that the proportion defective ($) is equal to.01. o The alternative hypothesis is that p is greater than.01. The tests were l' conducted at the 99.55 confidence level. Since the sample size (n) is large and p is very close to zero, the Poisson approximation of the binomial was used, as follows: r (x;.01n) , where the critical region is t.005. X= For four of the five groups (Mechanical. Electrical, Instrumentation and HVAC) a minimum of 782 Hilti Bolts were tested for correct length. The Fire Protection group had fewer than 782 bolts installed at the time of the sampling program; therefore all bolts in that group were tested. The results of the examinations are as follows: Fire Protection - All bolts UT'd (161). All bolts acceptable. Instrumentation - 783 bolts UT'd. All bolts acceptable. Mechanical - 816 bolts UT'd. All bolts acceptable. i HVAC - 818 bolts UT'd. Three bolts found to be unacceptable. Electrical - 796 bolts UT'd. All bolts acceptable. i The results of the sampling program indicated additional testing was required in the HVAC. Additional testing was also undertaken in the areas of the plant worked by the crew involved in the original allegation. Note that the sampling program results indicate that unauthorized modification in electrical installations was isolated. '} M me-ere + w h g m_ .mmms r, g,

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• December 4,1981 TXX-3442 i

Expanded Examination in HVAC and Electrical Installations Based on the UT examinations discussed previously, expanded examinations ) were initiated in the HVAC and Electrical Hilti bolt groups. Except as noted herein, all installed HVAC and electrical conduit support Hilti expansion anchor bolts installed in areas worked by the involved crew were examined. In isolated instances, where personnel safety would have been comoromised or where subsequent installation of material rendered installed Hilti expansion anchor bolts inaccessible (as detemined by the Quality Engineer), ultrasonic examinations were not performed. The majority of the inaccessible installations involved ceiling mounted supports with the predominant mounting device being the Richmond insert. i One-quarter inch (1/4") Hilti bolts were also excluded from the UT examination for the following reasons: 1. Hilti Super-Kwik bolts are not available in 1/4" diameter. 2. All Hil,ti-Xwik bolts, by design, only have one mandrel. The findings in the electrical arena show that a Hilti-Kwik bolt with a modified mandrel cannot achieve the required torque value. j 3. 1/4" Hilti-Kwik bolts have not been utilized in HVAC installations. 4. Comparisons of embedment requirements for the 1/4" Hilti-Xwik bolt versus the amount of concrete coverage over rebar indicate that the probability of rebar interference is minimal, and thus there should have been no reason to field modify the Hilti-Xwik l bolts. t A total of 4870 additional Hilti bolts were examined in the HVAC group; 1 ,i bolt was found to be nonconfoming due to mandrel modifications. In the Electrical group 3641 additional Hilti bolts were examined; 7 bolts were identified as nonconforming due to mandrel modifications. In summary, a total of 8511 additional Hilti bolts were examined, representing all bolts installed with the exception of the inaccessible and 1/4-inch bolts described above.

efe - s UNITED STATES ps a* o,%,' .) NUCLEAR REGULATORY COMMISSION l l f l R EGloN IV =J ' C* 1 $ $11 RY AN PLAZA oRIVE. SUITE 1000 I. / 8 ARLINGTON, TEX AS 76012 5%,,,.e September 19, 1980 ? In Reply Refer To: RIV u Docket No. 50-445/Rpt. 80-18 M M ' 50-446/Rpt. 80-18 l Texas Utilities Generating Company ATTN: Mr. R. J. Gary, T.xecutive Vice President and General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen: This refers to the inspection conducted by our Resident Reactor Inspector, Mr. R. G. Taylor, during August 1980, and by Mr. L. D. Gilbert on August 27, 1980, of activities authorized by NRC Construction Permits No. CPPR-126 and 127 for the Comanche Peak facility, Units No. I and 2, and to the discussion of our findings with Mr. R. G. Tolson and other members of your staff at the conclusion of the inspection. Areas av=4ned during the inspection and our findings are discussed in the 3 i enclosed inspection report. Within these areas, the inspection consisted of l selective av=ination of procedures and representative records, interviews with personnel, and observations by the inspector. e During the inspection, it was found that certain activities under your license appear to be in noncompliance with paragraph 50.55(e) of 10 C7R 50 of the NRC Regulations. The item of noncompliance and references to the pertinent require-ments are identified in the enclosed Notice of Violation. Corrective actions to prevent recurrence were complaced prior to the conclusion of this inspec-I tion; therefore a reply to the Notice of Violation is not requested. 9 Two new unresolved items are identifie'd in paragraph 4 of the enclosed report. l t t i t 1 u e 1 jg Spy ep-U E01A-85-59 6m

y Docket No. 50-445/80-18 50-446/80-18 Appendix A NOTICE OF VIOI.ATION Based on*the results of an NRC inspection conducted during August 1980, it appears that certain of your activities were not conducted in full com-pliance with the conditions of your NRC Construction Permits No. CPPR-126 and 127 as indicated below: Failure to Reoort a Significant Construction Deficiencv r 10 CFR 50.55(e) states in part, "(1) If the permit is for construction of a nuclear power plant, the holder of the permit shall notify the Commission of each deficiency found in design and construction which, were it to have gone uncorrected, could have affected adversely the safety of operations at any time throughout the expected lifetime of the plant. Contrary to the above: i[ The NRC Resident Reactor Inspector determined that nearly two hundred welds in safety-related piping systems have been reported by Brown & Root Quality Assurance as being under-sized (and therefore presumably O

==d r-=tr =sta) vi x c==<ar =c a > art a "-2215, initia117 dated April 23, 1980. Records available to the RRI indicated that the condi-tion had not been properly reported to the NRC as required by CHL 50.55(e). ~! 1 This is a deficiency. .f a t i

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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 50-445/80-18; 50-446/80-18 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 and 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection Conducted: August 1980 f f//)/2 Inspectors: a fp Y G. Taylo f, ion ~ Resident Reactor Inspector tatre Projects Sdet i f/ f//$/$e L. D.' Gilbert, Reactor Inspector, Engineering Support tate' Section (paragraph 2.e.) (August 27, 1980 only) Approved: /I 9///x/ b

1. /$#/A 4 W. W Crossma Chief, Projects Section

/ Da(e .620 14 4 .{ R. E. Hall, Chief, Engineering Support Section /Dat,4 l i j Inspections Summary: Inspection During August 1980 (Report No. 50-445/80-18; 50-446/80-18) ( Areas Inspected: Routine, announced inspection by the Resident Reactor Inspec-l tor (RRI) with support by a Regional II Inspector including general site tours; follow up on previous inspection findings and significant construction defi-ciencies; electrical installation activities; protection of installed equip-l ment; piping installation activities; and instrument installation activities. The inspection involved one hundred-four inspector-hours by the RRI and three inspector-hours by an Engineering Support inspector. Results: No items of noncompliance or deviations were identified in five of i i the areas. One item of noncompliance was identified in the sixth area (deficiency - failure to report a significant construction deficiency - { paragraph 6). . j l +4P aNq_ & _g g 4 'M _? M g* t*WMe,M * -' "-TWW

i 4 Texas Utilities Generating Company 2 Septe=ber 19, 1980 0 In accord =ce with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a, copy of this letter and the enclosed inspection report will be placed in the NRC's Public Docu=ent Room. If the report contains any information that you believe to be proprietary, it is necessary that you submit a written application to this office, within 20 days of the date of this letter, requesting that such infornation be withheld from public disclosure. The application must include a full statement of the reasons why it is claimed that the information is proprietary. The application should be prepared so that any proprietary infornation identified is contained in an enclosure to the application, since the application without the enclosure will also be placed in the Public Document Room. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.- Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely, W.C.Seidi,I, Chief Raactor Construction and O z=s1= ri== sePPort ar =ch f

Enclosures:

1. Appendix A, Notice of Violation 2. II Inspection Report No. 50-445/80-18 1 50-445/80-18 cc: w/ enclosures Texas Utilities Generating Company ATTN: Mr. H. C. Schmidt, Project Manager j 2001 Bryan Tower Dallas, Texas 75201 I I } l l' li ~^ ~~ "Z D -*

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j l i The RRI had no further questions regarding this matter. O (Closed) Infraction (50-445/80-13; 50-446/80-13): Failure to Follow c. Electrical Inspection Procedures. The licensee committed to a sampling verification program relative to the multi-pin connector installation in his letter to RIV, dated June 6, 1980..The RRI has reviewed inspection records and has interviewed one of the two QC , inspectors involved in the reverification which was accomplished on a 100 percent basis rather than by sampling and is satisfied that all installed connectors are properly wired as of this time. The licensee has revised Quality Instruction QI-QP-11.3.28.17 " Verify Prefabricated Cable Crimping and Pin Orientation," to require either a visual or electrical continuity verification that each wire has been installed in the correct position in its connector. This i Revision 2 became effective on July 16, 1980. ~ The RRI had no further questions regarding this matter. d. (Closed) Infraction (50/445/80-15;50-446/80-15): Failure to Establish Quality Assurance Program for Class 5 Pipe Support Systems. The licensee's letter to RIV, dated August 18, 1980, committed to appro-priate revision of the governing project specification (2323-MS-46B) and to the issuance of quality procedures necessary to carry out a quality assurance program for ongoing fabrication and installation of the Class 5 type pipe supports and to accomplish a back-fit inspec-tion of those already fabricated and installed. To this end, the licensee has issued QI-QP-11.11-1, " Installation Inspections of (] Class 5 Pipe Supports," as of July 25,, 1980. The RRI has interviewed v QC personnel assigned to the back-fit inspection effort and is

i assured that the effort has been initiated and will be ongoing for a sustained period until properly completed.

The RRI has reviewed the above indicated procedure and has found it to be consistent with the revised requirements of Project Specification 2323-MS-46B and related standards, primarily those of the American Institute for Steel Construction. The RRI had no further immediate questions on this matter but will follow the effort through routine inspection. ,5 (Closed) Infraction (50-445/80- 7;50-446/80-17): Failure to Follow Drawing for Weld Prep Details. }'The licensee informed RIV by letter, e. dated August 29, 1980, that they were of the opinion that the !f reported condition was not a violation, based primarily on the advice (j of the NSSS supplier, Westinghouse. The RRI was cogniaant of the j licensee and Westinghouse position and asked that the RIV Engineering i Support inspector that made the finding return to the site to again review the applicable drawings and the actual installed components. I This was accomplished on August 27 with the result that it was deter-mined that the drawings had been misinterpreted and that the com-l ponents were satisfactory. j o .i a 1 n.-

_.2.. ~ - DETAII.S O 1. Persons Contacted i Principal Licensee Emolovees

• B. R. Clements, TUGCO, Vice-President, Nuclear Operations
• D. N. Chapman, TUGCO, Quality Assurance Manager
• R. G. Tolson, TUGCO, Site Quality Assurance-Supervisor
• J. R. Ainsworth, TUGCO, Quality Engineering Supervisor
• A. Vega, TUGCO, Lead Engineer, QA Central Staff Other Persons 1

J. V. Hawkins, Brown & Root, Acting Project Quality Assurance Manager F. W. Gettler, Gibbs & Hill, Vice-President for Power Engineering The RRI also interviewed other licensee and Brown & Root employees during the inspection period including both craft labor and QA/QC personnel.

• Denotes those persons with whom the RRI held on-site management meetings during the inspection period.

2. Action on Previous Inspection Findings 1 (Closed) Infraci. ion (50-445/79-27; 50-446/79-26): Failure to Follow a. Q Procedures for Hoisting Safety-Related. Components. The licensee i stated in his letter, dated December 18, 1979, that specific pro-I cedures would be revised to provide appropriate instructions to the craft labor for rigging and hoisting of safety-related equipment to the extent that the component manufacturersprovided special instructions. The RRI has verified that the indicated procedures . l have been appropriately revised and that those components requiring special care have been identified and necessary instructions have been prepared. The RRI had no further questions on this matter. b. (Closed) Infraction (50-445/80-13; 50-446/80-13): Failure to Follow Welding Procedures. The licensee notified RIV by letter, dated ] June 6, 1980, that in their opinion the weld observed being welded out of procedure was technically sound as welded but acknowledged j that the welders must better understand the need for strict welding ' j procedure compliance. Toward this end, the licensee committed to a retraining program for both welder supervision and for the welders. l The RRI has verified that such retraining has been provided and has observed no further instances where the weld procedures involving different processes have not been followed. 4 4 O - 1 2 5 7 3__3p ,_ -: y _.._ _ ----.-.3;; __ _ ~y

This matter is considered closed. Q f. (Closed) Deficiency (50-445/80-18;50-446/80-18): Failure to Report Significant Construction Deficiency. This item is considered closed on the basis that corrective action of training nature that should prevent recurrence was initiated and completed during June 1980. The RRI has verified through interviews that the training has taken . place and has apparently been effective. This matter is considered closed. 3. Site Tours The RRI toured the safety-related plant areas several times weekly during the inspection period to observe the general progress of construction and the practices involved. Six of the tours were accomplished duri'2g portions of the second shift where the main activity involves the installation of electrical cables and the application of protective coatings. li No items of noncompliance or deviation were identified. 4. Electrical Installation Activities q The RRI made a number of observations of electrical cable pulling operations during the period. The RRI observed the activities of four of the seven cable pulling crews in order to ascertain whether they were working within the parameters of the site installation procedures and good practices. The O a*1 1== e6= r a ta =cti iti or ta o^/ac 9 r e== 1===1s= 4 te = iter the pulling crew work. The RRI found both groups (craft and QA/QC) to be working consistently within their respective procedures, EEI-7 and QI-QP-11.3-26. The RRI also inspected randomly selected cable tray segments in the Unit I cable spread room and in the Unit 1 Safeguards Building for freedom from cable damaging burrs and excessive debris. These segments were found to be in satisfactory condition for cable pulling activities. 6 The RRI also conducted a sampling inspection of the extensive steel struc-ture in the Unit 2 cable spread room that will utimately support the cable tray installation. This structure is defined on drawings 2323-E2-0712-01-S (Revision 4), 2323-E2-0712-02-S,(Revision 5) and 2323-S-0922 (Revi-

I sion 4) and twenty-one subordinate detailed engineered drawings.

Due to the complexity of the engineered drawing system, the site engineering 7 force was found to have developed a second set of drawings based upon ll the above identified drawings which provide a succinct number for each column and beam in the structure. The site originated drawings are FSE-00242 and FSE-00243. These FSE drawings in turn provide a reference to an entire series of drawings captured under FSE-00159 which provide details of each beam 4 and column for the purposes of fabrication, installation and inspection. The RRI obtained a total of fifteen of the FSE-00159 detail drawings and O 4

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o determined that each was consistent with the primary engineering drawing and then inspected the selected assembly for compliance. The assembly, ' Q which may be initially identified by column location 8317, and the four-teen connecting horizontal beams were found to have been fabricated and installed in accordance with the detail drawings. The RRI noted that beam seat clips welded to the column flanges and web did not-have weld returns as would appear to be required by the applicable overall standard, " Specification for Steel Building Construction," published by the American I Institute for Steel Construction. This situation is comparable to the , ), unresolved item appearing in paragraph 6 of Inspection Report 50-445/80-15; 50-446/80-15 even though that unresolved item appears in the context of v ASME Section III relative to pipe supports rather than electrical cable tray supports. The comparability arises in that the requirements of Section III appear to have been based upon the earlier AISC Specification and appear in essentially the same phraseology and in the same context. Discussions between the RRI and the licensee indicate that the licensee is developing an engineering justification for not providing the weld end returns and that this justification will be provided to the RRI for review. This matter will be considered an unresolved item pending receipt of the } licensee analysis and review by the RRI and other NRC personnel. On or about July 2,1980, the licensee notified RIV that a concern had been identified wherein concrete anchor bolts embedded in certain areas of the building floors might not develope design strength values due to having been embedded' through an architectural concrete floor topping. Q Where the floor topping is involved and not. accounted for during instal-lation of the anchor bolt, the bolt embedded length might not be adequate p to develope the design loads required. The licensee followed up his l [yg initial report of the matter in a letter, dated August 8, 1980, wherein yy [ he provided the information required by 10 CFR 50.55(e) and also stated the corrective actions being taken. Within the area of the Unit 2 cable I

• y spread room which has the floor topping applied, the RRI verified that the j

anchor bolts attaching column 8317 to the floor have an embedded length s adequate to develop the necessary strength even when the two inch thick O floor topping is subtracted from the design embedded length. The licensee p' Op ' l has identified all areas where floor toppings have been applied and is i I further identifying all embedded anchor bolts in these areas. Since each P bolt is coded with a unique letter which identifies its length, it then becomes possible to determine the actual embedded length in the structural concrete, based on an assumed two inch thick floor topping. The RRI will follow this matter until the entire effort of identification and analysis is complete, expected by the licensee to be early in 1981. 1 { Until completed, this matter will be considered to be an unresolved item. No items of noncompliance or deviations were identified. 4 O 5 m m n-n m-m.~

5. Protection of Major Installed Eouipment The RRI observed that the Reactor Pressure Vessels in Units 1 and 2 were covered and protected in the manner recommended by the supplier. The Unit I reactor core support structures (internals) remained in their enclosures within the Reactor Containment Building, while those for Unit 2 continue to be protected by outdoor area enclosures. The RRI observed that randomly selected electric prime movers for pumps have their space heaters energized as have the motors associated with motor operated valves. Random visits to the main control room indicated to the RRI that adequate air conditioning was being maintained to protect the installed components. No items of noncompliance or deviations were identified. .] 6. Safety-Related Piping System and Hanger Installation. The RRI made several general observations of the handling practices for piping components during the inspection period both in the on-site fabri-cation shop and within the main plant buildings. These practices were consistent with the requirements of Construction Procedure 35-1195-CPM 6.9 and good industry practice. The RRI observed the following welds being made during the inspection period: Weld No. Isometric Filler Ht. Welder (s) Process FW-7 RC-2-520-001 434788 AWT-BDR-BRS GTAW (machine) W-41 SI-1-RB-008 946100 AZC GTAW (manual)

k W 1 MS-1-R3-003 A82394 AGL-BBN SMAW FW-1 SI-2-SB-01 434788 BBI-AFP GTAW (machine) f The RRI verified that the weld procedures 99024, 99025 and 99028 for the j

GTAW (machine) process had been qualified and were being followed by the j welders as were procedures 88025 and 11010 for the manual processes. The

j noted welders were also verified to have been qualified, all in accordance

'l with ASME Section IX. r i The weld filler metals were previously determined to satisfy the require-ments of ASME Section II. j In addition to the above pipe joint welding, the RRI observed welder BKP l! during a portion of his work on pipe support CC-1-057-020. The welder 'j was observed to be using E-7018 type rod identified as heat number 643875 1 under procedure 11032. The RRI subsequently verified from documentation that the welder, weld procedure and weld filler metal were all qualified { as required by ASME Section III and other referenced ASME Sections. The RRI also obtained the support drawing (same number as support) and verified O 6 l 37 7, g.r.,

that the welder was providing the veld configuration required for the joint observed and that the support was of the overall configuration i required by the designer. During a review of the documentation in the possession of welder A2C as noted above, the RRI found a reference to 3rown & Root Nonconfornance Report (NCR) M-2215 and Nondestructive Examination Report (NDER) 3263. Discussions between the welder and the RRI indicated that the NCE docu-mented a substantial number of undersized fillet velds associated with s so'cket type weldments with the NDER capturing only the specific joint. The welder indicated that he had been working for some time adding filler metal or making entire new welds where finer metal could not be added to the initial welds. The welder was not sure just how many joints were involved but that there was a substantial number. The RRI subsequently reviewed the master copy of the NCR and found that the specific joint observed was identified as the one-hundred sixty-first joint involved and that there was an ongoing inspection program of all safety-related lj socket veld joints to identify those with undersize fillets, all to be reported under NCR M-2215. The RRI noted, during the review of NCR M-2215 and from discussions with welder AZC, that most of the welds being corrected had been accomplished 3 in the on-site fabrication shop and had been previously inspected and l accepted by QA/QC. Further discussions with cognizant QA/QC personnel revealed that the situation had apparently arisen as a result of the inspection and nondestructive examination sequence wherein a typical joint would be visually inspected (including fillet size). And subse-quently, which could be up to several days later, another QC inspector j would perform the liquid penetrant nondestructive examination of the same i joint. The QC enmher might well at that time require that the fillet surface be-ground as necessary to provide a smoother surface in order to f obtain a more meaningfull examination. Provided that the fillet size was essentially at minimum acceptable value initially, the subsequent grinding would reduce the size below the minimum allowable size. The QC examiners had not been instructed to recheck the fillet size after grinding and, i j therefore, the veldment was fully accepted. It appears that an alert QC j inspector in the field installation activity became concerned when he found some previously accepted shop fabricated spools had uniier-sized welds and brought it to his management's attention within Brown & Root via the referenced NCR. Since an under-size weld must also be considered ,j an under-strength veld and in consideration of the quantity involved, the i RRI inquired as to why the item had not been reported as a 10 C7R 50.55(e) lll procedures for doing ASME Section III work do not address 50.55(e) although item. Subsequent research by the RRI established that the Brown & Root l ij the licensee had issued procedures and instructions incumbent upon Brown i! & Root which do cover 50.55(e). Review of the licensee QA Site Super- ,i viscr's log of "Significant Deficiency Reports" revealed that the item had not been reported to the licensee and had, therefore, never been i' given the necessary analysis relative to reportability. Discussions with the licensee's Site QA Supervisor indicated that he had become aware that ji the then 3rown & Root Project QA Manager did not have an awareness of and/ or an appreciation of the need to communicate significant deficiencies R O la l l l -e 7 7

7 - i = to the licensee for purposes of satisfying 50.55(e). The licensee =anager stated that the already provided instructions were re-emphasized at all supervisory levels within the Brown & Root QA/QC organization during train-0 1== 1=== siv = i= =1a '== 198o-th ^^t 1== rv1 e== x 7 a t== 1 in the supervisory group and established that they had received the neces-sary instructions and appeared to understand their responsibilities in this area. The RRI also reviewed Brown and Root NCRs issued since early May 1980 (NCR 2250 and up to current date) for evidence of other signifi-cant reportable deficiencies within the Brown & Root scope cf work. With dut consideration to the subjective judgements required by 50.55(e), the RRI concluded that none of the Brown & Root NCRs included within the scope of the review contained reportable deficiencies. A comparable review of the entire forty NCRs that have been filed with the TUGC0 scope of QA/QC controlled effort also indicated adequate compliance to 50.55(e) require-ments. Based on the complete scope of information related above, the RRI concluded that an item of noncompliance had occurred but also that effective corrective action had been taken by licensee management. Also see paragraph 2 of this report. During the inspection period, the RRI also examined the radiographs of the below identified pipe weld joints for compliance to ASME Section III for ?, weld quality and to ASME Section V for radiographic quality. No inconsis-tencies were identified. Wald No. Isometric Line No. W-25 BRP-RC-1-RB-033 3-RC-1-158-2501R1 O W-3 & W-2 BRP-CS-2-RB-a 2 3-CS-2-225-2501R1 W-7A BRP-SI-1-RB-021 3-SI-1-303-2501R1 ii W-3 & W-2 BRP-RC-2-RB-045 6-RC-2-46-250111 W-12 BRP-SI-1-RB-016 6-SI-1-024-2501R1 W-13 BRP-CS-1-RB-01 2-CS-1-107-2501R1 W-3 BRP-SI-1-R3-059 10-SI-1-180-2501R1 W-17A BRP-SI-2-RB-0485 2-SI-2-199-2501R1 i W-9,W-6,W-8 BRP-SI-2-RB-019 6-SI-2-101-2501R1 l W-12 BRP-SI-1-RB-027 3-SI-1-057-2501R1 l t W-28 BRP-SI-1-RB-025 3-SI-1-039-2501R1

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Except as noted above, no items of noncompliance or deviations were i identified. 3 l 4:O M" 'n._-*O4MMW 7 h_ $'Wf__'___*

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7. Instrument Installation Activities The RRI observed during the inspection period that various instruments are being placed on their supports throughout the Unit 1 plant areas. The in-place instruments are generally covered and physically protected by wired in place wooden boxes which prevented the RRI'from observing whether the instrument ports were properly plugged, if not connected, and that the proper type of instrument has been installed. These inspection elements will be accomplished at a later stage of construction when the need for strong protective covers is lessened. The RRI also observed that instru-ment tubing runs are being or have been installed consistent with normal practice. The RRI noted that at the present time the FSAR does not identify specific instrument channels which are safety-related but rather provides only a general system functional analysis. Discussions with cognizant licensee engineering personnel indicated that drawings in the 2323-M1-2500 series provide the identification of those instruments with a safety func-tion and also those with no safety function but are connected via tubing to ASME safety-related piping systems. The RRI reviewed the referenced drawings series and found the identification substantially complete based on his knowledge of reactor systems except that the N-16 and Reactor Excore instrumentation were absent from the tabulation. Further discussions with engineering personnel revealed that this problem had been identified by both the licensee and NRC:NRR in their licensing reviews and would be corrected during future drawing releases and by FSAR amendment. This is not to say that the information does not exist in an engineering sense but rather that the information retrievability is very difficult and not Q in a concise, easily understood document. As an example, it appears that the only present identification of the manufacturer's brand and model number for a particular instrument is in a purchase order for the instrument. Again, it appears that this type of information will be collated and placed into the FSAR in a future amendment. This area of concern will be followed closely by the RRI and other assigned II personnel in future inspections. No items of noncompliance or deviations were identified. I I; 8. Unresolved Items ll l Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of concom-pliance, or deviations. Unresolved ite:ns disclosed during the inspection are discussed in paragraph 4 and will, in future inspection reports, be ieferred to as: l a. 50-445/80-18; 50-446/80-18 Absence of Weld Returns '!l* b. 50-445/80-18; 50-446/80-18 Embedment of Anchor Bolts Through ll Floor Topping i O l 9 l l'--. ,,; n _ c.

. _ = _. .. + _.. ..a. . ++-- ~ ~ < 9. Management Interviews - (] The RRI met with one or more of the persons identified in paragraph 1 on August 6, 8, 11, 18, 20, 21, and 27, 1980, to discuss inspection findings and the licensee's actions and positons. k - 9 + e T is 4 4 3f f i O i, i! 4! I 1 l i I 'i Ih 'i O 10 l 6Y

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.g %,e n.- .? i U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 50-445/80-16; 50-446/80-16 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 and 2 l Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas i Inspection conducted: July 7-31, 1980 [/&L[4 Inspector: . G. Taylor, Reside (pt Reactor Inspector 'Dats Projects Section hk/O Approved: W. A. Crossman, ChMf, Projects Section Date i Inspection Summary: Inspection During July 1980 (Report No. 50-445/80-16; 50-446/80-16) l Areas Inspected: Routine, announced inspection by the Resident Reactor Inspec-l- tor (RRI) including general site tours; safety-related piping installation and I welding; electrical cable installation; protection of major components; and follow up on allegations received. The inspection involved eighty-nine inspec-tor-hours by one NRC inspector. Results: No items of noncompliance or deviations were identified. L gegm FO\\A-85-5 i

n g:, DETAILS 1. Persons Contacted t Principa_1 Lienesee Employees 1 I

• J. B. George, TUSI, Project General Manager t
• J. T. Merritt, TUSI, Construction and Engineering Manger
• D. N. Chapman, TUGCO, Quality Assurance Manager i
• R. G. Tolson, TUGCO, Site Quality Assurance Supervisor

.i f Other Persons i F. W. Gettler, Gibbs & Hill, Vice-President for Power Engineering i E. Horowitz, Gibbs & Hill, Assistant Chief Mechanical Engineer R. L. Moller, Westinghouse, Site Manager The RRI also interviewed other licensee and Brown & Root employees during the inspection period including both craft labor and QA/QC personnel, i

• Denotes those persons with whom the RRI held on-site management meetings j

during the inspection period. 1 1 2. Site Tours i The RRI toured the safety-related plant areas several times weekly during the inspection period to observe the general progress of construction and the practices involved. Eight of the tours were accomplished during por-tions of the second shift where the main construction activity involves the installation of electrical cables and the application of protective .s coatings. No items of noncompliance or deviations were identified. 3. Electrical Installation Activities The RRI made a number of1 observations of electrical cable pulling opera-tions during the inspection period. The RRI observed the activities of each of the seven cable pulling crews one or more times in order to ascertain whether they were working within the parameters of the site installation procedures and good practices. The RRI also observed the activities of the QA/QC personnel assigned to monitor the pulling i: activities to evaluate their knowledge of requirements and their dili-l gence in assuring conformance. The RRI also reviewed the qualification l and training files of several randomly selected QA/QC personnel assigned to various electrical installation activities for conformance to the gui-i dance provided by Regulatory Guide 1.58 and ANSI N45.2.6, both of which are titled " Qualification of Nuclear Power Plant Inspection, Examination, and l I i 2 1, G

.c .~- Testing Personnel." The RRI examined randomly selected segments of installed cable tray for freedom from burrs and other sharp surfaces, identification, and freedom from undesirable debris. No items of noncompliance or deviations were identified. !j 4. Protection of Major Installed Equipment 1 i; The RRI observed, during the general plant tours, that the Reactor Pressure Vessels in Unit No. 1 and 2 were covered and protected in the manner recom- ' i mended by the supplier. The Unit 1 Reactor support structure (internals) remain in their enclosures within the Reactor Containment Building while those for Unit 2 remain in protective outdoor enclosures. The RRI observed I ;, that pump electric prime movers and motors for safety-related valves have protective space heaters powered by temporary wiring or by per:::anent wiring

l as evidenced by a hand warm condition. The main control room and the Hot Shutdown Panel are adequately air conditioned to limit the temperature rise due to the recent extremely hot weather from adversely affecting the rela-tively delicate electronic solid state devices and instruments. Other electrical components such as switchgear and motor control centers which need protection from excess humidity until permanently energized are heated above ambient by means of electric lights within the enclosure in accordance with the supplier recommendations and normal industry practice.

No items of noncompliance or deviations were identified. 5. Special Inspection Relative to Allegations Received The RRI performed a special inspection relative to a series of three allegations received via another NRC inspector assigned at another nuclear

l power station from an employee of that station who was a former employee of the general contractor at CPSES, Brown & Root. The alleger will here-I after be referred to as Individual A for purposes of identification. The allegations as received by the RRI were:

a. Anchor bolts used in the concrete walls of the safety-related build-D ings were pulling out under load. No specific location was given. b. There are numerous concrete voids in the building walls that can be l;l l located by sounding the walls with a hammer and listening for a hollow sound. 2 c. The Safeguards I panel on the 790' level has loose bus bars and

j ground wire connections. He had told his supervisor of the problem, but was told to forget the matter.

i

!l The RRI reviewed the Brown & Root maintained personnel record jacket relative to Individual A and interviewed his former General Foreman rela-i I tive to Individual A's activities while employed at CPSES. The following

j facts were gleaned from the records and the interview

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. ~.? T R S o -) (v 7/a5/Po ~ c Individual A was employed by Brown & Root at CPSES as an electrician a. for a period of five weeks early in 1980. The records indicate Individual A was terminated due to insubordination. b. The interview with the General Foreman indicated that Individual A had worked most, if not all, of his period of employment on the task of installing grounding cable on cable trays with run-offs to equip-ment such as Safeguards Panels. The task was predominately performed on the 790' elevation area of the Unit 1 Safeguards Building. The General Foreman stated that Individual A had been a good reliable worker. The General Foreman also stated that he had terminated Individual A for insubordination after a Brown & Root Personnel Safety Inspector had found Individual A on a high scaffold without his safety belt in violation of published site safety standards. When admonished I by the Safety Inspector, Individual A gave the Safety Inspector a "hard time". The Safety Inspector then contacted Individual A's Fore-man who was also given a "hard time" by Individual A. The foreman asked the General Foreman to terminate Individual A for failure to follow the site safety rules even when ordered to do so; i.e.,, insubordination. 9 The RRI's inspection relative to each of the allegations revealed the following facts and conclusions: (b.Allegationa It is true that the anchor bolts in use at CPSES will on occasion pull out when loaded which is the exact purpose of loading each bolt to a prescribed value, i.e., will the bolt pull out under load. This situation comes about because of the design of the bolt wherein a hole of a prescribed drill bit size is drilled into the concrete, the bolt with its pre-assembled wedges is driven in and then pulled back to set the wedges. On occasion, i [h for one of the several reasons such as not holding the drill steady, the hole in the concrete actually is larger than the drill bit size. When this happens, the bolt may not develop full wedging action and will pull out at some force less than required depending on just how much the hole is oversized. The RRI could find no evidence that the anchor bolts are pulling out after the bolt has achieved its full load as indicated by being torqued to a specified value. When a bolt does pull out because of hole oversizing, the condition can be relatively easily corrected by installation of the next larger size bolt. The allegation has no technical merit. Allegation b The RRI identified two locations where a hollow sound could be obtained by tapping a wall with a hammer. The condition was identified to the licensee, i who found several more in the same general vicinity, all in the walls of the r 4 t I _ _ _ - _ +_.

~ ~ (R; Safeguards Building I on the 790' elevation. Each of the five areas ranged from about I to 2 square feet. Each area was marked out and excavated down to the reinforcing steel or about two inches. No voids or other faulty condition was found in this stage of the investigation. The licensee then proceeded to excavate behind the reinforcing steel such that the excavations could be repaired in a sound manner. During several of the excavation, efforts, the RRI was present and observed nothing of an unusual nature. The RRI also queried the craft personnel doing the work about what might have occurred when the RRI was not present and was informed that all of the excavations revealed nothing except sound concrete. The RRI again tapped the concrete after it was excavated to a depth of about four inches and determined that in each instance the hollow sound was no longer present. The RRI can offer no explanation as to why the hollow sound exists in t some localized areas but can assert that the sound is not indicative of f }i actual hollowness or otherwise inferior concrete based on the observations l l and examinations described above. The allegation appears to have been based on an inference that a hollow sound indicates a hollow wall which has been shown to be an incorrect inference. The allegation is thus without merit. r.: Allegation e l} The RRI has inspected the Safeguards Panel referenced in the allegation, i.e., Motor Control Center 1EB3-1, and found that the power bus and ground bus connections were tight as of the time of inspection. It appears that

f the allegation might have been representative of the panel's condition

'I early in 1980 when the alleger was employed at CPSES. The panel in ques-tion, however, has been partially turned over to the licensee's test and startup organization for initial energization. The test and startup procedures provide that each power and ground bus bar connection be veri-fied to be tight prior to initial energizing of a panel. Based upon an 't interview of the cognizant test engineer and upon personal observation by the RRI, all bus connections were tight prior to initial power application to the panel. The purpose of grounding the bus is to assure the safety of operations and maintenance personnel in the event that a power connection were to become inadvertently grounded. Any loose power bus connections would be revealed during startup testing of the various devices powered from the Motor Control Center either by the device being non-opera-j ;i tive or by arcing in the loose connection. The allegation has no technical merit as of this time. 6. Pipe Hansers and Supports

?

The RRI continued the inspection of the Class 2 and 3 pipe hangers and supports initiated in June 1980 (see Inspection Report No. 50-445/80-13; 1 s ,i k

r f 50-446/80-13) with particular interest being directed toward a design configuration of several supports that were not typical of that experienced during the inspection of other major piping installations. This configur-ation is one wherein the pipe is supported on an essentially line contact basis by a simple steel member such as a box beam or a "w" beam rather than by a shoe or clamp which would have the effect of spreading the load over a larger area. Review of ASME, Section III Code, Subsection NB, NC and ND for Class 1, 2 and 3 systems respectively, and Subsection NF for supports, failed to reveal any definitive requirements although Subsection NF seemed to imply that such shoes or clamps should be utilized. The RRI l contacted knowledgeable NRC personnel in the areas of stress analysis, fracture mechanics and pipe support design for technical advice on the subject. The RRI was advised that the design configuration observed had become a relatively common feature in designing supports for piping systems which must be seismically sound. The line contact induced stresses would be reduced by a slight and unharmful deformation of the pipe and the support elements to a point well within the strength characteristics of each ]' element. The RRI had no further concern in this area. li 7. Safety-Related Piping System Installation i The RRI made several observations of the handling practices relative to stainless steel safety-related piping, particularily in regard to that '? involving the Reactor Coolant piping in Unit 2 which was being adjusted for fit-up preparatory to welding to the Reactor Pressure Vessel and Reactor Coolant Pump in No. 3 loop. The actions observed were consistent l with applicable site procedures and those of good industry practice. On or about May 30, 1980, the RRI was informed by the licensee that per-sonnel representing the fabricator of the Unit 2 Reactor Pressure Vessel were on the site to conduct a special examination of the-outlet nozzle safe-ends. The licensee stated that the fabricator had discovered that I, a comparable vessel still in the fabrication shop did not conform to the fabrication design drawing requirements in that the safe-end thickness on the outside diameter was less than specified. The reduced thickness was postulated to potentially cause cracking in the factory accomplished Inconel weld of the safe-end to the primary nozzle during the subsequent field weld of the safe-end to the Reactor Coolant pipe, particularily if 5 the field weld stainless steel were to overlap onto the Inconel weld. It was ascertained that one of the four outlet nozzles was substantially 3 thinner than specified but not so thin that any special welding technique would be required. The vessel fabricator and the supplier (Westinghouse) further established to the satisfaction of the licensee and the NRC that any cracking of the Inconel weld which might occur would be readily detectable during the normally conducted pre-service baseline inspection 6 i .I - = - =

p p. by mecas of liquid penetrant examination. Westinghouse was also able to establish that even if the Inconel weld did crack, the resulting crack would be such that the welded connection would not fail under either operating or seismic loading conditions or both in combination. The RRI 4 has reviewed the licensee's file on this matter and has had discussions of the matter with cognizant NRC personnel and has no question on the correctness of the licensee's decision that the matter was not formally reportable in accordance with 10 CFR 50.55(e) as indicated in the licensee's letter dated June 26, 1980. i The RRI observed welder BBN during a portion of the welding of joints FW-5 and FW-6 as. identified on Isometric drawing WP-1-RB-05 representing line No. 2-RC-1-035-2501R1 in the Reactor Coolant primary pressure boundry. The welder was verified to be welding within the parameters of welding procedure No. 88021 using weld wire heat No. 746100. The RRI reviewed Weld Data Card and Weld Filler Metal Log in the possession of the welder and verified that assurate data had been recorded. The RRI subsequently verified that the welder, weld procedure and weld filler metal had each been properly qualified and certified in accordance with applicable portions of the ASME Code. The RRI examined the following radiographs for completed safety Class 1 and 2 welds as denoted by the last number in the designation. The radio-graphs indicated that the welds conform to the requirements of ASME, Section III for the applicable class. The radiographs themselves met the requirements for radiograph quality as required by ASME Section V. Weld Isometric Line W-13 & 31 BRP-RC-1-RB-15 6-RC-1-108-2501R1 FW-4 BRP-SI-1-RB-54 6-SI-1-328-2501R1 FW-2 BRP-SI-1-RB-54 6-SI-1-327-2501R1 FW-6 BRP-SL-1RB-40 10-SI-1-021-2501R1 t FW-14 BRP-SI-IRB-037 10-SI-1-021-2501R1 1 FW-2 BRP-CS-2-AB-079 4-CS-2-55-151R2 W-16 BRP-CS-1-RB-01 2-CS-1-107-2501R1 W-2 BRP-CS-2-RB-037A 2-CS-2-103-2501R1 W-14 BRP-CS-2-RB-22 3-CS-2-079-2501R1 W-2 & 3 BRP-RC-2-RB-044 6-RC-2-029-2501R1 7 l woe e,-~ r _cpe-- - m -e-wee =+ -w. -~ =.

?:, .+ BRP-CS-1-RB-028 2-CS-1-112-2501R1 FW-7 W-9 BRP-SI-2-RB-52 6-SI-2-172-2501R1 W-7A BRP-SI-1-RB-16 6-SI-1-101-2501R1 W-12 BRP-CS-2-RB-013 2-CS-2-107-2501R1 JV-1 BRP-RC-1-RB-020 4-RC-1-075-2501R1 FW-11 RRP-RC-1-RB-08 2-RC-1-053-2501R1 The RRI verified that piping lines 12-SI-1-031-151R2, 8-SI-1-323-151R2 and 8-SI-1-324-151R2 as shown on flow diagram 2323-M1-0261 have been installed as required. In conjunction with this verification, the RRI also examined the piping runs of lines 8-CS-1-567-151R2, 8-CS-1-063-151R2 and valves 1-LCV-112D, 1-LCV-112E and 1-8546 as shown on piping flow diagram 2323-M1-0255. The RRI examined the training and qualification records of six of the QA/QC personnel assigned to the inspection of piping system fabrication, installation and examination. The personnel were randomly selected by the RRI and represent approximately 17% of the QA/QC force assigned to

l the activity. The training and qualification requirements were based on SNT-TC-1A as required by the ASME, Section III Code and supplemented by the Guidance of ANSI N45.2.6.

The records indicated an educational / experience level and training background that meet these requirements. No items of noncompliance or deviations were identified. 8. Management Interviews The RRI met with one or more of the persons identified in paragraph 1 on July 7, 8, 9, 10, 11, 14, 15, 16, 21, 25 and 29, 1980, to discuss inspection findings and the licensee's actions and positions. A i.

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r i INSPECIION PLAN II Inspection Raport No. 50-445/80-16; 50-446/80-16 Licensee: Texas Utilities Generating Company Location:- Glen Rose, Texas Facility: Comanche Peak, Units 1 & 2 Type of Licensee: W, PWR, 1150 MWe Type of Inspection: Resident, routine, announced Dates of Inspection: July 1980 Dates of Previous Inspection: June 1980 ? Inspectors: RGTaylor SCOPE OF INSPECTION Resident Inspection Program for Construction e 1 l l n SO 1 /**y"g.'~ d1.6 g)4o c-- [ Approve y, //.R% p / pggj t . _ _ _ _ _. - - _ _}}