ML20198C904
| ML20198C904 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/19/1986 |
| From: | Grimsley D NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Garde B GOVERNMENT ACCOUNTABILITY PROJECT |
| Shared Package | |
| ML20198C917 | List:
|
| References | |
| FOIA-85-59 NUDOCS 8605230091 | |
| Download: ML20198C904 (26) | |
Text
{{#Wiki_filter:, b-h M ~%, UNITED STATES
. 8 o Nt/ CLEAR REGULATORY COMMISSION r, a wAsmNoTow, p. c. 20sss I
%,*****,o MAY 19 E Docket Nos. 50-445/50-446 Ms. Billie Pirner Garde Director, Environmental Whistleblower Clinic Government Accountability Project IN RESPONSE REFER 1555 Connecticut Avenue, NW, Suite 202 TO F0IA-85-59 Washington, DC 20036
Dear Ms. Garde:
This is in further response to your letter dated January 21, 1985, in which you requested, pursuant to the Freedom of Information Act (F0IA), records related to the NRC Comanche Peak Task Force established in March 1984. We are now placing one additional bcx of records (box 8) in the QA/QC category, covering Appendix I records numbered one through 364, into the NRC Public Document Room (PDR) in file F0IA-85-59 under your name. The Appendix I records numbered 1 and 19 contain deletions without invocation of an FOIA exemption in accordance with the agreement reached between you and the NRC staff at a meeting held on May 8,1985, concerning your pending FOIA 3 requests pertaining to the Comanche Peak Technical Review Team activities. J This agreement was summarized in a letter to you from J. M. Felton, dated May 15, 1985. The information deleted consists of the names of allegers. 5 A separate set of the disclosed records is being made available for public inspection and copying at the Local Public Document Room (LPDR) located at the University of Texas Library, 701 South Cooper Street, Arlington, Texas. Additional boxes of records in the QA/QC category are currently under review and will be placed in the PDR and LPDR in the coming weeks. Sincerely, md W d, j Donnie H. Grimsley, Director i Division of Rules and Records l Office of Administration l l
Enclosure:
Appendix I l 8605230091 860519 PDR FDIA ( CARDEG5-59 PDR I L d
r .
\
RE: F01A-85-59
. I APPENDIX I i
BOX 8 - QA/QC DATE PAGES
- 1. 4/24/84 0.E. Eisenhut. NRC, ltr to 17 M.D. Spence TUGCO, " Comanche Peak Steam Electric Stations Units 1 and 2 Allegations"
- 2. 3/19/84 GAP ltr to Richard C. DeYoung, 19 Request Filed Pursu nt to 10 CFR 2.206-In PDR-ANO 8601270243
.3. 8/18/84 CASE's Partial Answer in 34 Opposition to Applicant's Motion for Authorization to Issue a License to Load Fuel and Conduct Certain Precritical Testing and Motion for Addi-tional Time to Responde
- 4. 9/26/84 GAP ltr to D. E. Eisenhut, 8
" Answer to Request for further information regarding several items under review by TRT."
l RELEASED
- 5. 2/15/83 G. L. Madsen, NRC, ltr to R. J. 80 Gary TUGC0 which refers to the special inspection conductid-as result of concerns expressed
~
during July and Sept.1982 evidentiary hearing sessions i
.e
t
~6. 4/11/83 NRC Ltr to R. J. Gary, TUGCO. ~
50
" Construction Appraisal In- f ,
spection 30-445/83.18, 50-446/ . 83-12
- 7. 8/19/83 Region IV Report for CAT 17 followup
- 8. 10/3/83 G. L. M'adsen, NRC Lti to R. J-.
34 Gary, RE: " Refers to Inspection conducted during period 6 8-16-84"
- 9. 1/30/84 Memorandum (Records Retrieval) 5 Application for Operating l License 1
- 10. 6/5/84 Region IV Plan for the 16 Completion of Outstanding Regulatory Actions.
11- 8/28/84 A. Vega TUGCO, Office memo 11 to S. L. Spencer, "-CPSES NRC Special Review Team Report"
- 12. 9/7/84 D. N. Chapman, TUGCO, Office 13 memo to 8. R. Clements, " Status of Items Assigned to QA"
- 13. 11/5/84 D. R. Hunter, NRC, ltr to D.
' ' 30 R. Hunter, TUGCO, ' Comanche Peak Report 84; 32; 84-11"
- 14. 11/28/84 D. R. Hunter, NRC Ltr to 19 M. D. Spence. TUGCO, RE:
Inspection Conducted under Resident Inspection Program, Aug 26, 1984/0ct 20, 1984
- 15. 1/9/85 D. R. Hunter, NRC to M. D. 11 Spense, TUGCO, RE: Inspection Conducted under Resident In-spection Program Oct 21, 1984/
December 18, 1984 e 6 2
,9
' ~
- 16. No Date -
TestimoniRegarding. '
- 13 inspection.and Testing of Non - l' ASMEcomporientsynd. Systems ,
i
'17..? No Date ,Te.stimony R'egarding Pre - 5 -
- . . Servi'ce Inspection ,and In-Service. Inspection of ASME ' Components and Systems -
- 18. No Date TUGC0 Corporate QA-Site QA 2 Activities.. Interface
- 19. No Date Safety Evaluation Heport 81 NUREG-0797 Supplement #5, 20 No Date Special Review Team Report 78 21 No Date Index of Comanche Peak 55 Operating License Proceeding AVAILABLE IN'PDR
- 22. 7-10-84 Deposition of Gordon Raymond 181 Purdy before the NRC Atomic Safety & Licensing Board Filed in PDR - AN0> 8407170l178
. 23. 8-16-84 Prefiled Testimony of Gordon 6 Purdy (in the fann of a Deposition. Filed in PDR.' ,
ANO 8408230416 24, 8/16/84 Prefiled Testimony of Gregory 20 Bennetzen before the NRC Atomic Safety & Licerising Board Fi, led in PDR. Ah0 8408230403
- 25. 11/26/84 Hearing before the.NRC Atomic Safety & Licensing Board in 326 the matter of TUGCO. Filed in the PDR - AN0' 8411280285.
- 26. 11/27/84 Heating before the NRC Atomic
.383 Safety & Licensing Board in the matter of TUGCO. Filed in POR.- ANO 841129'0272' 3 -
U
-27. 11/28/84 llearing Before the NRC 315 Atomic Safety & Licensing )
1 Board 'in the matter of TUGCO.- ' Filed in the PDR.'-
- I
.* ANO 8412010148 9 e
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- 32. 3/23/79 Component Modification Card 1
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- 33. 3/27/79 Component Modification Card 1
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- 34. 4/12/79 Component Modification Card 1
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- 35. 4/12/79 Component Modification Card 1
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- 82. 2/13/79 Control Room Drawing Job i 1 2323
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- 98. 12/30/81 Component Modification Card 2
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#2447 107. 9/9/78 Component Modification Card 1
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#00161 109. 9/20/78 Component Modification Card 1
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#4512 140, 3/1/79 Component Modification Card 1 e4516 141, 3/15/79 Design Change Authorization 2
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, & Elec. Control B1dg.
i 350. No Date 2323-Al-0500 Primary Plant Unit 1
, 1 Containment. & Safeguard Bldgs.
351. No Date 2323-El-0713-01-S-Aux & Elect. 1
- Cont. 81dgs. Cable Tray Support Plan ,
352. No Date 2323-El-0510-S-Reactor Bldg. 1 Elec. Penetration Area 24
4 ; s .'s . FSC-00423 Aux Bldg. Elev. 7920 3 353. No Date 1 354. No Date FSE-00230 -(MAP) Control Rm & Aux. Bldg. 355. No Date FSE-00221 Reactor Bldg - Cable 1 Tray Support Plan 356. No Date FSE-00191 Reference Drwg for 2 Hanger No's No Date FSE-00191-Cable Rm & Aux. Bldg. 1 357. Cable Tray Support Plan FSE-00185 Ref Dwg for Hanger Nos 3
., 358.. No Date 359. No Date BRP-51-1-YD-003 Safety Inje'ction 1 360. No Date 2323-S1-0618 Safeguards Bldg 1 Misc. Part Plan Outline 361. No Date 2323-S1-0638 Safeguards Bldg. 1 Structural Stl.
7 362. No Date CPSES Organization Chart 2323-S-0831,.0832, 0833, 0834 4 363. No Date FB Spent Fuel Pool 364. No Date 2323-5-0909 Cable Tray Support 1 - Sheet 9 e g 6
^ 25
*- Ad@18ts no
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GOVERNMENT ACCOUNTABILITY PROJECT 1335 Connecticut Avenue N.W., Suite 202 Washington, D.C. 20036 (202)232 4 550 G:85 :104 January 21, 1985
/
FREEDOM 0F INFORMATION ACT REQUEST FREriUU'.A OF INFOftMATION ACT. REQUESI, Director Mk S$$ ~ Office of Administration Nuclear Regulatory Commission Washington, D.C. 20555 To Whom It May Concern: Pursuant to the Freedom of Information Act ("F0IA"), 5 U.S.C. 5 552, the Government Accountability Project (" GAP") requests copies of any and all agency records and information, including but not limited to notes, letters, memoranda, drafts, minutes, diaries, logs, calendars, tapes, transcripts, summaries, interview re-ports, procedures, instructions, engineering analyses, drawings, files, graphs, charts, maps, photographs, agreements, handwritten notes, studies, data sheets, notebooks, books, telephone messages, computations, voice recordings, computer runoffs, any other data compilations, interim and/or final reports, status reg ports, and any and all other records relevant to and/or generated in connection with the overview, ragulation and investigation of the Comanche Peak Nuclear Plant by any person, branch, or department of the NRC since January 18, 1985. This request includes all agency records as defined in 10 C.F.R. 5 9.3a(b) and the NRC Manual. Appendix 0211 Parts 1. A.2 and A.3 (approved October 8,1980) whether they currently exist in the NRC official, " working," investigative or other files, or at any other location, including private residences. If any records as defined in 10 C.F.R. 5 9.3a(b) and the NRC Manual, supra, and covered by this request have been destroyed and/or removed after this request, please provide all surrcunding records, including but not limited to a list of all records which have been or are destroyed and/or removed, a description of the l action (s) taken relevant to, generated in connection with, and/or issued in order to implement the action (s). GAP requests that fees be waived, because " finding the information can be con-sidered as primarily benefitting the general public," 5 U.S.C. 5 552(a)(4)(a). GAP is a non-profit, non-partisan public interest organization concerned with honest and open government. Through public outreach, the Project promotes whistleblowers as agents of government accountability. Through its Citizens Clinic, GAP offers assistance to local public interest and citizens groups seeking to ensure the health and safety of their communities. The Citizens , Clinic is currently assisting several citizens groups, local governments and .
'intervenors in the central Texas area concerning the construction of the Comanche Peak nuclear power plant. . -
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Director Office of Administration . Page Two We are requesting the above information as part of an ongoing monitoring project cn the adequacy of the NP.C's efforts to protect public safety and health at nuclear power plants. For any documents or portions that you deny due to a specific F0IA exemption, please provide an index itemizing and describing the documents or portions of documents withheld. The index should provide a detailed jusitfication of your grounds for claiming each exemption, explaining why each exemption is relevant to the document or portion of the document. withheld. This index is required under Vaughn v. Rosen (I), 484 F.2d 820 (D.C. Cir.1973), cert. denied, 415 U.S. 977 (1974). We look forward to your response to this request within ten days. Sincerely, LA Billie Pirner Garde Citizens Clinic Director P 4 e
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Y {(((j' Docket Mos.: 50 445 and 50-446 Mr.11. D. Spence - [,()f/0? President Texas Utilities Generating Company 400 !! orth Olive Street Cl l l "Q O = L. B 81 Dallas, Texas 75201
Dear Mr. Spence:
Sub.iect: Conanche Peak Steam Electric Station Units 1 and 2 Allegations As you are aware the NRC staff has received allegations of improper practices during the construction of the Comanche Peak facility. Most of these allegations have been received non-technical areas.' The over the last several nanths and cover both technical and NRC staff has made an initial segre the technical issues and they are provided in the enclosure. gation ofrequested You are sone of to respond fully to' the attached. issues with your assessment of: i t (1) an evaluation of the validity of each of the specific issues identified, clearly stating the basis for your conclusions,
'(2) the safety significance of each issue in view of your conclusion in iten (1),and 1 (3) the generic implications on other systems or contractors for any iten fcund to have nerit.
Your responses should be forwarded to the NRC not later than May PS,1984 As the staff review of the allegations continues, there nay be nore technical areas provided for your action. You should also note that the NP.C Office of
-Investigations is pursuing the examination of non-technical . areas.
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COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 . cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory
-Washington, D. C. 20036 Commission P. O. Box 38 Robert. A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin 2001 Bryan Tower , 114 W. 7th, Suite 220
- Dallas, Texas 75201 Austin, Texas 78701 Mr. H. R. Rock B. R. Clements
- Gibbs and Hil1, Inc. Vice President Nuclear .
393 Seventh Avenue Texas Utilities Generating Company
.New York, New York
~
10001 Skyway Tower 400 North Olive Street ' Mr. A. T. Parker L. B. 81 Westinghouse Electric Corporation Dallas, Texas 75201 P. O. Box 355 Pittsburgh, Pennsylvania 15230 William Burchette, Esq. Law Office of Northcutt Ely David J. Preister Watergate 600 Building Assistant Attorney General Washington, D. C. 20037 Environmental Protection Division, P. O. Box 12548, Capitol Station Ms. Billie Pirner Garde Austin, Texas 78711 Citizens Clinic Director Government Accountability Project Mrs. Juanita Ellis, President 1901 Que Street, N. W. . Citizens Association for Sound Washington, D. C. 20009 Energy 1426 South Polk Dallas, Texas 75224 J
- 1. It has been alleged that docunent control clerks have issued incomplete
; fc0 " packages" to quality control and craft personnel, at the direction of
. U8 6 supervision, in violation of procedures.
- 2. It has been alleged that craft personnel rade unauthorized desicr. document chcoces in the field. They did this by writing a " traveller" which A^t 70 '
allowed them to use an incomplete design document package in the field, g q 3. It has been alleged that the document control conputer is not accurate in that it does not match the documentation. J\ 4 It has been alleged that instances have occurred where documents were lost. PTh'?2_ No log was kept of lost documents, and on occasion craft and management would find a way around the missing docunentation. An examole is they would call up a missing control modification card (CliC) on the ccmputer and delete it. AQ 735 It has been alleged that document control clerks received poor training. These clerks learned how to handle " travellers" and other types of documentation on the ,iob.
%g . It has been alleged that original, permanent records (i.e., weld data cards) are not being stored in a fire, proof vault.
7* It has been alleged that hundreds of packages'of permanent records have Y. been lost in the course of various noves.
, 8. It has been stated that document control clerks are issued " controlled stamps" which they use to certify that a document package contains the
_q 'atest information and is ready to be used to perform the work or the inspection. It is alleged that these stamps were issued to the Ouality
- nntrol Department, by management, and that they would stamp their own drawings and declare them legitimate. .
M -7 9. It has been alleged that sometimes documents are outright falsified. An g_gg example is that a date was changed on a weld data card by a cuality control (OC) inspector.
- 10. It has been alleged that craft would " bootleg" rework by per# arming repairs h-NS without any documentation.
- 11. Describe the utility and contractor programs for verifying that the "as-built" conditions accurately reflect des *ign. It has been alleged that instances h
w % - g "(ave occurred where craft, instead of following the appropriate docu . as-built" on the document. Then the actual "as-built" condition would not be sent back to Gibbs & Hill for evaluation. An exansle is a non-conformance report (NCR) that was written because material did not neet ninimum wall thickness and an engineer voided it by writing "as-built". 6
, . 12. It has been alleged that quality centrol inspectors are not qualified 13cr have insufficient training.
- 13. What is the purpose of the N-5 program? Could the program hinder the document controller from carrying out his/her . job? It has been alleged 45- that the N-5. program is making it difficult for document controllers to review records in the records vault when necessary and makes i_t impossible to review packages going into the vault.
- 14. It has been alleged that management has taken away an approved set of procedures for reviewing documents. Therefore, document controllers
-.rJQ-donothaveanythingtoreviewby.
g 15. Describe the utility and contractor programs for rec ~eiving vendor parts and certification of vendor components. It has been alleged that if
% y parts were lost from a vendor component, the fabrication shop manufactured replacements without procedures. The practice of fabricating "Q" (safety related) material without quality control oversight occurred regularly,
- 16. It has been alleged that non-conforming material has been used in safety systems. Examples include:
g /[ a) Craft " buttered" pipe', material is welded 360* around the pipe to make it thicker, to achieve rec,uired wall thickness. Mf-f b) Fipe numbers have been changed in an effort to bypass NCRs.
'q-i :. 4 c) Out of round pipe was heated up by craft and made round without procedures or analysis.
"- d d) CMC's would be " lost" so that unauthorized work could not be traced.
- 17. Describe the utility and centractor programs for changing the. class of i a material and procedures used to assure the appropriate class of material is used in the construction of systems. It has been alleced that:
a) documents were reassigned from "Q list" (safety-related) items to "non-0" (non-safety related) items to circumvent ANI review. The 4 documentation would be left as non-0; however, the material was placed in a safety system. b) a department upgrades the class of a material to fill an order, if material requested is not available. An example is that craft will look for material requested on drawings and if it cannot be fourd,
. -1 they will substitute a sinflar looking material. The material will then be stamped with a number that incorrectly identifies the class of the material.
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, 18. It has been alleged that there was inproper sign off of " hold points" g on travellers.
- 19. It has been alleged that craft personnel would satisfy a CMC on an inadeouate weld by welding over it instead of following the procedure A Q tC O 2 of cutting it nut and then welding.
- 20. It has been alleged that there are undocumented weld repairs. Modification I%$)[-L433Lwere made to material, such as a hanger, after OC hed acoroved it.
- 21. It has been alleged that safety related welds were repaired with weld toch (W. T.) holdpoints, instead of QC hold points in violation of
/kd}bl)' add) procedures. Ilhat are.the effects of this?
- 22. Although the Ouality Control (OC) and Construction Organizations are represented as. independent, it has been alleged that current practices fk()-k between QC an construction compromise the independence of these two organizations.
~,9I .
'2 It has been alleged that craft bypassed procedures by telephoning orders to the fabrication shop in lieu of sending drawings.
It has been alleged that there is constant pressure, by craft and N L4 U3f 7 EU nanagenent, on 0A/0C inspectors not to write non-conformance reports. 0 0 e e e D 0 I i - t 4 E
TEXAS UTILITIES GENERATING COMPANY IBKYWAY TOWER
- 400 NOMTH OLIVE STREET. L.B. 58
- DALLAm.TEXAss 73:401
. "4t%" "'"'01 June 1, 1984 TXX-4187 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Dockets: 50-445 Washington, DC 20555 50-446 Comanche Peak Steam Electric Station Units 1 and 2 Allegations Transmitted by Letter of April 24, 1984 Additional Response File No.: 10125
Dear Mr. Eisenhut:
This letter provides our additional response to your letters of April 24, 1984 and May 1, 1984 containing a list of allegations about certain practices at the Comanche Peak Steam tiectric dtation. Our TXX-4180 dated May 25.. 1984 provided our response to 16 of the 24 allegations. Responses to the remaining eight allegations are contained herein. We have responded to these allegations to the fullest extent possible commensurate with the amount of specific information provided to us. To the best of our management's knowledge, no documentation relating to this evaluation has been discarded or destroyed. We trust you will find the enclosed information helpful in expediting closure of these issues. Please advise if you require further information. Very truly yours, BRC:In m cc: John Collins NRC, Region IV f r;lL f O\($*l4 gi . . r A D8VtERON OF TEXA0 (?TSL8 TIES EE.ECTRIC COMPANY I
TXX-4187 6/1/84 . Allegation No.1 It has been alleged that document control clerks have issued incomplete 40 O ~ " packages" to quality control and craft personne'1, at the direction of supervision, in violation of procedures. Evaluation of Validity Applicants are unaware of any instance where a supervisor directed document control clerks to violate a procedure by issuing incomplete packages. Our review indicates that this allegation may stem from a misunderstanding of certain CPSES procedures. Procedure DCP-3, Rev. 17, at Section 3.4 authorizes the issuance of documents into " traveler" packages. This process allows the issuance of only those documents required for the specific work covered by the traveler. These documents.are then stamped "This document shall be used only in - conjunction with Operation Traveler #
- This practice precludes the documents from being used in a manner other than that intended by management. This procedure is certainly preferable to burdening the inspectors or craft personnel with documents not related to their assigned task.
Personnel may have been confused with what constitutes a " complete" package. It is our position that the procedure provides for issuance of a package complete for the intended use. QC inspections, verification of design changes and the walkdown programs provide an additional level of assurance that all required documents have been utilized. Safety Significance None Generic Implications on Other Systems or Contractors Not. applicable J
TXX-4187 6/1/84 Allegation No. 5
.t e , 2 It has been alleged that document control clerks received poor
/\ ' j' J training. These clerks learned how to handle " travelers" and other types of documentation on the job.
Evaluation of Validity In accordance with Procedure DCP-3, document control clerks are required to receive training commensurate with the complexity of their job function. This training comprises document control orientation (including instruction in handling of travelers) and a passing grade (15 out of 16 questions correct) on a test which is based on a thorough knowledge of DCP-3. Records of training and test results are maintained by DCC. We have reviewed these records and have confirmed that all clerks presently employed and employed within the last six
, months underwent this training and testing. Obviously, additional training in the broad sense may be received "on the job" as individual clerks are faced with opportunities to learn even more about document control.
Safety Significance None Generic Implications on Other Systems or Contractors Not applicable
TXX-4187 6/1/84 Allegation No. 14
. It has been alleged that management has taken away an approved set of
- c),i'# procedures for reviewing documents. Therefore, document controllers do not have anything to review by.
Evaluation of Validity Our investigation of this allegation indicates that on Transmittal #57, dated 4/18/84, Brown & Root QA procedures CP-QAP-18.2 "QA Review of ASME III Documentation," and CP-QAP-18.3 "QA ASME III N-5 Certification," were deleted. This was done in anticipation of ' incorporation of the contents of these procedures into CP-QAP-12.1. As. this was Easter weekend, document reviewers did not work. On Monday, April 23, it was evident that the new revision of CP-QAP-12.1 was not going to be ready for issuance. Transmittal #58 was sent out explaining that CP-QAP-18.2 and 18.3 were deleted in error and reinstated those procedures. They are.still in effect as the new
~
revision of CP-QAP-12.1.has yet to be issued. We are not aware of any other instances. Safety Significance None Generic Implications on Other Systems or Contractors Not Applicable
)
TXX-4187 6/1/84 Allegation No. 16 It has been alleged that non-conforming material' has been used in safety systems. Examples include:
. a) Craft " buttered" pipe, material is welded 3600 around the pipe to I, J , J make it thicker, to achieve required wall thickness.
/ b) Pipe numbers have been changed in an effort to bypass NCR's.
c) Out of round pipe was heated up by craft and made round without procedures or analysis. d) CMC's would be " lost" so that unauthorized work could not be
- traced.
Evaluation of Validity Our investigation of the four cited examples indicates the following:
, a) Piping.which was identified as being in violation of minimum wall requirements has been repaired by welding to achieve the required wall thickness. This practice was approved in Construction Procedure CP-CPM-6.90 and is considered a standard repair method for such situations. The repair of minimum wall violations was documented by a repair process sheet or by Welding Engineering adding additional steps on the Weld Data Card for the weld involved.
b) ASME Section III requires heat traceability for all pressure retaining materials. Heat number verification is documented on the Manufacturer's Record Sheet (MtS) and verified by both QC inspection personnel and QA Document Review personnel. There have been some instances when QC inspections have identified mismatches between numbers on the drawing and on the hardware. Pieces identified on the drawing have been incorrectly identified on the hardware, or in some cases, the fabrication drawing or Component Modification Card (CMC) would incorrectly duplicate piece numbers,
.i.e., use the same piece number for more than one piece of the
TXX-4187 6/1/84 subassembly. In the first case, QC inspection personnel would have the craft correctly identify the spool. In'the latter case, QC would notify the craft that the drawing (or CMC) was in error. If engineering agreed and revised the drawing (or CMC), the craft would correctly identify the piece, at which time QC would perform its inspection. c) Our investigation of this allegation on safety related piping did not substantiate that this has occurred. The basis for this statement is interviews with construction supervisors and numerous QC inspectors. Additionally, both TUGC0 and Brown & Root NCR logs have been' searched, and no evidence of this type of activity was encountered. d) The purpose of the CMC is to document a design change which is required because of interference, minor fabrication error, etc. A
" lost" CMC would make an as-constructed condition appear nonconforming', rather than making " unauthorized work" untraceable.
CMC's are issued to construction through the Document Control Center (DCC) and logged as they are issued. As an additional safeguard, the verification of design changes and the N-5 walkdown verify that the as-installed condition matches the latest design. Hardware which does not match the latest design is identified by QC as nonconforming. Safety Significance None Generic Impact on Other Systems or Contractors Not applicable
1 e- TXX-4187 6/1/84 Allegation No. 17 Describe the utility and contractor programs for changing the class of I a material and procedures used to assure the appropriate class of material is used in the construction of systems. It has been alleged , that: . I a) Documents were reassigned from "Q list" (safety-related) items to q $ "non-Q" (non-safety related) items to circumvent ANI review. The documentation would be left as non-Q, however, the material was placed in r,afety system. i! b) A department upgrades the class of a material to fill an order if~ ' material, requested is not available. An example is that craft will Ig look for material requested on drawings and if it cannot be found, I . they will substitute a similar looking material. The material will then be stamped with a number that incorrectly identifies the class of the material. l Evaluation of Validity > Material classification may properly be changed either downgrading material or upgrading. Material, on occasion, has been upgraded, i.e., ASME Class 2 upgraded to ASME Class 1. Both the Brown and Root QA Manual and ASME QA/QC procedures define specific provisions for doing so. A higher classification, i.e., Class 1, is always acceptable for use in a Class 2, Class 3, Class 5, or 80P system. Material is not ' upgraded from non-safety to Class 5 or ASME. a) Our interpretation of this , allegation is that systems or portions of systems have been changed from ASME to Class V (Non-ASME) or nonsafety. This has happened as a result of Engineering reclassifying systems or portions of systems. These changes were based on design considerations, and were not made in order to l preclude or circumvent ANI review. l
TXX-4187 6/1/84 b) Construction procedures allow the substitution of alternate materials which have been approved by Engineering. If neither the specified material nor the approved alternate is available, Engineering is contacted to provide additional alternatives or to work with Construction and Procurement to prepare Field Requisitions to procure necessary materials. The quality program requires that this material is checked at receiving by QC to assure compliance with purchase order requirements. QC field inspection personnel verify material type, grade, and heat numbers (where required). This verification is designed to assure that material used meets design requirements. Two safeguards preclude the possibility of changing the classification of material by simply "re-stamping" the material:
,1) If material reclassification were to occur as alleged, the old , classification would have to be ground off prior to stamping the new number and would be readily apparent to QC; and
- 2) For material requiring' heat number traceability, the material classification would be caught in Document Review by the mismatch between ASME Class and heat number, i.e., heat number would be for a different Class than indicated.
In sunnary, our investigation indicates that material reclassification has occurred, but it was done in accordance with approved procedures. Portions of ASME systems have been downgraded to non-Code; however, it was~not done in violation of Code or regulatory requirements. Safety Significance None Generic Implications on Other Systems or Contractors Not applicable +
- ~ , . , , . - -
,--n ..-
.- l- .
r TXX-4187 6/1/84 - Allegation No. 18 It has been alleged that there was improper sign off of " hold points" on travelers. .e - Evaluation of Validity We have assumed that the term " improper sign off" means the signitg of a traveler by someone unauthorized to do so. Our review of this matter indicates that there have been isolated instances o'f this nature. Wh'en anunauthorizedsignaturehasbeendetected,appropriagecorrective~ action has been taken. Safety Significance None Generic Implications on Other Systems or Contractors
, Not Applicable 9
Y n I e
TXX-4187 6/1/84 Allegation No. 19 j It has been alleged that craft personnel would s'atisfy a CMC on an k h ,>y g ofhcutting inadequate it out and then weld by welding over it instead of welding. Evaluation of Validity The term " inadequate weld" used in this allegation is vague. For example, a weld which was undersize and therefore " inadequate" would require no weld metal removal prior to performance of additional welding. On the other hand, welds may be termed " inadequate" due to relevant indications and rejectable by applicable codes or standards. In these cases, procedures require removal of the defect, verification of defect removal by the same nondestructive examination method which detected the defect, rewelding, and repetition of required NDE. In neither case would " inadequate welds" be welded over in violation of procedure. Without any specific information to aid in the evaluation, we are unable to address this allegation more specifically. Safety Sionificance None Impact on Other Systems or Contractors Not applicable t
-.- - . - - _. . _.__ . . . - - - - - - - - , _ . . _ . - . . _ , . . - - ~ , _ . - , - . - _ . . . _ - - -
e , i I TXX-4187 6/1/84 Allegation No. 20 - It has been alleged that there ar,e under.umented weld repairs.
;L('f t 52 Modifications were made to material, such as a hanger, after QC had
/s approved it.
i[b. Evaluation of Validity This allegation' lacks sufficient specificity to allow an evaluation. Our review has determined that construction has, on occasion, performed additional welding after QC acceptance of a pipe support. Unauthorized welding, when identified, has been resolved through approved nonconformance procedures. There are instances, however, where post < inspection modifications have been performed. This is, however, controlled procedurally and results in a subsequent QC inspection of the modification. Safety Significance ' None Generic Impact on Other Systems or fontractors ; Not applicable i k 4
1 f - STATE OF TEXAS ) i
)
COUNTY OF E0.'t:"J.'CLL - ) Hooo g . Billy R. Clements being duly sworn deposes and says: That he is Vice President, Nuclear Operations, Texas Utilities Generating Company and knows the contents of the foregoing Applicants' additional response to Darrell G. Eisenhut's April 24, 1984 letter transmitting allegations; that the same is true of his own knowledge except as to matters therein stated on information and belief, and as to that he believes them to be true. Billy R. C/lements
~
Subscribed and sworn to before me this 1 day of June 1984.
-:.. c:: .. m: . i e .~,,c h fgg . o r ::n 1: a 5~
Eva Anz - N6t'ary Public i State: Texas , . County: Hood ,, ?
, 0&V fO. .j
,gg f A 8/18/
fg UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of l 1 TEXAS UTILITIES GENERATING i Docket Nos. 50-445-1 COMPANY, et al. l and 50-446-1 i (Comanche Peak Steam Electric Station i Station, Units 1 and 2) 1 CASE'S PARTIAL ANSWER IN OPPOSITION TO APPLICANTS' MOTION FOR AUTHORIZATION TO ISSUE A LICENSE TO LOAD FUEL AND CONDUCT CERTAIN PRECRITICAL TESTING , AND MOTION FOR ADDITIONAL TIME TO RESPOND On August 7, 1984, Applicants filed their Motion for Authorization to Issue a License to Load Fuel and Conduct Certain Precritical Testing, pursuant to 10 CFR 50.57(c). CASE (Citizens Association for Sound Energy), the only remaining Intervenor herein, received that pleading on August 8, 1984 (which would make CASE's Answer due to be mailed on August 18, 1984). . We hereby submit CASE's Partial Answer in opposition to Applicants' Motion. Applicants argue (Motlon at page 2) that the Board shculd summarily grant this motion because "(1) the activities for which authorization is sought will not endanger public health and safety, and (2) the contention which is presently pending before this Board is not relevant to the proposed activities." However, both of Applicants' representations ire false, as CASE will demonstrate in the following. To begin with, by their own wording, Applicants have placed their Motion in the same category as their other recent Motions for Summary t F0lA-85-59 1 T 3 ##8 a ]
l e Disposition, and CASE submits that we should be accorded necessary discovery and additional time to adequately respond to this Motion. This will enable CASE to further research and develop the points made in this Partial Answer, allow us to include Affidavits where applicable, and is in keeping with the requirements of 10 CFR 50.57(c), under which Applicants filed their Motion. CASE moves that the Board grant us discovery on Applicants' Motion, and twenty days from the time we receive answers and documents requested on discovery, in which to respond to Applicants' Motion. Applicable NRC regulations under which Applicants have flied their Motion,'and by which they and the Licensing Board are bound, are discussed below. 10 CFR 50.57(c) states, in part:
"50.57 Issuance of operating license
-"(c) Action on such a motion by the presiding officer shall be taken with due regard to the rights of the parties to the proceedings, including the right of any party to be heard to the extent that his contentions are relevant to the activity to be authorized. Prior to taking any action on such a motion which any party opposes, the .
presiding officer shall make findings on the matters specified in paragraph (a) of this section as to which there 1% a controversy, in the form of an initial decision with respect to the contested activity sought to be authorized." 10 CFR 50.57(a) states, in part:
, "(a) Pursuant to 50.56, an operating license may be issued by the Commission, up to the full term authorized by 50.51, upon finding that:
"(1) Construction of the facility has been substantially -
completed, in conformity with the construction permit and the application as amended, the provisions of the Act, and the rules and regulations of the Commission; and
"(2) The facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission; and 2
L: _ _ _ _ _ _J
1 .
,6 . ,/
"(3) There is reasonable assurance (1) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations in this chapter; and
"(4) The applicant is technically and financially qualified to engage in the activities authorized by the operating license in accordance with the regulations in this chaptet. ..
"(5) The applicable provisions of Part 140 of this chapter have been satisfied; and -
"(6) The issuance of the license will not be inimical to the common defense and security or to the health and safety of the public."
Applicants attempt to persuade the Board that the Diablo Canyon case and the Comanche Peak case are so similar that the Board should rely on the decision in Diablo Canyon as a precedent for a similar ruling here. (Motion at pages 8 and 9.) However, Applicants have failed to make the necessary comparisons to support such a conclusion. The Diablo Canyon case is different from Comanche Peak in many regards. One of the most important dif ferences is that Diablo Canyon had already been granted an operating license once, which means that those Applicants had already proved their case to the satisfaction of the Licensing Board /1/. Certainly this is not the case with- Comanche Peak, where there are many very serious, hotly contested issues still to be resolved before the Board can consider whether or not to grant an operating license. In addition, the issues raised at Diablo Canyon prior to the decision cited by Applicants were considerably different from those in Comanche Peak. f1f Even though that license was suspended in November of 1981, within about a week of its having been granted, and even though the Commission acknowledged at that time that the plant should probably never have been granted a license to begin with. 3 1
r
,/..
CASE submits that, for these reasons, Diablo Canyon does not present an appropriate comparison to Comanche Peak, and that the decision in that proceeding should not be used as a precedent for Comanche Peak. ' Applicants also attempt to draw a comparison between Catawba and Comanche Peak (Motion at page 9). However, at Catawba there were several differences which render it an unacceptable comparison to Comanche Peak. - For example, in Catawba, the Intervenors (for their own reasons, which Applicants have not documented) did not attempt to make a case that there . were contentions which were relevant to the fuel load authority which was sought;'and the Intervenors did not oppose the license to load fuel. That obviously is not the case with Comanche Peak, where the Intervenor is strongly opposing fuel load and citing specific reasons for doing so. ,It is also inappropriate and unfounded for Applicants to draw the inferences which they have attempted to draw from the lack of opposition to fuel load in the Catawba case. At Catawba, the Intervenors did not have the number or type of issues which CASE is presenting in the Comanche Pedk hearings. (For example, they did not have the design issues which we have in our 1 i proceedings.) It is obvious that the decision in Catawba is not binding on the Board in this instance and is, in fact, totally irrelevant and l immaterial to any issue at hand in these proceedings. For these reasons, CASE submits that Catawba does not present an - appropriate comparison for Comanche Peak, and that the decision in that proceeding should not be used as a precedent for the Comanche Peak proceedings. 4 J
r
*. . ?-
Clearly, the decisions in Diablo Canyon and Catawba (the only two precedents cited by Applicants) were unusual decisions applicable only to those particular cases. It should be noted that there is no provision in NRC regulations for the specific type of no,-criticality testing Applicants propose. The provision under which Applicants are filing is 10 CFR 50.57(c) which states very clearly that it is "for an operating license authorizing low power testing (operation at not mor.e than 1 percent of full power operation), and further operations short of full power operation" (emphases added). This is quite.different from what Applicants propose. But if they are granted a license under this provision, CASE firmly believes that they will attempt to use it as a foot in the door to conduct further testing (either accidentally or " accidentally on purpose") at low criticality. Applicants argue that they do not seek a low-power license which would permit Comanche Peak Unit 1 to go critical, and that they " seek only authorization to load fuel and to conduct certain testing that must be completed before initial criticality may be achieved." /2/. Applicants . also state, however, that "The first two fuel assembikps loaded contain the neutron sources." (Motion at page 3.) Further, they do not claim that njt risk to the public health.and safety can occur. They claim that there will be no significant risk (Motion at page 7); however, there is no explanation as to what constitutes "significant risk" in the minds of Applicants. It should be assumed that Applicants do not believe operating the reactor at full power will pose a significant risk, based on their past representations j l f2/ Applicants also state that "(in fact it will be shut down by a margin of at least 5%)" (Motion at page 7.) There is no further explanation of precisely what is meant by this terminology, and CASE does not understar.d what is meant by it. This is one of the questions we would like to have answered on discovery. 5 l
.c to the Board (which have not been made for some time now) that there is sufficient evidence in the record for the Board to make a decision favorable to the Applicants' receiving an operating license.
In addition, the Commission's decision in Diablo Canyon, which Applicants cite as a precedent (Motion at top of page 8) does not clata that there is nji risk to the public health and safety from fuel loading and pre-criticality testing -- only that it is " extremely low since no self-sustaining nuclear chain reaction will take place under the terms of the license and therefore no radioactive fission products will be produced." (Motion at page 8; emphasis added.) CASE submits that implicit in the Commission's decision in Diablo Canyon is the assumption (we believe an erroneous one in the case of the . Comanche Peak Applicants) that Applicants will indeed comply with the terms of the license. However, Applicants' trustworthiness and compliance with the terms of their present (construction) permit and applicable NRC regulations goes to , the very heart of CASE's Contention 5, as the wording of our contention plainly indicates: -
"The Applicants' failure to adhere to the quality assurance /quallty control provisions required by the construction permits for Comanche Peak Units 1 and 2, and the requirements of Appendix B of 10 CFR Part 50, and the construction practices employed, specifically in regard to concrete work, mortar blocks, steel, fracture toughness testing, expansion joints, placement of the reactor vessel for Unit 2, welding, inspection and testing, materials used, 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 construction of the facility. As as result, the Commission cannot make the findings required by 10 CFR 50.57(a) necessary for issuance of an operating license for Comanche Peak."
-- (Emphases added.)
6
T Thus, contrary to Applicants' assertions, CASE's Contention 5 is directly concerned with whether or not Applicants can be relied upon to comply with their construction permits and applicable NRC regulations, and thus is directly applicable to the issues at hand. Furthermore, as discussed in more detail herein, the Applicants in the Comanche Peak proceedings, by their own actions and statements, have called Into serious question any assumption that they can be relied upon.to comply with the terms of a license to load fuel and do pre-criticality testing, or that they will or have the ability to keep the plant from reaching criticality. CASE submits that, in order to comply with the plain language of NRC regulations, the Licensing Board in the Comanche Peak proceedings, prior , to granting Applicants' current Motion, must make the same findings which it would be necessary to make for a full operating license. For the Board to make the decision now to allow Applicants to load fuel and engage in non-criticality testing would amount to a prejudgement on the part of the Board on virtually all of the important safety . issues which have been hotly contested over the past few years -- but without having all the facts necessary to make such a judgement. Were the Board to rule favorably on l l Applicants' Motion, it would be saying in effect that the Board members are I ready to state without reservation that Applicants are completely . trustworthy and that they believe everything Applicancs have told them, that l they do not believe what CASE's witnesses have told them,"that they believe the plant has been designed and constructed correctly, and that Applicants have proven their case. CASE submits that not only would this be patently unfair and extremely 7
r-
,- .i prejudicial to CASE, but it is in fact contrary to the NRC's own regulations as set forth in 10 CFR 50.57(c). Further, we do not believe that the Board can, based on what is in the record (and what is soon to be be in the record, such as additional CASE Answers to Motions for Summary Disposition),
make a favorable finding at this time. Were the Licensing Board to rule favorably on Applicants' Motion at this time -- without having all the facts -- it would in effect amount to a - prejudgement by the Board that Comanche Peak will be granted an operating license and that the possibility that it will not be granted is in actuality - non-existent. If this is true, what have we all been doing in these proceedings for the past five years? - Were the Board to rule favorably on Applicants' Motion at this time -- without having all the facts -- it would send a clear and unmistakable message to the whistleblowers who testified in the past and in the recent intimidation' depositions (and who may yet co.me forward), who have placed their future livelihoods in jeopardy and altered their lives forever, that this Licensing Board and the NRC had a complete lack of concern about their sacrifices or about the important issues they have raised. It would say to the world that the wording of 10 CFR 50.7 f 3/ is hollow and meaningless. f3/ 10 CFR 50.7 states, in part:
"(a) Discrimination by a Commission licensee, permittee, an applicant for a Commission license or permit, or a contr tor or subcontractor of a Commission licensee, permittee, or applicant against an employee for engaging in certain protected activities is prohibited.
i l
"(c) A violation of paragraph (a) of this section by a Commission I
licensee, permittee, an applicant for a Commission license or permit, or a contractor or subcontractor of a Commission licensee, permittee, or applicant may be grounds for: [
"(1) Denial, revocation, or suspension of the license. .."
(Emphasis added.) I ! 8 l
Were the Board to rule favorably on Applicants' Motion at this time - without having all the facts - it would send a clear message to the Applicants in the Comanche Peak proceedings that. they will not be punished for violating NRC regulations.but will be instead rewarded. In addition, it would send a clear message to Applicants in other proceedings that it really doesn't matter whether or not they comply with 10 CFR 50.7 or 10 CFR Part 50, Appendix B, or 10 CFR 50.57, they will still be allowed to load fuel and eventually get their operating licenses. Such a ruling at this point would have repercussions far beyond the Comanche Peak proceedings. Were the Board to rule favorably on Applicants' Motion at this time - without having all the f acts - it would send a clear message tohessrs. Walsh and Doyl who have sacrificed their time, noney, and effort for almost two years in order to bring facts to the Board (even af ter having been told that Applicants would be allowed to relitigate the design / design QA issues admittedly without having to show good cause), that their efforts
. have been unappreciated and in vain, that the Board is not really interested .
Inthedesign/designQAissues,andthachssrs.WalshandDoyhhavein actuality been participating in what amounts to a farce. I Were the Board to rule f avorably on Applicants' Motion at this time - without having all the facts - it would send a clear message to individuals l with vitally important information regarding problems with documentation (including falsification of documents) that the Board is not interested in finding out the facts about Applicants' important " paper trail," upon which the NRC is so dependent for its reasonable assurance that the plant has been constructed and designed properly. i l l 9 E _. _ _ . _
- s. ,n Were the Board to rule favorably on Applicants' Motion at this time --
without having all the facts -- it would send a clear message to CASE that the NRC intended CASE's participation in these proceedings for all these years to serve one purpose and one purpose only, to add credibility to a system designed not to arrive at the facts, but only to allay the concerns of the public about Comanche Peak. CASE does not and cannot believe that this was or is this Licensing Board's intent. The NRC is under close scrutiny already for other actions taken. In fact, Congressaan Udall's committee is holding hearings on August 30 about the very case cited by Applicants -- Diablo Canyon. In addition, the competence and credibility of the NRC has already been called into question and the public confidence in the NRC severely damaged recently in these very proceedings by the violation of confidentiality for whistleblowers by NRC Staff and counsel. This Licensing Board, in deciding whether or not to , grant Applicants' Motion to load fuel, has a clear choice to make in the Comanche Peak case -- a choice which will either increase and help to i
- restore the public's confidence in the Nuclear Regulatory Commission, or deal what may well prove to be a death blow to the public's trust.
Further, CASE submits that Applicants' arguments are not logical and that they have failed their burden of proof as proponents of the requested Order /4/. 14/ 10 CFR 2.732 (Burden of proof) states:
"Unless otherwise ordered by the presiding officer, the applicant or the proponent of an order has the burden of proof."
l 10
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Nowhere in their pleading have they explained exactly why it is necessary to load fuel in order to make the tests which they propose. If there is no need for the neutron sources, why is it necessary to load them into the reactor prior to tests? CASE submits that there is no, good reason for Applicants being allowed to load fuel prematurely. In fact, a brief re'riew of Applicants' recent biweekly updates of the schedule for fuel loading for Unit I reveals that Applicants have been increasingly deferring preoperational testing items until after fuel load /5,/. This was a willful management decision. Applicants have totally failed to meet their burden of proof in this regard, and CASE submits that they are using this as a deliberate ploy at this time to put additional pressure on the Licensing Board to rush to a favorable decision which the Board cannot possibly make based upon the information it now has. Applicants claim that Comanche Peak will not be allowed to become critical (Motion at page 7). However, CASE believes that it should be obvious to the Board, from'the Answers to Applicants' ,Mor. ions for Summary Disposition which we have filed to date and the numerous ([alsh/Doylj} allegations which Applicants have not yet even addressed, that there are still many unresolved questions regarding the design an Comanche Peak, sufficient to call into question Applicants' ability to assure that the plant will not be allowed to reach criticality. f5/ Compare, for example, Applicants' third biweekly status report (under 6/18/84 cover letter to Messrs. Eisenhut and Collins), page 2, to the fourth biweekly status report (under 7/5/84 cover letter to the Board), page 2, where Applicants have added yet another test (Control Room Air Balance) to their list of deferred preoperational testing items to be conducted after fuel load. See also discussions in CASE's 10/13/83 (1) Motion to Add a New Contention, (2) Motion for Discovery, and (3) Offer of Proof. 11
. . I p t In addition, there is another problem of which the Board is probably not aware which CASE believes should be considered before the Board makes its decision. This matter first came to CASE's attention on 8/9/84, when we received a copy of NRC Inspection and Enforcement (I&E) Report 50-445/84-08, 50-446/84-04 (copy attached). Like Applicants' Motion, our discussion here is premature. However, although we have not completed our evaluation of the probles at this point in time, we believe that the potential impact of the problem is so great that we must call it to the Board's attention now for
, the Board's consideration before making its decision.
In this I&E Report, Applicants were cited with a Notice of Violation (Appendix A of Report) for: 4
" Failure to Perform Inspections of Installation Activities Related to Unit 1, Main Coolant System Crossover Le2 Restraints
" Criterion X of Appendix B to 10 CFR Part 50 requires that inspections of activities affecting quality shall be established and executed by or for the organizations performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity.
" Texas Utilities Electric Company Quality Assurance Plan, in Section 10.0, requires that planned written inspection procedures be utilized.
It further requires that inspection activities include the types of characteristics to bd measured, the methods of examination, and the criteria. -
" Contrary to the above, it was determined that inspections were not made of the installations of the Unit 1 crossover leg restraints, nor were any documents requiring such an inspection issued. Specifically, the requirements for installation, as specified in Gibbs & Hill Drawing
, 2323-S1-0550, were not inspected and documented. The eight crossover restraints (2 per loop) are asior components of the main coolant piping seismic restraints and support system.
"This is a Severity Level IV Violation. (Supplement II.D) (445/8408-02)"
(First emphasis (title) in the original; remaining emphases added.) 4 I 12
This matter is discussed further on pages 9 and 10 of Appendix C to the Report. Although the total disregard for NRC regulations (and the Applicants' own requirements) for inspection and documentation of these vitally important restraints is, in and of itself, disturbing enough, there is another aspect not discussed in the I&E Report to which we now call the Board's attention. As CASE understands it, these restraints are rupture restraints, used for energy dissipation. They are passive under normal conditions (normal, upset, and emergency), but they are absolutely critical to the survivability of the plant in the event of a double guillotine break. (It should be noted that Applicants have made an assumption of where the pipe can break, but that does not necessarily mean that that is where it's going to break. There could also be a horizontal break at the nozzles of the steam generator or the recirculating pump, for instance.) If these cross-over leg restraints cannot take the load during such an event (i.e., if they fail), the effects of the double guillotine break are , transferred by couple into the upper and lower lateral. restraints for the steam generator. It is CASE's belief that these two cross-over leg restraints (for each loop) take a vertical component in one direction and a horizontal component in one direction, and that they are bi-directional supports, whereas they should be tri-directional supports (only restrain 2 degrees of freedom, whereas they should restrain 6 degrees)'. If the cross-over breaks, it would be similar to a jet engine (i.e., the stesa comes out like a jet); it causes the steam generator to, in effect, take a flip, but 13
0 the angle there stops it from doing that. If the cross-over restraints fall in their function, then the only restraint left for the steam generator (at least) are the upper and lower lateral restraints (which are already in question); and the loading into the effects has not been included into the analysis of the upper and lower lateral restraints because Applicants are relying on the ability of these cross-over restraints to dissipate their portion of the energy. One must consider that each element of the restraint system has got to contribute its own weight in the dissipation of energy in the event of an accident, and the failure of any one of these elements transfers an additional, unanticipated, and unanalyzed load to other parts of the system. Further, CASE questions whether or not the upper and lower lateral restraints have been inspected either. (To what criterion, to what requirement were they inspected? To the earlier criterion, which failed completely? Or to the latest, where both the upper and the lower lateral restraints were included?) . There will be further discussions hf ter Messrs. Walsh and Doy have had sufficient time to review the details of this matter) in CASE's Answer to Applicants' Motion for Summary Disposition regarding the upper lateral restraint. However, CASE submits that there is sufficient doubt due to what we know so far to raise serious doubts in the Board's mind regarding these important matters - at least to the extent that Applicants should not be allowed to load fuel until these questions are resolved. 14
e . Also included in I&E Report 50-445/84-08, 50-446/84-04 is a Notice of - Violation regarding the Unit 1 Polar Crane:
" Caps on Unit 1 Polar Crane Bracket and Seismic Connections Exceed Design Requirements "10 CFR 50, Appendix B, Criterion V requires that, ' activities affecting quality shall be prescribed by documented instruction, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.' .
" Design change Authorization 9872 required that all gaps on the Unit 1 polar crane bracket and seismic connections greater than 1/16 inch be shimmed.
" Contrary to the above on February 13, 1984, the NRC inspector reviewed the polar crane bracket and seismic connections listed below and observed that there were unshimmed gaps that exceeded 1/16 inch."
(8 girders were listed, three of which had gaps of approximately 1/8", three with gaps 3/16", and two with gaps 5/32"; see copy of report attached.)
"This is a Severity Level IV Violation (Supplement 11.D) (445/8408-01)." (First emphasis (title) in the original; remaining emphases added.)
On page 4 (last paragraph) of Appendix C to the Report, the Inspector states that these were randomly selected girder connections, which , would seem to indicate that they may not be all of the girders with gaps greater than 1/16 inch. . This matter is laportant because the NRC Inspector discovered this violation while checking out Licensee Action on Previous Inspection Findings (see discussion on page 4, last paragraph, of Appendix C to the I&E Report). The previous inspection finding which was checked out and closed was in regard to a previous violation (445/82-11): Failure to Perfore Inspections of Installation Activities Related to Unit 1 Containment Polar Crane. This is the same matter discussed in the Board's 7/29/83 Proposed Initial Decision (Concerning aspects of construction quality control, emergency planning and Board questions) at page 19. The Board stated: 15
". . . the Board does not believe that it is a matter which the Board should pursue sua sponte because it appears that the staff and the applicant are addresing it. The staff issued a Notice of Violation in connection with the failure to inspect these shias /76/. The applicant has stated that all the shims in the polar crane girder support bracket assemblies will be removed and inspected. Shias which have clipped
' fingers' will be evaluated by an engineer to determine whether they are acceptable /77/.
"/76/ Staff Ex. 1488.
"/77/ Testimony of John T. Merritt, Jr. Regarding Placement of Shias -
in Polar Crane Glider (sic) Support Bracket Assemblies, App. Ex. 127, at 6." However, as stated in I&E Report 84-08/84-04 (page 4 of Appendix C): -
"The Licensee's Nonconformance Report (NCR) M-82-00894 documents the above violation. The disposition of NCR M-82-00894 directed that the polar crane girder connection finger shias previously installed per DCA 9872 were to be removed and inspected and any deviations from the requirements of DCA 9872 were to be identified to engineering for resolution. . .
"The shim inspection and rework was inspected and documented by quality control (QC) inspectors on NCR M-82-00894. This NCR was closed on January 24, 1983. The quality control inspection of the shia rework satisfies the requirements which were previously not met and which resulted in the original violation. This item is closed. However, the NRC inspector performed a random inspection of the polar crane girder connection shims and had the following two concerns: .
"(1) Design Change Authorization (DCA) 9872 required that all gaps greater than 1/16 inch be shimmed. In addition QC personnel I
verified that the gap for each polar crane girder connection was less than 1/16 inch and documented this on a shim documentation card which was attached to Traveler CE-82-370-8104 However, the NRC inspector observed that the following randomly selected girder connections had gaps that exceeded 1/16 inch:" (list of eight girders is given, with three have gaps of approximately 1/8", three with 3/16", and two with 5/32".)
"This is an apparent violation (445/8408-01).
l "(2) DCA 9872 required that the shias be tack welded as shown on Gibbs 1 and Hill sketch SK82032 (Sheet 3 of DCA 9872). General Note 4 of I SK82032 states that shims in the seismic connection may be welded to either vertical plate; however, on the seismic connections for girders 18, 22, 26, and 27 the tack welds which welded the shias together also tack welded the vertical plates together. This concern was discussed with licensee personnel. This is an unresolved ites (445/8408-04)." (Esphases added.) 16
- - .-.n. - - , y..-. ~. _ , - - - . - . - - . _ -
Obviously Applicants' corrective action was not adequate. There is yet another reason why this matter is noteworthy. In responding to a reporter's questions about this violation, a TUGC0 engineer stated (see attached copy of 8/14/84 DALLAS MORNING NEWS article):
"The most serious violation concerned earthquake supports for a heavy crane above the reactor core that previously had been approved by TUGC0 plant inspectors but later rejected by KRC inspectors, officials said. .
TUCCO engineer Tom Rose said that plant engineers believe that the TUGC0 inspections had been performed correctly, but that changes in temperature had caused the metal in the earthquake supports to expand .
and contract, resulting in the violations cited." (Emphasis added.) If the statement made by Mr. Rose is correct, and this expansion and , contraction came about at ambient temperatures, this obviously raises serious questions about what algbt happen under LOCA conditions, which was not discussed in the article but which is obviously a concern which must be . addressed. Applicants claim that Comanche Peak will not be allowed to become , critical (Motion at page 7). As stated by Applicants (Nbtion at page 7):
". . . the public health and safety can he at risk from nuclear power reactor activities only when fission products can be released to the
{ environment. Fission products are the by-products of the fission l process which occurs in the core after criticality." Applicants further argue that " Critical operation at significant power levels is required to generate enough fission products to be hazardous." . (Emphasis added.) However, once Applicants put in the high neutron source, l l they have a hot plant, whether it is operated at 1% or higher. Further, the l; I possibility of an accident would exist even then, although obviously there would not be anywhere near the reactivity of a plant being operated at 100%. 17 I I l l '_. ~
But once the plant is operated at 1% for a short time, there is low grade spent fuel. Once Applicants are allowed to go to low power, the fuel is now self-energizing and the neutron source is no longer required. All we would be talking about at that point is degree of reactivity; it would no longer be a question of whether or it there is spent fuel, but just the degree of degradation. There will be transuranics and other elements; it is just a matter of the degree. Applicants should be asked if they can return the spent fuel to the stockpile after, say, a month. CASE challenges Applicants' top officials to reach in there and pull a rod out with their bare hands af ter operating just a week at 1%. It'should also be noted that 1%. = about 13 MW electric.= about 40 MW thermal = the total capacity of a nuclear submarine reactor (which under full operation, normally never even gets to 13 MW). Further, 1% is equal to about 10% of the capacity of the FFTF = about 40 MW thermal. Once fission products have been produced (even at only 1% power) and decay heat has been produced, it is necessary for all safety-related systems - to be functional, due to the increasing level of fission products and decay heat. As discussed hereinF and in other recent CASE pleadings, there is strong question at this point in time as to whether or not all safety-related systems would indeed be functional. For the reasons discussed in this pleadir.3, CASE questions whether Applicants have either the desire or the ability to assure that the plant will not be allowed to become critical. Further, once the plant went critical, CASE questions the ability of Applicants, and the reliability of the safety-related systems in the plant, to keep it under control. 18
It is CASE's understanding that, in order to assure that the plant does not go critical, the neutron detectors must be fully operable, since they are the means used to measure the neutron flux. However, theta have been problema identified regarding the neutron instrument detector slota and wells which call into question the wisdom of allowing Applicants to proceed . with fuel load and testing without having further information in the record:
-(1) In the 7/6/77 letter from R. E. Hersperger, Project Macager, Gibbs
& Hill, to-H. C. Schmidt, Project Manager - Nuclear Plants. TUCCO (attached to CASE Exhibit 479, NCR C-669, admitted into evidence in accordance with the Board's Order (Proposed Findings of Fact; CASE Exhibits) of December 7, 1982, and admitted into evidence in the May 1983 hearings), Gibb's & Rill stated d
". . . G&H was notified that a series of rebsrs had been omitted from the reactor cavity concrete between Elevations 812'-0" and 819'-0-l/2".
The missing rebars were located adjacent to the neutton degection slots and had been added only xecently as a change in G4R drawings 2322-SI-0572, 2323-SI-0574 and 2323-SI-0575. . . the omission of this additional reinforcement does not in any way impair the structural integrity of the reactor primary shield structura under any postulaced , loading condition. The additional rebar had been added by G&H as a precaution against cracking which might possibly_ eccur in the vicinitt of the neutron detector slots following a LOCA. They provide a maana of uniformly distributing accident loading stressis around the slots precluding the possibility.of local cracking.
". . . These bars were added to the design because in the judgement of the design engineer they provided a. prudent improvement in the 7
performance of the reactor cavity structure." (Esphases added.) , The missing bars were never installed. (2) Inspection and Enforcement Report 50-445/83-34, 50-446/83-18, which was forwarded to.the Board by the NRC Staff Counsel under cover letter of September 26, 1983, stated (page 5): i 19
. )
i i
"The SRIO (Senior Resident Inspector-operatio'ns) reviewed the HFT (het functional test) Icg for any notation on.the shield wall / reactor vessel
?
interface. There were no entries related to this specific subject in 1 the log. There vae a notation that PT-45-06 ' Containment Ventilation * , failed to meet its acceptance critaris because_ the following areas more too hot (thermally {: '
"(1) All vessel supporta
"(2) Nnutron Instrument Detector Wells
"(3) Pressariser Room
- 4
"(4) All Steam Generitor Chapartments" (Emphases added.)
This problem was alto di'abussed la e'sgsts to Applicants' Containment i Tatyarature Survey (during hoc functional cesting) f 6/. Engineering escluation wae required p'rlor La ratesting section 7.1 containment , temper 4tura survey, which was t(ste& a total cf 3 tiaca. Upon comp!stion on , 3f 27/83 it was discovered chec the tesvarature indicators _were unreliabh in emb.ient truperatura above 134 degreen, test Deficiency 839 was 1seued - and section 7.1 uds requir64 to be retasted. Actest for 7DR-839 was completed on 4/5/S1 and T0a-908 tesued to idsocify arens that did not imet * ' acceptance criteria. As stated in TDR 908:
"The following areas dad not meet the acceptanc.s crtieria:
"1. A.?.V. (henctur Pressure Vessel) supports #1 Woc Leg. #2 Cold 34g, #3 Hot Leg, #4 Cold.
4 "2. Detector wells" (3) *
"3. PER Roon 905 *
"4. Steam Genecqtor Compartmentu #1, #2, #3, #4 f6/ See CASE Exdibit 857, attached to CASC's 10/13/83 (1) Motion to Add a New Contention, (2) Motion for Discovery, and (3) Offer of Proof; and discuselon on pages 35 and 36 of out 10/13/83 Mocien).
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"See CPPA-29.488 ' Minutes of Meeting' on April 20-21, 1983 (attached) for corrective Letions. K. F. Mcdonald 5/3/84" (sots: The Mlautes of the April 20-21 meeting were not attached to CASE,'s copy of TDR 908 when we received it from Applicants, and since CASE's Rocion was denied, we were not able to get it on discovery in the operating licensa hearings.)
Upon completien of engineering evaluation the Containment Ventilation System was adjusted. Due to Hot Functional Testing the changes in pressuriser area could not be. initiated. Section 7.1 was ratested on 5/18/83 and ,the same areas addressed in TDR-908 still did not meet this acceptance criteria. TDR-908 was closed and TDR-1221 issued against section 7.1. There were two test deficiencies which remained open at the conclusion (report'vas dated 6/1/83).of this test; one of them was: f
" Test de_ficiency report #1221 issued against Section 7.1 Containment Survey. Various areas in section 7.1 did not meet their respective
~
acceptance criteria. This area will be tested during initial startup. (Esphases added.) In addition, regarding the Control Room Heating & Ventilation System Performance test f 7/, the following was stated: ' " Tee status of Control Room HVAC System at the end
- of the preoperational test is as follows:
"Systse functions per design. Temporary modifications 124-127 were
.left installed because smoke detectors are not functional. Temporary modifications 130-133 were left installed because radiation monitors are not _ functional, chlorine detectors were lef t in the OFF position because they had not been calibrated, as of this date.
"The System Test Engineer recommends approval of the test as performed." .
i Dated 6/9/83. (Enphases added.) 17/ See CASE Exhibit 359, attached to CASE's 10/13/83 (1) Motion to Add a , New Contentien. (2) Motion for Discovery, and (3) Offer of Proof; and discussion on page 37 of our 10/13/83 Motion). 9 21
_ - _ __ _~. _ _ _ _ . _ _ C
. . . 'o -
If Applicants are to be allowed to load fuel and engage in testing, it is absolutely imperative that the neutron detectors be capable of fully performing their intended function. The combination of the preceding two identified and documented problems calls into question the. wisdom of allowing Applicants to proceed with fuel load and testing without having further information in the record. CASE opposes Applicants' being able to load fuel. However, if the Board disagrees in this regard, we urge that the Board, at a minious, require Applicants to provide sufficient information and documentation to resolve the concerns raised about the neutron detector wells and slots. t In this pleading and in recent CASE Answers to Motions for Summary , Disposition and other pleadings /8,/, the Board has been presented with j information and documentation which calls into serious question the adequacy i
/8/ See, for example:
CASE's 8/6/84 Answer to Applicants' Motion for Summary Disposition , of Certain CASE Allegations Regarding AWS and ASME Code Provisions Related to Design Issues (especially those pages identified on page 2 of cover letter); CASE's 8/6/84 Answer to Applicants' Motion for Summary Disposition Regarding Alleged Errors Made in Determining Damping Factors for OBE and SSE Loading Conditions (especially those pages identified on page 2 of cover letter); CASE's 8/6/84 Answer to Applicants' Motion for Summary Disposition Regarding Consideration of Friction Forces in the Design of Pipe Supports with Small Thermal Movements (especially those pages identified on page 2 of cover letter); . CASE's 8/13/84 Answer to Applicants' Motion for Sdmaary Disposition Regarding CASE Allegations Regarding Section Property Values (especially those pages identified on page 2 of cover letter); CASE's 8/13/84 Answer to Applicants' Notion for Summary Disposition Regarding the Ef fects of Caps on Structural Behavior Under Seismic Loading Conditions (especially those pages identified on page 2 of cover letter); i CASE's 8/14/84 Motions Regarding ANI Documents. See also CASE's 8/22/83 Proposed Findings of Fact and Conclusions of
' Law (Walsh/DoyleAllegations),entiredocument.
, 22
and effectiveness of Applicants' QA/QC program, the competence of Applicants with regard to the design of Comanche Peak, and the credibility of Applicants and their witnesses; further, the information and documentation presented strongly challenge the adequacy and intent of Applicants' management of the design and construction process at Comanche Peak. In addition, there are still many outstanding issues which CASE fully expects will continue to substantiate CASE's position, and which must be addressed before Applicants are allowed even to load fuel and perform tests. For one thing, all of the issues previously identified have not yet been resolved; for example, consider the following items listed in the Board's 3/15/84 Memorandum (Clarification of Open Issues): Intimidation The issue of intimidation, harassment, threatening, and firing of QC inspectors and others at Comanche Peak is a very important one, since a finding by the Board against Applicants could, in and of itself, be grounds for denial of an operating license, and/or revocation or suspension of Applicants' construction permit (see 10 CFR 50.7 and footnote on page 8 of this pleading). As CASE has discussed i j previously, we believe that intimidation (including a discouragement from doing the job right to begin with) is rampant at Comanche Peak.-- that it is, in fact, a way of life at the plant which is so ingrained that the quality of construction and design is indeterminate at best and deficient at worst.
- D. Intimidation in the Protective Coatings Area (page 7 of the Board's l
Memorandum). See item V. following. 23 l l
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F. Dismissal of bert Hamilton (page 8 . Already decided. See item V. following. J. Termination of (kenry Stiner (page 9jf See item V. following. V. Intimidation of QC Inspectors (pages 13 and 14). Extensive Depositions have already been taken; Expected Findings are due 8/31/8'; hearings begin 9/10/84. HH. Intimidation of(hrs. Stiner (page 18)h See item V. preceding. MM. Office of Investigations Reports (page 19). Protective Coatings As the Board is aware (although Applicants have not officially presented their position to the Board), Applicants are in the process of attempting to convince the NRC Staff that it would be all right if all the protective coatings inside the containment fell off the walls, because it would still not stop up the sump pumps /9/. However, we want,to call to the Board's attention that recently - the North Anna nuclear plant was shut down because the utility could I not provide the NRC w'ith documentation to prove that protective I coatings on the reactor's ventilation duct met NRC standards. (See I l
/9/ See Transcript of 6/7/84 and 7/27/84 NRC/ Applicants meetings in Bethesda; especially Tr. 85/3-13 of 7/27/84 meeting. .
See also 5/18/84, 5/23/84 (2 letters), letters to TUCCO from Richard L. l Bangart, Director, Region IV Comanche Peak Task Force, NRC, Region IV, Arlington, Texas; 5/22/84 Board Notification 84-106, Interim Report on Protective Coatings; 7/27/84 notice of meeting to be held 8/8/84 (which was subsequently cancelled with the understanding that Applicants would answer the questions attached in writing). I 24 l
t
'o, attached copy of August 8, 1984, WALL STREET JOURNAL article /10/.) As stated in the article:
"The problem was discovered when Unit I was closed recently for routine maintenance and inspection. Investigators could find no record indicating that paint on the reactor's ventilation duct met Nuclear Regulatory Commission standards . . . A paint that doesn't chip is esscacial because, during an accident, a flaking under intense heat could hinder safety operations.
"No records on paint could be found for Unit 2 and it was closed.last
~
week . . . +
"The utility is now wrapping mesh wire around the ducts to trap any paint that might chip under heat stress. The interim solution was approved by the NRC . . . "
i It should also be noted that there appears to have been no discussion to date about another aspect of the failure of protective coatings to perform their intended function -- that of helping with the ease of clean-up in the event of a nuclear accident and decreasing the amount of radioactivity to which members of. the clean-up crew would be subjected. The possibility of workers being needlessly exposed to increased radiation risk should the' plant go critical and have an - accident during Applicants' proposed fuel load and testing is one which we believe the Board must deal before making its decision on fuel loading. (E. Protective Coatings Technical Issues (page 7). It should be noted that CASE has decided to drop attempting to answer Applicants' pleading on l the issue of maximum roughness. We do not have any expert witnesses to
- /10/ We realize that newspaper articles are not the best documentation; however, we just received this article from one of our members yesterday and have not had time yet to obtain more information through the NRC public document room (if the information is even available there yet).
i 25 f
testify, and would have had to rely upon admissions from Applicants or the NRC Staff. Th'is does not mean, however, that we concede that this is not an important issue, or that Applicants' response is adequate or acceptable -- but only that CASE does not have the means available at this time to do anything about it within the context of the hearings process, through which we have always tried to work if at all possible. We believe that eventually the truth of the matter will come out, but in this instance apparently without CASE's help in the hearings.) H. Inadequate Disposition of Paint Defect Repairs (page 8). KK. Protective Coatings-(page 19). I. Undocumented Removal of Cable Trays (page 9). K. Welding Issues (page 9). Proposed Findings are due 8/31/84. Q. Heat Input for Welding (page 11).
- See also CASE's 8/14/84 Motions Regarding ANI Documents, especially pages 2 through 6 of Motion, and summary section on ANI REPORTS -- WELDING.
/3,/, page 2 -- Memorandum (Brandt Interpretation of Stiner Testimony),
February 10, 1984 (unpublished). Information relevant and material to this issue is contained in the same I&E Report discussed previously in this pleading, I&E Report 26
~J ,
84-08/84-04 (Appendix B, Notice of Deviation, and Appendix C, page 9, item 11.a. Platforms Inside Containment), which is self-explanatory. CASE Exhibits 1,033 and 1,051 -- major generic problem -- nonconforming material used on attachmenta for Class I attachments. CASE Exhibit 1,052 -- Class 1 piping attachment material installed in the field; Class 2 pressure retaining material after installation in Class 1 fabrication. CASE Exhibit 1,056 -- NCR's., used to upgrade supports from Class 2 to Class 1, possibly with nonconforming material. R. Unqualified QA/0C Supervisory Personnel (page 12). Vendor Surveillance
- S. QC Surveillanca of Chicago Bridge and Iron (page 12).
T. NPSI's Adequacy to Fabricate Pipe Restraints (page 13). Use of Polar Crane to Force 32" Main Steam Pipe into Positicn __ W. haserlyAllegations(page15]. BB. Cold Soringing of Pipe (page 16). Information which is relevant and material to these matters (auf ficient that CASE believes the record should be reopened, at least 27
, ~kC-e to the extent of allowing them to be included in Proposed Findings) is contained in CASE Exhibit 1,054 (attached to CASE's 8/14/84 Motions Regarding ANI Documents). As discussed in the summary section on ANI REPORTS - PROMPT IDENTIFICATION AND CORRECTION OF NONCONFORMANCES (page 9), this ANI Report documents a major problem: use of applied 7 force during fabrication of component supports (unauthorized use of a porta-power to spread the horizontal members of a box support in order to achieve required clearance, a practice which is not acceptable to the ANI, but which was closed by the'ANI on 5/16/84 " based on PSE Chief Engineer Jay Ryan assuming responsibility"). This ties in with and adds credibility to CASE Witness Bob Messerly's deposition regarding the use of the polar crane to force a 32" main steam line into position
/11/;alsowithtestimonybyCASEWitness([harlesAtchisoh that he had observed " cold springing" of two lines from reactor coolant pump compartment number three /12.
Document Control; Prompt Identification and Corrrection of Nonconforming Condition In addition to 'the following specific items mentioned in the Board's 3/15/84 Memorandum, CASE still plans to file a Motion for a New Contention regarding the subject of document control (including
/11/ See pages 25-32 of 4/14/84 Messerly NRC Deposition, attached to CASE's 8/3/83 letter to Board under Subject of Record Regarding Discouragement from Reporting Nonconforming Conditions at Comanche Peak Nuclear Plant.
/12/ See discussion on page 46 of Board's 7/29/83 Proposed Initial Decision (Concerning aspects of construction quality control, emergency planning and Board questions); closed with Board's 9/23/83 Memorandum and Order (Emergency Planning, Specific Quality Assurance Issues and Board Issues), page 36.
28
falsification of documentation), failure to promptly identify and correct nonconforming conditions, inadequacy of procedures, and related matters (including documentation problems with the important N-5 1 prograa /13/, lack of and inability to find adequate documentation regarding the fuel pool liner and transfer canal). This information calls into question the entire " paper trail" on which the NRC is so i dependent to assure that the plant has been constructed and designed properly and can operate without endangering the public health and safety. Earlier this week, we received approval from Mr. Roissan to obtain an'affidavitdirectfromffbieHatlejfregardingthereasonsfor withdrawingherFebruary1984 testimony;however,{jg.Hatlefhasbeen out of town ever since that time, and we do not yet have the necessary affidavit. In addition, the mini public document room does not yet have all the transcripts (of depositions taken during the Intimidation portion of the hearings) to which we need to refer in making our . Motion. We are hopeful that we will have all thE necessary documents in hand sometime during the coming week, and will file our Motion just I as soon as possible. X. Component Modification Cards (pages 15 and 16). FF. Computerization of Non-Conformance Reports (pages 17 and 18).
/3,/, page 2 -- Memorandum (Records Retrieval), LBP-84-8, 19 NRC , January 30, 1984.
/13/ See Testimony of NRC Staff Witness Robert Taylor during October 1983 hearings, Tr. 8875/8-8877/24 29
/
AA. Number of Inspectors (page 16). EE. Reactor Vessel Mirror Shield (page 17). CASE is not certain whether or not the Board is satisfied with the state of the record on this matter. II. Staff Walkdown Inspections (page 19). LL. Recent Changes in QA/QC Program (page 19). MM. Office of Investigations Reports (page 19). CASE believes that the Byron decision mandates that the Board not close the record until all 01 Reports on outstanding itema applicable to the hearings have been received and an opportunity provided for rebuttal to such reports. PP.$)halsh/DoyleAllegationa)(nowalsoreferredtoasDesignDecision , allegations) (page 20.). , JJ. Cygna Report (page 19).
/3,/, page 2 -- Partial Initial Decision (A-500 Steel), LBP 83-63, 18 NRC
, October 6, 1983. Being treated as a Motion for Summary Disposition.
ThedesignanddesignQAproblemsidentifiedby[kessrs.Walshand Doylh((includingthoseidentifiedregardingtheCygnaReport)become even more important because it appears that everyone is relying on Applicants to be certain that Comanche Peak has been designed correctly and adequately. The NRC's routine inspections are certainly not sufficient; and their Special Inspection Team (SIT), which looked into 30
e theallegationsof[Eessrs.WalshandDoylkfeventhendidnot adequately identify or address some of the problems. In addition, the Authorized Nuclear Inspectors (ANI's) do not inspect for design problems, but only for fabrication and installation /14/. This adds increased importance to the problem identified by Messrs. Walsh and Doyle, and since design problems are not inspected by QC Inspectors, or by ANI's, or routinely by the NRC inspectors, it is reasonable to assume that the design problems identified by Messrs. Walsh and Doyle are not isolated incidents, but are in fact only a small tip of the iceberg. Applicants' current Motion is based on some basic underlying assumptions -- that the plant has been constructed and designed and will operate in a predictable manner, and that there is a sufficient factor of safety to take care of any problems. However, it should be noted that CASE has not yet answered Applicants' Motion for Summary Disposition on safety, factors, and any decision by the Board at this - point in time on Applicants' current Motion would be reached without having seen CASE's response on the issue of safety factors. It should also be noted that, were the Board to rule based on its original 12/28/83 Board Memorandum and Order (Quality Assurance for Design), Applicants could not be granted an operating license at all (which would make the question of whether or not to grant a license to load fuel moot).
/14/ See Testimony of NRC Staff Witness Robert Taylor in the October 1983 hearings, Tr. 8875/8-8877/12.
31
T CASE believes that the Board will find little to resolve its concerns about the design / design OA issues in CASE's Answers to Applicants' Motions to Summary Disposition (some of which we have already sent in and some of which we are presently working on). In fact, in many instances in their Motions for Summary Disposition, Applicantshavebroughtnewinformationtotheattentionof[hessrs. WalshandDoylfwhichincreasesorsubstantiatesthoseconcerns. It should be obvious to the Board, from the Answers to Applicat.:s' Motions for Summary Disposition which we have filed to date and the numerouskaish/DoyljfallegationswhichApplicantshavenotyeteven addressed, that there are still many unresolved questions regarding the design and design OA at Comanche Peak -- not only about pipe supports, but about the entire plant -- some of which raise questions of sufficient significance that Applicants should not be granted a license even to load fuel and perform testing until those questions are resolved in Applicants' favor. CASE does not believe such resolution . is possible, based upon what is already in the record and what will soon be sent in. If there is no confidence in the design and design OA, there can be no reasonable assurance that Applicants can control precriticality, and they cannot be permitted to load fuel and perform tests without such assurance. RR. Trends or Patterns of Non-conforming Conditions Due to the heavy workload CASE has encountered, we have not as yet been able to complete our work on trends or patterns of non-conforming conditions. As we have discussed previously, this is a mammoth l l , 32 l
~
undertaking and we are still reviewing recently acquired information which is applicable to this. However, a rough sample of the type of information we would include in such trending is contained in CASE's 8/14/84 Motions Regarding ANI Documents, in the attached summaries. From a review of those relatively few documents, the amount of time and effort necessary can be readily understood.
/3/, page 2 -- Memorandum (Adequacy of Record: Delevel Diesel Generators),
January 31, 1984 (unpublished). In conclusion, as discussed herein, CASE has demonstrated (contrary to Applicants' argument) that there is a very good posIsibility that the activities for which auth'orization is sought will endanger public health and safety, and that CASE's Contention 5 is directly relevant to the proposed activities. As discussed herein, there is documentation and evidence to call into serious question Applicants' trustworthiness and compliance with the terms of their present (construction) permit and applicable NRC regulations. This goes to the very heart of CASE's Contention 5, as the wording of our contention plainly indicates, and is therefore directly relevant and material to Applicants' current Motion. Furthermore, Applicants in the Comanche Peak proceedings, by their own actions and statements, have called into serious q'uestion any assumption that they can be relied upon to comply with the terms of a license to load fuel and do pre-criticality testing, or that they will or have the ability to keep the plant from reaching criticality. 33
I
,e It should also be noted that Applicants' Motion is premature and that by their own admission (Footnote 1, page 3) they are already (apparently irretrievably) three weeks behind what CASE believes is their overly optimistic schedule to fuel load. There is no good reason (and none is offered by Appl 1~ cants) for their having to load fuel in order to conduct many of the tests, which they have by management decision deliberately and unnecessarily postponed until after fuel load.
In any event, CASE submits that, in order to comply with the plain language of NRC regulations, the Licensing Board in the Comanche Peak proceedings, prior to granting Applicants' current Motion, must make the same findings which it wouid be necessary to make for a full operating license. For the Board to make such findings now -- without having all the
.f acts necessary to make such a judgement -- would not only be patently unfair and extremely prejudicial to CASE, but is in fact contrary to the NRC's own regulations as set forth in 10 CFR 50.57(c). Further, for the reasons discussed herein, CASE does not believe that the Board can, ba' sed on .
what is in the record (and what is soon to be in the record), make a favorable ruling at this time. For these reasons and the other reasons discussed herein, CASE moves that the Board deny Applicants' Motion in its entirety. Raspectfully submitted,
&_M //L fMis.) Juanita Ellis, President i CASE (Citizens Association for Sound M Energy')
1 26 S. Polk Dallas, Texas 75224]P b- 214/946-9446 34
N RNMENT ACCOUNTABILITY PROJECT /
' 1555 Connecticut Awnue. N.W 202 '
Wo5Nngton, D.C. N (202)232-8550 September 26, 1984 RECElVED 00T 011984 JOHN .W. BECK Mr. Darrell G. Eisenhut Director Division of Licensing Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Eisenhut:
Recently, you provided to M. D. Spence, President of Texas Utilities Electric Company (TUEC), a request for further information regarding several items under review by the Technical Review Team (TRT) that have potential safety significance. It is my understanding that these items were also dis-cussed with TUEC at a public meeting in Bethesda last week. After reviewing your letter of September 13, 1984, and the transcript of the September 18 meeting, it is apparent that there are vital ianswered questions about both the methodology by which tne TRT is resolving issues that are brought to their attention by allegers, and the adequacy of the 'epth to which the TRT is looking at speci c issues. Your letter states that further background information regarding the issues identified in the September 13 letter and, presumably, the other items being reviewed by the TRT, will be published in a Supplement to a Safety Evaluation Report (SSER) which "will document the overall TRT's assessment of the significance of the issues examined." (September 13, 1984 letter from D. Eisenhut to M. D. Spence.) ' The Government Accountt 'lity Project (GAP) believes tha- any assess-ment of the overall significance 1ould be publicly disclosed pric to the issuance of the SSER. Frankly, we think that there is significant evidehce available to our investigators that the TRT, other members of your staff, and Region IV inspectors have fallen short of the in-depth inquiry which would provide the basis o ad uacy of any overa s A f/' / . qy
- f' . ap ,/P,,
/
,,3 y I
. * . .. ~ .
9 +
. Mr. Darrell G. Eisenhut
. September 26, 1984 Page Two Further, we believe that your request from TUEC fer additional infor-mation through "a program and schedule for completing a detailed and thorough assessment of the issues identified" falls short of the obviously necessary requirement 'for comprehensive programs to identify the full scope of the technical deficiencies listed in the letter. We note that in your request that a program pl.gn "should address the root cause of each problem identifled and its generic implication on safety-related systems, programs or areas," as well as the collective significance of these deficiencies." However, you have not required that TUEC's proposed response must first include ar identification of the extent of the identified problem. This " backward look" is critical to any
~
meaningful assessment of the adequacy of a TUEC "get well" program. Further, we do not understand the instruction to TPEC to submit a response which ..addresses those items .l.is.ted in.yo.ur .Septe.m.b.e.r 13., 1984, letter in the face of the large volume of outstanding items, yet to he identified, and the findings of the Quality Assurance / Quality Control (CA/QC) team. It appears to us that your direction to TUEC was vague and inappropriate, at ! this time. it neither ordered a reinspection commensurate with the level of deficiencies discovered, nor simply informed TUEC of some of the TRT's findings. i We r e that all of the identified p. alems in this letter were first , { identified by former employees at the Comanche Peak site, yet in no case that we are aware of, has the TRT re-contacted that employee with its findings to i get further direction or guidance, or clarifying information regarding the specific confirmed item. A good example of this is the finding under Test Y ( Program Areas; Prerequisite Testing (Section III(c).) Your request for addi-tional information stated that after a review of the Final Safety Analysis { Report (FSAPI commitments, appropriate procedures, records, and interviews, the I following p. lem was identified: The review of test records revealed that craft personnel M ' C-l : vere signing to verify initial conditions for tests in ' / O< violation of startup Administrative Procedure-21, en- 1 14M titled: " Conduct of Testing" (CP-SAP-21). This procedure requires this function to be performed by System Test ' d'/ Engineers (STE). Startup management had issued a memoran-4' 4 i [e dum improperly authorizing craft personnel to perform these verifications on selected tests. I
, r !
l t
Mr. Darrell G. Eisenhut
, September 26, 1984 Page Three Your letter than apparently instt ucts TUEC to: (1) rescind the startup memorandum (STM-83084); and (2) ensure that no other memoranda are issued which are in conflict with approved procedures.
After going over this finding with the allegers, we discovered several things which your report did not discover and does not address. First, the memoraEdum (STM-83084) which the TRT has identified, was not the beginning of the problem which permitted unqualified personnel to perform prerequisite tests. The memorandum, in fact, placed 'a limitation on the tests which un-qualified craft personnel were allowed to perform. Prior to the issuance of the mem6randum, there had been no, controls on ,the craft personnel performing , verification tests. A more detailed explanation has been provided to the TRT on"e, and a random review of prerequi.ite tests performed prior to the issu-ance of STM-83084 would have demonstrated that prerequisite tests were performed almost exclusively by~ unqualified personnel. -- - / -- Your requirement for TUEC to rescind the memorandum in conflict with CP-SAP.-21 is totally. inadequate to determine the consequences of an unknown majority of prerequisite testing done by.cnqualified, people. _ v Another example of 4.he inadequate approach being taken by the TRT and other branches of your agency is the handling of problems identified in the electrical / instrumentation area. Your September 13 Request for Additional Informatio'n identifies five l problems with electrical cable terminations, butt splices inside panels and vendor-installed terminal lugs. It also requires"certain procedural or re-inspection requirements for each of the problems. ! A cursory review af the corrective actions indicates that those
^
actions are totally inade; :e. For example, the TRT discovered that electrical QC inspectors 1. .erviewed did not even know that Inspection Reports (IR's) were supposed to include verification of witnesses to installation of ' certair. " nuclear heat shrinkable cable insulation sleeves." The solution for j this identified problem was only to improve training, daily procedures and to ! make sure that the problem does not happer in the future. 9 _-_._._.__.y _
1
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, Mr. Barrell G. Eisenhut
, September 26, 1984 Page Four The TRT failed to state that all electrical cables are installed, and all inspections are already performed. The corrective action for the future is essentially meaningless in-this case. Further, the TRT failed to discover j that this problem had been identified previously by electrical QC inspectors on MCR's, as well as in discussions with Quality Engineers (QE) who had written the inadequate procedures in the first place. Perhaps this can be explained by the fact that the TRT did not interview the electrical QC inspectors who in-
) itially identified the problem. Y-d - The September 13 letter also inappropriately places with TUEC the responsibility to give the NRC assurance that all the QC inspections which i required witnessing "for butt splices have been performed and properly docu- p ,1 mented," and verifying that the' butt splices are properly identified on the appropriate' drawings and are physically identified within the appropriate
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panels." 'Your letter did not even require that the methodology TUEC uses to reach its assurance be reviewed prior to the work being done, nor does your lette .specify tha.t the. assurances will come from a review of 100 percent of l' the affected inspections. . U
- - ~
Additionally, the TRT mentions a lack of splice qualification require-i ments. .4s you know, the lack of qualification requirements is an extremely significant deficiency. In some cases, the lack of qualification requirements for electrical cable has resulted in the requirement for cables to be replaced. ! Your suggestion to TUEC that the development of adequate installation and in-spection procedures for future wire splicing materials is grossly insufficient. The first step of any review should be to ascertain those cables which have I splices and all circuits affected by those splices. We note that without any idea of the extent of the problem, Mr.
- Youngb.csod of your office, has already judged that the use of wire splices l inside control panels is acceptable. (See September 14, 1984, letter to M. D. ,
Spence from B. J. Youagblood,
Subject:
Acceptability of Updated FSAR Commit-ment on the Use of IEEE-Std-420 for Comanche Peak Steam Electric Station (Units 1 and 2).), l I i k
" ~ '
Mr. Darrell G. Eisenhut
- September 26, 1984 Page Five We believe this microcosm of activity by your office is indicative of the approach being taken on serious technical and hardware deficiencies at the Comanche Peak plant. Apparently, the commitment to erase the problems is more important than the more prudent approach of first determining the extent of the problem, and then ascertaining whether a solution is acceptable. By separate letter today, we have requested under the Freedom of Information Act (FOIA), all material available to Mr. Youngblood which he used as a basis for the staff's evaluation report on wire splices.
We also note that no, QC inspectors who identified the splice qualifi-cation problem were contacted by the TRT. In another technical-related problem, we note that the TRT found cable terminations which did not agree with their location on the drawings.
' Your response is to require TUEC to do an "as built walkdown" of the locations ,
of all safety-related and associated terminations in the control room, panels and in the termination cabinets in the cable spreading room. This assumes that there is a' set of, drawings that is, in fact, accur' ate. We do not be'ieve that to be the case. 'We do believe that the requirement will amount to an NRC instructed update of electrical drawings instead of a documented reinspect %n. This is particularly likely given the lack of specicivity in your request r a reinspection--such as your failure to set what level of 'nonconformances u ; j an unacceptable level, and what the original scope of the inspection should be. The final electrical problem identified in the TRT's September 13 I report reveals incredible failure to expand the scope of the TRT's own review when problems are discovered. The issue, improperly closed NCR's on vendor-j installed terminal lugs, should have demonstrated again that the vendor inspection program at Comanche, Peak has significant problems. Evidence I available to us indicates that the flaw applies to NCR's on all vendor-installed components. Even more significant than the improperly disposit .ned NCR's, are site procedures and practices which improperly exempt all vendor-installed components from QC review during construction, including their exemption from the final quality document review. The TRT report drew conclusions about electrical equipment separation
, _ ___ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ . _ _. _ ._~
. . , . ~
. Mr. Darrell G. Eisenhut
< September 26, 1984 Page Six violations, but failed to explain what standard was violated--ES 100 or Nu Reg Guide 1.75. Without specifying what the minimum separation requirements are, it is impossible for either TUEC or the public to ascertain what the TRT is talking about. Without such clarification, it is a meaningless exercise to attempt to evaluate TUEC's analysis justifying their violations.
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The language of the entire section is an argument in the abstract, since the fundamental issue--the separation criteria which is approved by the
.NRC--is not identified.
It is also appropriate to note that this item was also brought to the attention of the TRT by a former employee, and that employee has never been subsequently contacted for clarification of this issue. { There are, numerou,s other [xamples of significant findings which' we believe'are not adequately addressed by scope or corrective ac, tion request. In summary, the TRT's update report provides the following insights into the million dollar effort launched by your office: .
- 1. It has continued to take a piecemeal approach to the increasing number of identified problem:
- 2. It fatis to provide any assurance that proolems other than those identified by the whistleblowers, inter-venors, or others will be independently found.
- 3. There is no attempt made to quali,fy the percentage of deficiencies discovered.
- 4. No " backward look" is being required to examine completed systems which have been inst ' led with the use of inaccurate documentation, ur ear pro-cedures, or unqualified craftsmen or inspectors.
Your review team is to be commended for their diligent pursuit of a mammoth number of deficiencies which have accu'mulated over the last seven years. Unfortunately, their ability to do a comprehensive job is being
7c_. Mr. Darrell G. Eisenhut
- September 26, 1984 Page Seven hampered by a methodology which fails to incorportte the best information available--that of those employees, quality control inspectors, and engineers who know the scope, extent, and root cause of the problems they brought to the attention of the NRC.
I,t is unfortunate that your agency cannot rise above the mistaken im-pression that only NRC investigators, inspectors and engineers are able to { identify and evaluate serious deficiencies. GAP's concern continues to be that your TRT will conclude with a compendium of over 1,000 allegations from current sources, with a matched set of 1,000 resolutions advanced by TUEC and endorsed by the NRC. That scenario does very little to answer the ultimate questions about management competence, quality assurance breakdowns, documentation chaos, and indeterminate systems, structures, and components throughout this plant. ,-
- y .. . ...
As you are aware, I was not able to attend the TRT briefing last week because of the ongoing harassment and intimidation hearings in Ft. Worth. After a review of the recent IE, O. and other NRR reports, including this TRT report I respectfully request to meet with you and Mr. Ippilito at your earliest conveience. - Sincerely, Ab i Billie Pirner Garde l { Citizens Clinic Director ) BG:me cc: Service List 4 1 1 4
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T v .... .. . . SERVICE LIST Mr. M. D. Spence Mr. James E. Cumins President Resident Inspector / Comanche Peak Texas Utilities Generating Company Nuclear Power Station 400 North Olive Street, L.B. 81 c/o U.S. Nuclear Regulatory Dallas, Texas 75201 Comission Post Office Box 38 Nicholas S. Reynolds, Esquire Glen Rose Texas 76043 Bishop, Liberman, Cook, Purcell & Reynolds Mr. John T. Collins 1200 Seventeenth Street, N.W. U.S. Nuclear Regulatory Commission Washington, D.C. 20036 F.egion IV 611 Ryan Plaza Drive Robert A. Wooldridge, Esquire Suite 1000 Worsham, Forsythe, Sampels Arlington, Texas 76011
& Wooldridge 2001 Bryan Tower, Suite 2500 Mr. Lanny Alt.n Sinkin Dallas, Texas 75201 114 West Seventh, Suite 220 I
Mr. Homer C. Schmidt l Manager - Nuclear Services Mr. B. R. Claments
] Texas Utilities Generating :ompany Vice Preside.it Nuclear Skyway Tower Texas Utilities Generating Compan 400 North Olive Street, L.B. 81 Skyway Tower l 400 North Olive. Street, L.B. 81
+ Dallas, Texas 75201
'- - Dallas, Texa: 75201
'9 Mr. H. R. Rock .
Gibbs and Hi", Inc. William A. Burchette, Esquire
~393 Seventh raenue : 1200 New Hampshire Avenue, N.W.
New York, New York 10001 Suite 420 Washington, D.C. 20036 Mr. A. T. Par r Westinghouse , actric Corporation David R. Pigott, Esquire Post Office Box 355 Orrick, Herrington & Sutcliffe Pittsburgh, Pennsylvania 1E230 600 Montgomery Street San Francisco, California 94111 Renea Hicks, Esquire Assistant Attorney General Anthony Z. Roisman, Esquire Environmental Protection Division Trial Lawyers for Public Justice Post Office Box 12548. Capitol Station
- 2000 P Street, N.W.
Austin, Texas 78711 Suite 611 Washington, D.C. 20036 President
. Citizens Ast lation for Sound Energy l 1426 South P k Dallas, Texas 75224 .W Ms. Nancy H. Williams CYGNA 101 California Street i San Francisco, California 94111
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Texas t. ;iities Gererati ; C:::any ATTN: R. J. Gary, Execu:1ve Vice Presicen-and General Manager 2001 3ryan 7:we.r _ ..
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i:c0: Gen:1e:en: This refers :: the special inspeccion conducted as a resul: cf concerr.s expressed by Me:srs. M. Walsh and J. Coyle during :na July and Se:: amber 1952 avidentiary hearing sessi0ns en Cc:anche Peak. The special ins;ection .ias
- enducted by Massrs. J. !. Tapia, R. G. Taylor, anc Cr. J. R. N ?ajan Of cur 5 aff, anc Cr. U. P. Chen of the Cepart cn . Of Energy's Energy Tech.nci:;y Er.ginaering Center (iTEC) during the ;eried Oc::'er c 13.Cece :er 2, 1952 ard
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Permi s CF?R.125 and CFFR-127 f:r :ne C:mancne Fcck S:sa E!a: ric S:2-1:n, Units 1 anc 2. . n exit in erview ..as c:nduc:ad en February 3,1333. Areas examinec durir.g the ins;ec-icn anc cur fincings are discussec in - e
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, 4 selec-ive examination of precedures anc re:resen stive rec:r:s, 'nterviews .:' "
- erscnnel, and evalua:':n of cesign :echnicues.
'a'i .nin the sc:pe of :he ins;:ec*.icn, 4 new unresolved items were fier.-if f e: ' . '
Ce: ail Sec:fcn, :aragra;hs 3.c and 3.j cf :he enclosed re::r .
- n 1 :cedance wita IC C??. 2.790(3), a c:;y of : .is ie::er and - e en:!::u e W.*il ba. .l.=.=.. '. . ...a. . . N.Rc. 3'rw i . .d '. w w- . . . .a .q . .'%. . . . . . . '. a. s a ' ,v 'J. ..-.'.'/- ...#.-. .-'...-a.,
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... . t;..---;::J ra Texas Utilities Generating 2 Co.pany Should you have any questions concerning nis inscection, we will'$e ples.2cc ::
discuss 0.9em with you. Sincerely, "CnsmM I; * *' *
- c.L. MAOSEN" G. L. Macsen, Chief Reactor Pr: Ject Branen 1 cnclosure: - .
Accendix - NRC Inspection Recort 50-445/32-25 50-446/82-14 f c: w/ enc 1: Texas Utilities Generating Company ATTN; H. C. Schmict, Project Manager ,
,, 2001 Bryan Tower Callas, Texas 75201 ' .
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APcEN0!X U.S. NUCLEAR REGULATORY CCPHISS CN REGICN IV Report: 50-a45/32-25 50-446/82-14
.; Ocekets: 50-445; 50-445 Catescry: AI i
Licensee: Texas Utilities Generating Comcany (TUGCC)
, 2001 Bryan Tower
; 0allas, Texas 75201 5
4 Facility Nems: Comanene Peak Steam Electric Station, Units 112 Inspection At: Cemanene Peak Steam Electric 5tation; Gibos & Hill in New '.orx
'l City; and Nuclear Power Services, Inc. (NPS!), in Secaucus, I.! Nw Jersey i
Insp'ection Conducted:' October 13-Oecemcer 2,1522 and January 18, 1983
. Exit Interview Conducted: February 8, 1563 a
?
- nspectors: _
,. /nN- 2ffal32, J./D. Tacta', Reacter Inspector Cate' Eng/neering Sec* / cn, Reactor Project Branch 2
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R. G. Taylor [, Resicent Reacter inspec .or - Jaty
', Construction i
J. R. N.
$ lbRajan,/ Meenanical d A r:hs 1 Engineer, 10!53 Oa e
- Mechanical Engineering 3ranen, NRR _
$ /b i ,,~ t. 1 ~c A. 2* ?.h
- W. 9. Chen, Manager, Stress Analysis unit, Energy Da e .
Technology Engineering Centar (ETEC) g>0 c S e -e. We = . 8 * . E
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Other NRC and Contractor Technical Personnel: 9 0. Rotheerg, Structural Engineer, Structursi Engineering Branen, NRR
! 0. Smith, Senior Materials Engineer, Materials Engineering Branen, NRR
- H. Fleck, Staff Member, Stress Analysis Unit, ETEC
,,' J. 8 rammer, Staff Member, Stress Analysis Unit, ETEC V J. Fair, Senior Mechanical Engineer, Inspection & Enforcement
R. Sosnak, Branch Chief, Mechanical Engineering Branch, NRR l.
Aoproved: [b T. F. Westerman, Chief Data M/ y Reactor Project Section A O
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- 0. M. Hunnicutt, Chief Y/ E.5
~0 ate '
J ' Engineering Section
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,~_. / 7 // 7 m, 5. S. Surwell, Licensing P'roject Manager Oate
] Division of Licensing, NRR insoection Summari .
Inscoction Durine Period of Octooer'13-Oecember 2.1992 and Januarv is.1983 (Recort 50-445/82-26: 50-446/82-14)
. Areas Inseectad: Special, announced inspection of the pipe succort engineering program in response to concerns exoressed at the ASLS hearing by witnesses Messrs. Walsh and Doyle. The inspection identified 19 br:ad areas of concer-expressed by Messrs. Walsh and Ooyle, determined the cesign status of the pice supports used as examples of these concerns, evaluated the validity and safety i significance of each concarn, inscected the design crocacures and Oracti:as of the pipe succort design organi:ations, and inscected a sample of 100 1:e succort casigns which had passed tnrougn tne comclete design rev'ew Orc:sss.
The inscaction involved 1,322 ins:ecter-nours by tne NRC ins:ectors an:
- nsultant personnel.
Results:
- 1. The results of the scecial inscection of the 19 broad areas of concern
, excressed Dy yessrs. Walsn and Ocyle are summari:ec in tne fo11: wing i f st.
In its insoection of all of these concerns, the Specisl Inscecti:n Tasm ! cid not finc any viciations of NEC regulations. The Special :nsce:-fon , Team did find two areas in wnica tnere are a total of four matters -nicn t l J
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1 , d n 3 3 the 5:ecial Ins: action Team c:nsfears : te "unresolvec"; that is acci-3 tional informatten is reeded in order to reacn a conclusion as to comoliance l ^with the provistens of 10 CF2 Part 50 Ap;endix 3 or Otner a;;itcac!e Commission regulations. In addition, there were four matters en wr.icn accttional ef fert is recuired to complete resolution of the cercarns. :n all of these cases the Applican:
, nas identifiac a similar preolem in :ne c:urse Of i:3 casign evicw program and is uneartaking ccerective action. However, since One worx is still in progress, :nese catte-s are ident!?ied as 0;en 1: ems wnica fil te followed 3
in the course of ne .NRC's c:nstruction inspe::ica :r: gram. The Applicant's design pr gram and design review crecedures are acecuate :: crevics reasonacle issurance that appropriata correcti"re acti:n will :e taxen. The Special Inspection Team's c:nclusions esgarcing asen of tne l? breac
. areas are given below. The only remaining catter is verification that the corrective action has teen comcletad. The order anc identification letters for each concern are the same as utiti:ed in Paragrapn 3 cf the Details Section of this re ort.
- a. The interfacino between cice sue:cri desien creues: No vic14tions of
' NRC'recuirements or acectec stancarcs were 1cen:1fied. The cencern regarding the interface between the cipe su;;crt design grou s has 1 not been substantiated.
- b. Interfacing between cice succert desien croues and cice stress analvsis orcanizations: The Applicant's itarative design review program =re-vices sucstantial assurance that pipe sucpert design defects will be identified and correctac prior to or curing the As-Buit: Ve ri fication
- Program. Mr. Doyle's concern was not substantiated.
- c. Design analvses for Richmond inserts and Hilti bolts: Mr. Ocyle's concarn acout large loacs on cencrete ancnors cue ;o LCCA induced thermal expansion of pipe support tuce steel is not substantiatec.
Mr. Doyle's concern about high bending stresses in the c:ncrete - anchcrage colts for Richmenc insarts is in par: confirmec, sicce such stresses are not calculated :y ne Acclicant anc snould te calculatec in order := assure :nat these stresses co not exceed :ne A5ME C:de allewacle stress for 201tirq. This is an unres:17ed item. However, one Special Inspection Team consicers it unlikely ena: suen stress will lead to failure of tne belt. In accition, curing the course of its review, the 5:ecial :ns: action Team identified one issue, not raised by Mr. Coyle, relating :: the sufficiency of test data usec to succor: the use of i 1/2 inch Ricnmond' inserts. The Acolicant's tas; program in res ense ::
- ne Sc,ec'ial Inscection Team's finding is an unresolvec item.
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- d. Dif ferential thermal exoansion ef fects in cice succorts: The Soectal
. Inspection Team agrees that the cifferential thermal ex6ension effec s
- resulting frem LOCA conditions coes not need to te consicerac in the
- 3 design of pipe support members. The loacs and stresses in the members ,
i will be reduced cue to the flexibility of the anchor connection. The y Special Inspection Team concluced that Messrs. Doyle's anc Walsh's ] 1 concern does not have a valid technical basis.
- e. Di ff erential thermal exoansion and otner Pfects in w.tli-to-w.all .
- ) floor-to-cei lino , ano floor-to-wal l o1ce succohs: The effecta of
'j cifferential hermal expansion due to LCCA anc concrete credo aere found to be acceptable based upon a review of dtsig,7 guiceldres, worst-case analyses, and construction practices. With respect to . spismic disclacement the Aoplicant icentified a similar problem witn ' respect to certain service water system scoports during the coursa of its design review program and is correcting this matter. The
', redesign resolved the concern. The NRC staff will verify the ecm-
. plation of the modifications in a follow-on inspection as part of its construction inspection program.
- f. Stabilitv of ofee sucoort destens: With respect to Messrs. Walsh's anc Ocyle's concerns aoout tne staoility of non-rigid box frames with
, gaps, the Applicant identified the same proolem as a result of its design review program. The Special Inscection Team concluded that
. Messrs. Walsh's and Doyle's concerns relating to instability of pipe , supports is resolved by the Applicant's stability reassessment prpgram.
The NRC staff will verify that these modifications are completed in a follcw-on inspection as a part of its construction inspection program.
. g. Use of U-bolts in cice succor: cesien: Mr. Doyle's concern acout the restraint by U-bolts of lateral movement of the pipe cue to thermal expansion at one-way restraint points, anc his concern aoout the preloading stresses have also been identified in the course of tne Applicant's normal review program and tnesa problems have been .
rectified. Mr. Doyle's other concerns about the use of U-colts have . i been found to be without a valid technical basis,
- h. Leadino cue to seismic accelerttien of tne oice succort structure:
Mr. Walsn's concern regarcing a neto to incluce seismic accelerstf ons in the pipe suoport cesign analysis and Mr. Walsn's analysis project-ing failure of the pipe succorts under seismic loads are witacut a valid technical basis. Mr. Ocyle's concern :nat tne pice stress analysis did not acec;uately consider the acced weignt of the succort was also witnout a valic technicai asis.
- 1. Mement rsstraint and local ofoe stress cue to welcec stancnions on ofees: Wita respect to Messrs. walsn's anc Ocyle's concern tnat tne effect of welcea stancnions on piping had not been inclucac in tne pipe stress analysis, the Special Inspection Team found tnat tne Aeolicant has included these ef f ects in the As-Buiit Verificatica m
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{y Program. Witn respect to moment restraints, the Applicant nac icant.1-fied a similar proclam witn respect to a unique design usec for some
$ of the main steam supports and was correcting ne proolem. The I Special Inspection Team found the method of analysis for the correc-f'. tion to be acceptable. This concern is resolved. .i j With respect to local pipe stresses, the Special Ins;:ection Team con-1 cluded that the Applicant's method of analysis is ac:ectacle. This 2
concern is resolvec, however, during tne course of its review of S this concern t?.e Specia? Inspection Team icentified a su;oort involv-
- .! ing a special condition warranting consideration of cifferential d thermal expansion. The Applicant incicated that it had catermir.ec
] that the stressas were accectacle and agreed to provice tne Scecial . % Inspection Team witn its analysis. The Special Inspection Team will ,
p, verify the ecceptaottity of this analysis. 4 ['.! j. Deflections and local stresses in cice suecort structuresi Mr. Doyle's ( concerns accut excesstve ceflections in certain suppor s nad in two i instances aiss been icer.tified by the Applicant's design review pro- }l gram. In er.e case the pacblem has alreacy been rectified and in tne
.: other the proclam is te be rectified by redesign. The corrective '!. action will be verified by the NRC staff in a follow-on ins:ection.
- i. Mr. Ocyle's concerns in two other instances nave not been substan-g tiated. Thus, the cor.cerns raised by Mr. Ocyle are resolved.
Ot. ring :he course .cf its review of these concerns, the Scecial
, Inspection Team icentified ar.cther matter, not raised by Mr. Ocyle,
'j which required accitional information relating to supoort 3 f ffness. Two studies which the Apolicant has agreed to provide remain 9 unresolved itams.
'o ; 1. Consfeerstion of friction loads: Frictional load critaria between-1 pipe and sup or me.t:ers used by the cipe sup ort design gr:uos, t a!though different, were found to be acceptable. The concern is .
d resolved. I N I 1. Corsicoattien of kick 1 cads: Mr. Ocyle's concern was found to te . incorrect.
- m. Modelinq of wide flance mem:ers as infinitelv ricid in :Psien:
Mr. Coyle's concern was found to De incorrect.
- n. Effect of cele-fermine on the ductilit'/ of tuce steel: The ASCO Grace S calc-formed tuce steel is sufficiently cuctile to :erfcrm i:3 design intent. The concern is resolved.
- o. 0:eratin'c condition loads aceear to be in error: The Scecial Ins:ec-tion Team conclucac :nat tnis corcern was witncut a valid tecnnical basis. The concern is resolvec.
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- c. p. Welded stecced connections, fillet welds and skewec weles: Mr. Ocyle's
'.' concerns about welded stepped connections in circular tucular joints, i undersized fillet welds, and skewed T-joint welds have not been sub-
. stantiated. This concern is resolved. One unresolved item previously identified by the NRC dealing with QC inspection procedures for skewed a welds is still under review by the NRC Region IV staff.
- q. Section crecerty values utilized bv pice Suecort Encineerinc: The
. Special Insoection Team concluded tnat Mr. Walsn's concern a:put different tuce steel section property values utili:ed cy One 3SE 4 pipe support design group is resolved. The Applicant is currently 5 reanalyzing all large bore and Class 1 small bore pipe su port designs using consistent mem:er property values. The differences in section property values for small bore Class 2 and 3 succorts are less taan 8 percent, and will not result in unanticipatec support behavior..
T This concern is resolved. 3 r. Succort cads welded over oice cirth welds: Mr. Ooyle's concern that pipe sucport pads on Class 2 pipe supports were welded over the pipe
', girth welds is not correct.
, s. Damace to cice succort durine hycrostatic testinc: The Scecial
. Inspection Team found tnat ne pice sucport tuce steel was camagec
' prior to hydrostatic pressure testing and the damaged tuce steel was in place during hydrostatic pressure testing. Mr. Coyle's allegation regarding the cause of the damage to this succort was incorrect. ,The replacement of the damaged tube steel was verified by tne NRC staff.
- 2. The Special Inspection Team conducted a scecial inspection of 100 pi;e sucport designs wnich had received their cesign review by ITT-Grinnell and NPSI, and were " vender-certi fied. " Eacn support design was reviewec for fifteen design attributes. The review did not disclose any discrecancies in the random sample wnich would indicata a failure of the Acplicant's design verification program to identify and correct sucports to assure c:mpliance with acclicable design criteria.
- 3. Witnin the areas inspected, a unresolved items and 4 coen i ems t were i centi fi ed.
Summarv and Conclusions Mr. Walsh and Mr. Ocyle made numerous allegations of widescread design ceficien-dies in the design of 1:e supports at the Comanene 3eak clant. They su::orted their allegations with a numcer of preliminary designs anc sketenes for various succorts. The Special Inscaction Team looked not only at One specific succorts alleged to be defective tut also into related design Oractices in s:me 19 =r:ac areas enc:meassing the Walsn/Coyle concerns. The various drawings anc s<etenes offered as examples of the design deficiencies alleged by Mr. Walsh and Mr. Coyie reflected initial succort designs wnica had not completed the Acolicant's iterative design and review procasses.
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W= . The Scacial Inspection Team found in some 12 of enese broad areas (Paragrape 3, T. 9 suosections a, b, d, h, k, 1, m, n, o, p, r, and s) that tne concerns alleged by Walsh and Doyle were not suostantiated. In 6 of these broad areas (Para-
.! graph 3, subsections e, f, g, i, j, and g) some aspects of the concerns l expressed by Walsh and Ocyle had also been identified by the Applicant during g?
the course of its design review processes and the problems have been or are being rectified; other aspects of the concern were not substantiated. In one
$ broad area (Paragraph 3.c), one. aspect to Mr. Doyle's concerns relating to the N bending stress in the bolt were in part confirmed. Other ascects of Mr. Ocyle's
( concerns in this area were not sucstantiated. None of the concerns raisec ey Walsh and Doyle were sucstantiated as demonstrating serious ceficiencies in the Y Applicant's pipe support design program. Even in the area of bending stress in 3 the bolt of Richmond inserts, the Special Inspection Team considers the stressas 8 involved are unitkely to lead to colt failure. Ouring the course of its assessment of the Walsh/Doyle concerns, the Special E Inspection Team identified two areas related to Richmond inserts and the suoport stiffness values used in the pipe stress analyses, not raised by Walsh and
, Ooyle, for which further supporting information was needed with respect to e certain of the Appitcant's design assumptions. But, even with rescact to these issues, they do not appear to involve situations in which tne plant 3_ piping systems would fail to function under any design loading condition.
3 Rather, these questions relate to the need for tne Applicant to provice y additional data to verify certain assumptions used in the design analyses in order to substantiate that ample margins are available uncer all casign
; loading conditions.
(; The examples of design preolems offered by Messrs. Walsn and Doyle were interim designs and did not represent designs which had completed the Apolicant's design d review process. For this reason, the Special Inspection Team conducted a review
". of a sample of 100 vendor certified supports for 15 design attributes wnich would be inoicative of the problems alleged by Messrs. Walsh and Ooyle. The y purpose of this review was to determine whether design deficiencies had survived ? the Appif cant's itarative design review process. The review did not disclose 4 any discrepancies which would indicate a failure of the Apolicant's design :
A verification program to identify and correct supports to assure compliance with apolicable design criteria.
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3 DETAILS
.l a -
Q 1. Persons Contacted 1
- Licensee Personnel
?!
j J. B. George, Vice President and Project General Manager 4 *H. C. Schmidt, Manager Nuclear Services . 1 "J. C. Finneran, Pice Suoport Engineer 1 "J. S. Marshall, Licensing Manager, d) S. Dacko, Senior Licensing Engineer
', O.-H. Wade, Licensing Engineer 3 "J. T. Merritt, Startuo Manager 1 R. M. Kissinger, Project Civil Engineer .
3 H. Harrison, Technical Services Supervisor q 0. Rancher, Supervisor, Technical Support Design Reviewy P.S.Y. Chang, Chief Engineer, Small Sore 1f
' *G. Krishnan, Sit'a Stress Group Supervisor-{Jpe Design Grouo j ~
- 0. Westbrook, Technical Services As-Su{} Coordinator G, Abele _ Supervisor, Site Engineering-
.i "M. McBay, Engineering Manager
.! *R. Jones, Manager Plant Operations -
. Other personnel "M. A. Vivirito, Manager, Analytical Engineering, Gibbs & Hill P. R. Rajan, Senior Project Engineer, Gibbs & Hill 1 "R. E. Ballard, Project Manager, Gibbs & Hill i F. A. Colucci, Applied Me.chanics, Gibbs & Hill C. I. Corean, Chief Engineer, Applied Meenanics, Gibos & Hill E. L. 8etkor, Supervising Engineer, Structural, Gibes & Hill
, H. W. Mental, Group Supervisor, Pipe Stress Analysis, Gibos & Hill
- 3. Bayles, Metallurgist, Gibbs & Hill E. Eramia, Engineering Manager, Site Engineering, ITT-Grinnell .
T. Smith, Manager, Apo11 cations Engineering, ITT-Grinnell P. J. Fang, Manager, Piping Structural Analysis, ITT-Grinnell O. Powers. Engineering Manager, IT'*-Grinnell J. Mangasarian, Supervisor, Applications Engineering, ITT-Grirnell G. Breidenbacn, Engineering Nanager, Nuclear Power Services, Inc. (NPSI) H. 0' Errico, Project Manager, 'NPSI F. Samaan, Structural Group Sucervisor, NPSI T. Sharati, Assistant Structural-Group Sucervisor. NPSI,
, C. Maitey, Supervisor, Apolied Meenanics, NPS!
H. t.ancelot. Director of Engineering, Richmond Screw Anenor Comcany C. W. Gay, Manager, CPSES Structural Services, Westingnouse
- R. Henrajani,' Lead Engineer in Review Certification. NPSI
, " Denotes attendance at tne Exit Interview.
1/ Contract ecoloyee, managed and sucervised by licensee.
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3 . 1 i, 9 ' II 2 2. Introcuetion g 3 Ouring the Comanche Peak evidentiary hearing sessions on July 29, and
! September 13 and 14, 1982, before the presiding Atomic Safety Licansing Soard (ASLS) regarding Contention 5 (construction QA/QC), Citi: ens Asso-S ciation for Sound Energy (CASE) witnesses M. Walsh and J. Doyle expressed 3 concerns related to the overall pipe support engineering procedures being i utilized for the Comanche Peak facility. In response to the concerns of .1 Messrs. Walsh and Ooyle, the NRC formed a Soecial Insoection Team to si address the concerns anc evaluate their significiance. Memoers of *he ej Special Inspection Team and other personnel who assisted them in their ] tasks are listed above. The inspection was conducted in several steos as follows: .. a. The Special Inspection Team reviewed the testimony and depositions q with exhibits provided by Messrs. Walsh and Doyle, the testimony and deposition with exhibits provided by the Applicant, and the,tran-M script of the proceeding. The objective of this review was to y' identify arid catalog the concerns expressed by Messrs. Walsn and Ooyle. These concerns were identified and are addressed in Para-graph 3 of this inspection report. < b. The Special Inspection Team then conducted a special insoection .at the Comanche Peak facility to determine the design status of each of ,; the pipe supports identified by Messrs. Walsh and Doyle. The design , review had been completed for only one of these succorts (Support No. CC-1-107-008-E23R; Ocyle Deposition Attachment 11TT).
- c. A series of special inspections were conducted by tne Special Inspec-tion Team at the Comanche Peak facility, Gibbs & Hill in New York City, and Nuclear Power Services, Inc. (NPSI) in Secaucus, New Jersey to determine: a) the validity and safety significance of each of :ne concerns expressed by Messrs. Walsn and Ocyle; b) the role ano responsibilities of each of the pipe support design groups, ooth on- .
site and off-site, and the piping design grouo, Gibbs & Mill; and ; c) the design procedures and practices used by each of the oice succort design groups and by Gibbs & Hill.
- d. Finally, tne Soecial Inspection Team conductec an inscection of a samole of 100 pipe suoports designed by ITT-Grinnell and NPSI wnion had been " vendor-certi fied," 1. e. they had passed througn the recuired design review precedure and nad been found acceotaole by ne responsi-ble pipe succort design group. This inspection consisted of a review of randomly selected pipe succorts for the concerns identified oy Messrs, Walsn and Doyle. The results of this insoection are cascribed in Parayrson ,
4 of this report.
- e. Af ter c'onsidering ne information received during ne aoove inscec-
- tion, the Special Inspection Team scheduled another inspection visit to the station on January 18, 1983. In these suosecuent distuss10ns certain additional details were provided or clarified.
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- f. An exit interview was conducted at the station on February 3,1933.
1 3. Concerns Related to Pioe Sucoort Desians 0 The following expressed concerns relating to the design of pipe scoports were reviewed during the inspection: j a. The Interfacinq 9etween Pioe Succort Desien Groucs 1 Mr. Ooyle expressed concern that the interface between the Aeoli-
-i cant's various design groups, primarily in the cipe suoport area but .. also including otner areas, were inadequate and were the cause of i
4 design inconsistencies (Tr. 3706, 3852, 3864, 3925 and 3973). The Special Inspection Team could not determine if the concern was i~ntended to address all of the possible interfaces involved in pi,oing and support design or only the interfaces between the specific group
.; in which Mr. Ooyle worked and the other design groups. Regarding tne latter and more narrow interpretation of the concern, the Special i Inspection ~ Team determined that Mr. Ooyle had been assigned to work ~
in the " Site Stress Analysis Group" (SSAG). .Mr. Oeyle's more specific assignment was to a subgroup witnin SSAG tnat analy:ed supoort frames utilizing the Structural Design t.anguage (STRUOL) computer progra:n
- and hence this subgrouc became known as the STRUOL group. The other
., subgroup in SSAG performs pipe stress analyser using the AOLPIDE computer program. Both programs are In a main-frame coecuter located l' -
in New York City with the site communicating with that computer by ' telephone lines. App 1tcant's engineering instruction No. CP-EI-4.5-9, " Performance Instruction for SSAG," revision 0, dated Sectember 5,1980, and the current revision 1, dated August 3, 1981 were reviewed during tne inspection. The procedure in both revisions describes. the SSAG as a group which receives requests from the various design groups for analysis of stresses in either pipe systems or in suoport frames. The stated purpose of the SSAG is to provide an intermeciata check of ; the stressas before a design is finalized or when a design change is being made to a previously finalizac design. The procedure requires that all stress analysis requests must te in writing, and accrovec cy the sucervisor of the requesting grouo. The results of tne SSAG analysis are returned to tne originating cesign group. 'The various design grous instructions indicate taa'. design groups are resconsible for the analysis entner than SSAG.
. Discussions with cogni: ant Applicant personnel and severs 1 tours of the office area housing the SSAG indicata that with one excection, the groue is a service organi:ation to the design groues and 9as no in-lineefunction in the design process. The one exception relates to the pipe analysis group wnich performs tne official pipe stress analysis of pipe runs from 2.5 intnes to 4 inches in diameter casad on soecific instructions from the resconsible engineer in the Gf bos &
Hill New York office. The STRUOL grouc is rot involved in :nis
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particular effort. The Applicant's personnel anc were involvec in ~
,i the formation of the STRUOL group have stated that the group was
{ 1argely staffed with technician lovel people when originally formed,
; but was later restaffed with engireering level personnel when-'it was 4
found that the technicians required too much help in modeling frames for input into the computer prog am.- The Special Inspection Team 3
' concluded that the Applicant has defined and documented the responsi-bilities of each engineering organization, and has also defined anc documented the communications p'ths a between the SSAG and the other
~,
groups in an effective manner, based upon review of the rt oplicant's '
.. documents, Ifsted below: .
s -
, a. CP'EI-4.6-9
" Performance Instruction For SSAG."
j: ~
- b. CP-E?-2.1 " General Program For. Pipe Support Design, Fabrication and; Installation Activities"
- c. CP-EI-4.0-4 " Field Structu p1 Engineering Grouc Design Control Instruction"
- d. CP-EI-4.0-1. " Design and Design Verification Control For
- Pipe Support Enginesring#
t
- e. CP-EI-4.0-13 " Control of Stress Analysis For Pipe Sucoort' Engineeringd - .
- f. CP-EI-4.5-4 i' Tech ical Services Engineering Instruction-for Pipe Hanger Design Review" ,
The narrow interpretation of the concern is resoived.
, Considering the concern in its $ header sense and including ne alleged inconsistent design requirements, :he Special Iqscaction Team found that Messrs. Oeyle and Walsh, as members of the STRUOL groce.
, o f the strild- ;
. tural frame engineering organizations and the differirare exposed nto the des i criteria used by these- organizations. Further, due to tneir office location, they had rwady access to the design basis documonts used oy the two offsite p'pe support design organi:ations. Messrs. Coyle and Walsh noted that the three pipe support organi:ations were eaca using different cesign anproacnes and that another approach was used by ne onsite civil / structural design grous charged with the design of caole tray and conduit suoports. In addition to differing; design acercacnes. '
they noted that each of the organi:ations appeared to be using differtnt , section prop'erty val'ues for the structural shapes involved. Mr. Ocyle in particular seemed to feel that, had the design basis inputs and interfacas been adequate, these differences ould'not have occur ed. He further st'ates that since such cifferences have occurrec, the Applicant nas violated the requirements of the NRC as exoressed in 10 CFR and other documents such as ANSI 15tandards N45.2, " Quality Assuranct Pacgram Recuirements for Nuclear Pycc Plants," and c _y_ y_ , y __ _ ." - * --
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l l N45.2.11, " Quality Assurance Requirements for Oesign of Nuclear Power Plants." The NRC has endorsed N45.2 via Regulatory Guide 1.28 and has also endorsed N45.'2.11 via Regulatory Guide 1.64 N45.2 is a general requirement document essentially equivalent to Appendix 3 of 10 CFR 50 while N45.2.11 is specific to the design controls require-ments contained in Critarion III of Appendix 8 and N45.2. N45.2 anc 45.2.11 were promulgated in their present form after the Apol,icant was granted a construction permit for Comanene Peak. The Special Inspection Team concluded that there is no evidence that the intanded objectives of N45.2.11 have not been achieved nor is there any evi-dance the Quality Assurance programmatic requirements of 10 CFR 50, Appendix S; N45.2 and/or See:1on NA of tne ASME Code have not caen satisfied. The Special Inspection Team found that the alleged inadequate inter-faces are not the cause of the differences in design approaches. The differences appear to be the natural outgrrwth of the Applicant's utilization of three separata pipe support design organizations and yet a fourth organization for the design of other structural succorts such as those for cabla trays and conduits. An early decision was made by the Applicant that the pipe sucport detail design would be contracted out to dne of several companies who are in tne business of designing and facricating pipe support components. In order to satisfy the American Society of Mechanical Engineers Soiler and Pres-surs Vessel Code (ASME Code) requirements and in order to. set a basis for competitive bidding between the companies, it was necessary to , provide them with the overall design criteria to be met. The Gibbs & Hill document to accomplish this objective was Specification MS-46A which crevices the information required by the ASME Code and also satisfies the requirements for design input information as described by N45.2.11, paragraph 3.2. A contract for the design of oipe supports was awarded to ITT-Grinnell Comoany in 1975. The means by whien ITT-Grinnell would satisfy the detailed requirements of MS-46A and the ASME Code became that company's responsibility. Later, it became separenc to the Apolicant enat ITT-Grinnell was not acle to maintain an soprocriate schedule for either cesign or f abri-cation of the supports. In 1977, the Applicant enterea into a con-tract with Nuclear Power Services, Inc. (NPSI) on essentially toe same specification basis. As with the ITT-Grinnell contract, the details of functional compliance witn soecification MS-46A became NPSI's responsibility. Still later, the Applicant reali:ec that an onsite cesign/ redesign grouc was necessary if an acoropriate schecule was to te maintairec. The Acplicant tnerefore created what cecame the projec* Pipe Succor Engineering (PSE) organi:ation, whien also utili:ed the same specifi-cation basis as the other two design groucs. j
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13 3 L l Since neither the specification nor tne ASME Code dictate in cetail l} the means by which an engineer is to satisfy the design criteria, d differences in engineering approaches occurred between the tnree q parallel pipe support groups. Again, in reference to N45.2.11 and W its requirements regarding interfaces, the overall purpose is to y assure that each design organization has a clear, documented scope of
;j responsibility and that there are documented paths for communication }- when the responsibility shifts from o.ie organization to the other or is shared by both. In the case of three pipe support design organi- $ zations, each has its own specific scope of responsibility since eacn 4
has been assigned the responsibility for a specific group of supports. 3 There is no apparent need for cross communication between tne three 5 groups since they share no common detailed responsibility. Furtner-
'6 more, the lines of communication between the Applicant, the A/E and j aach pipe support design organization are clear and documented.
4-
,j Based upon the above considerations and upon review of Specification
] .
MS-46A, par,ticularly pages 3-7 through 3-12: Applicant's letter dated December 21, 1981 which provides the correspondence matrix
, regarding orders CP-0046A (ITT-Grinnell) and CP-0046A.1 (NPSI); and
- Procedure CP-EP.1 " General Program For Pipe Design, Fabrication and
; Installation"; the Special Inspection Team concluded that the Appli-cant has adequately defined and documented the responsibilities and
$' paths of communications between the architect / engineer (Gibbs & Hill) and the pipe support design groups, including the responsibilities and communications with the SSAG. No NRC regulation has been violated, and the programmatic objectives of Section NA of tne ASME Code, N45.2.11, and N45.2 appear to have been satisfied. This concern is resolved. The concern expressed by Mr. Doyle regarcing the interface between the pipe support design groups has not been substantiated.
- b. Interfacinc Between the Pice Succort Desien Groucs and the Pice Stress Analvsis Orcani:ations .
a 4 Mr. Ocyle implied a general concern througneut his deposition tnat the three pipe suoport design organi:ations were utilizing casigns which induced stresses in the piping that are not considerec in the pipe stress analysis (Ooyle Deposition). Relative to this concern tne Scecial Inspection Team reviewed: 1) One cesign process for tne piping and the pipe supports in a series of discussions with cogni-zant memcers of the Acolicant's staff and tne tnree pipe succort design groups at the Comanene Peak facility, and 2) the procedures 1 for concucting the pipe stress analysis in discussions with cognizant members of Gibbs & Hill at their offices in New Yorx. The Scecial Inspectitn Team also reviewed the Gibbs & Hill instructions, "As Built Ve'rification Instruction," Revision 2 cated June 7, 1982, and the TUSI Engineering Instruction CP-EI-4.5-1, " General Program for As-Suilt Piping Verification," Revision 6 cated August 30, 1982 wnien provide the necessary steps and guidelines used by the Applicant to ). p , , , , , ., - ... _ -e.***us* *me * * * * * = * * * * * * =
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. implement its responsibilities in the as-built stress verification of j the designated piping.
j Gibbs & Hill, as the architect / engineer, is the designer of all ASME 4 Code Class 2 and 3 1arge bore piping, i.e. greater than 2-1/2 inches j in diameter. This is an iterative process involving numerous exchanges 3 of information with various subgroups of the Applicant's organiza-
- ] tion, including PSE, and with the two contract pipe support cesign
.i groups (ITT-Grinnell and NPSI). The Special Insoection Team ::en-j cluded that an understanding of this iterative design process was 3 necessary to establish the significance of the numerous exhibits put
] forth by Messrs. Walsh and Doyle. These exhibits, without exception 2 were found to be still in the design process. Thus, the Applicant's 1 d3 sign and review process had the potential for correcting all j alleged design and analysis deficiencies prior to the pipe support
.: becoming operational. This portion of the inspection attempted to j address the question whether or not approcriate procedures or guide-i lines were ,in place which assure that the corrective actions would be f, taken, i, A description of the App 1tcant's process of designing and analyzing a 1ength of pipe and its pipe supports, a " stress problem," is provided q to place the allegations and this portion of the inspection in-perspective. The process is described for a length of pipe rather than an individual pipe support because the design unit is the length i of pipe; that is, the pipe support is an accessory in the total ,
+
design probl'em and cannot be designed separately from the length of N pipe. The following simplified steps give the highlights in the
; design procedure.
(1) Gibbs & Hill prepares a conceptual design for the length of 4 pipe. The length of pipe is generally chosen to run between two i > anchor points. The conceptual design consists of a piping
;- layout which defines the proposed routing of the pipe including i
.: piping components between the ancnor points. The location and .
'{ 1engths of the straight and curved sections of piping are defined by pioing plan and elevation drawings and/or isometric drawings.
(2) Gibbs & Hill performs a pipe stress analysis on the conceptual j design to procuce an acceptable design which will meet tne ASME B&PVC (Code) allowable stress requirements. Comc11ance with aopropriate provisions of the ASME Code is requireo oy NRC regulations 10 CFR 50.55a(d). If the sonceptual design dcas
': not satisfy the Code criteria, design / analysis iterations are
?
perdermed to produce an acceotable design. Changes to the con-cep'tual pice routing and the location and numoer of proceseo l succorts are usually required curing this process. The pipe l stress problem calculates the forces and type of loads on the ! procosed pipe succorts. I i ___)
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(3) The description of the acceptacle piping layout, proposec
.. support locations, pipe movements at the support locations, anc , the directions of restraint and magnitudes of the forces for each support are sent to one of the three pipe support design groups (ITT-Grinnell Corporation, at Providence, RI, NPSI at Secaucus, NJ, or PSE at the Comanche Peak Offices) depending upon the scope of their contract or assignment. Using that information and other design data (e.g. , structural arrangement drawings) the pipe support casign groues prepare a design for each of the pipe supports given in the concactual casign for tne
- length of pipe.
(4) If the pipe support installation personnel (craft) determine
, that a succort cannot be installed as designed, PSE field engineers are notified and make changes as necessary to procuce a design that can be used. Based on their judgment on the ', . impact of their changes, the PSE. field engineers may~ request an analysis of the change from the site stress analysis groue (SSAG).
4 (5) When the pipe and some of its suoports have been installed, the Quality Assurance Group starts its as-built inspection cocument-ing the as-built dimensions of the pipe and installed pipe supports. The drawings for the pipe and pipe supports are
; revised to reflect the as-built configurations, and are stampec
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"as-built verified." When a significant portion of the succorts on the length of pipe have been as-built verified, a package is assembled and forwarded to Gibbs & Hill for a preliminary stress analysis.
' (6) The as-built package for the length of pipe is sent to Gibbs &
Hill where it is reviewed and adjusted for any new factors which may imcact the pipe stresses e.g. , pipe routing changes; supeort i relocations, orientation, deviations and restraint charactaris-tics; minimum wall violations; addition / deletion of valves, fittings and other appurtenances; valve weights, orientation of ; the operators and center of gravity; sleeve clearances / types of seal; ano changes in thermal modes of coeration. The stress preolem is rerun to determine new stresses in the pipe anc new loads on tne pipe supoorts. (7) The stress problem package is returned to the Comanche Peak where the responsible pipe suoport design grouc reviews the new pipe loads on the sucport and the final as-cuilt suoport
. configuration to assure that the suoport will meet the new functional requirements. If the sucport is found to satisfy tne new= recuirements it is stamped " vendor certified." If the sucport is found unsatisfactory, it is modified anc the new as-built design is sent to Gibbs & Hill to te assassed for its imcact on t.5e pipe stress problem.
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, (8) When all of the pipe supports are installed (and conceivably at
-r intermediate steps), data are added to the stress problem package i on the pipe supports installed since the stress problem was last
, run. The package is returned to Gibbs & Hill to assess whether
, the new as-built configuration impacts.the pipe stresses. If so, the pipe stress problem is rerun, or alternatively the
! supports are redesigned, until the pipe stresses are found 1 acceptable with the as-built configuration of the pipe and all pipe supports reflected in the stress problem input.,
(9) The stress problem package is returned to the Comanche Peak site where any changes to the loads on the pipe support are reviewec by the responsible pipe support design grouc and if satisfactory,
,, the remaining pipe supports are " vendor certified." If any pipe
~
; supports are found unsatisfactory for the new loads, the supcort ,
must be modified and the stress problem package is recycled j through Gibbs & Hill and the pipe support design group until all
, pipe stresses are acceptable and all pipe supports are vendor ,] certified for the loads developed in the last run of the stress probl em.
- It should be recognized that the acove description is simolified in
, that it does not include any recognition of the constraints imoosed O. by construction schedules / status. It is further simolified by restricting the description to only the principal or main stream participating organizations. The Special Inspection Team believes the design status of the pipe supports identified by Messrs. Walsh and Doyle fell into steps (4) and (5) of the above procedure. In essence, the suoports in question had not entered the as-built
_ verification program. In its investigation of the specific allegations of Messrs. Walsh and
, Ooyle discussed in Paragraph 3 of this report, the Soecial Inscection Team found most to be without a valid technical basis. For other concerns the Special Inspection Team found' that they had been j resolved by the Applicant's normal iterative design rev-f ew process.
On the basis of its review, the Special Inscection Team concluded that the Aeplicant's iterative design review program provides sucstantial assurance that pipe supoort design cefects will te identified and corrected prior to or during the Applicant's As-Suilt Verification Program.
- c. Desien Analvses of Richmond Inserts and Hitti Solts Mr. Ocyle's concerns in the area of pipe support concreta anchor
. design are twofold: (1) very large loads on concrete ancnors due to thermal expansion of the mice succort tuce steel uncer LOCA conditions have,been excluded in tne cesign of the ancnorage; anc (2) the method of shear and mcment analysis at the point of ancnorage is in error and may significantly affect the performance of the anchor.
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. ihe Special Inspection Team reviewed the following reference documents to assess the Applicant's overall compliance with Section 3.8 of the CPSES Final Safety Analysis Report (FSAR), and to evaluate the-engineering design adequacy of pipe supports utilizing Richmond inserts or Hilti-bolts to anchor the support to the structural concrete:
i
- 1. Gibbs & Hill Specification 2323-55-30, " Structural Embecments ,"
March 19, 1981. ,
- 2. Gibbs & Hill Report, " Evaluation of LOCA Temperature Effects on Pipe Supports," August 25, 1982.
3.
~
NPSI Report, " Load Transformation Study on Richmond Insart & Tube Steel Assemblies," September 1982.
- 4. PSE Guidelines, Section V, "Hilti Concrete Anener Bolts."
- 5. PSE Guidelines, Section VI, " Richmond Inserts and Anchor Bolts Stress Allowables."
- 6. TUGC0 Procedure CP-HBM-0.1, "Hilti Salt Inspection Manual,"
Revision 31.
- 7. Polytechnic Institute of Brooklyn Test Reports for Richmond Screw Anchor Company.
- 8. PSE Report, " Richmond Inserts - Prepared for 1-17-83 meeting with NRC.
Richmond Inserts The Special Inspection Team reviewed the Applicant's method of designing pipe supports utilizing Richmond inserts to anchor the pipe support tube steel to the structural concrete. The following . describes this review and the Special Inspection Team's conclusions relating to the Richmonc insert concrete ar.cher. (1) Thermal Exoansion Loads With respect to the concern acout the exclusion of One tnermal expansion load, the Special Inspection Team assessac the mag-nitude of the excluded load, the Aoplicant's cesign criteria with factors of safety, and finally the adequacy of tne available test data used to generate the design allowables. Th Soecial Inspection Team determined, from interviews with
, cognizant design engineers anc from calculation reviews, that the Aop11 cant had not consicered LCCA thermal exoansion effects on concrete' inserts and bolts in the design of individual pice supports anc associated concrete anchors. [A concrete anchor is
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$ composed on an insert in the concrete and a bolt whicn is used
$ to attach the support to the insert.] This decision was based
} primarily on the ASME Code Section III, Appendix F, " Rules for -
; Evaluation 'of Faulted Conditions," which does not require that 2
differential thermal expansion stresses resulting frem faulted
.i conditions be included in the design procedure. This exclusion I is based on the ASME Code rationale that these stresses occur once in the lifetime of the plant, are self-limiting in nature ., and are relieved by small deformations and displacements.
[; Although the ASME Code is not directly applicable to the design of the concrete anchorages, the Applicant adcoted the ASME Code
! philosophy in the design of the concrete inserts. This design >
approach is documentad in Sections 3.S.3.3.3 anc 3.3.4.3.3 of _ the FSAR, where it states, ". . . thermal loads are neglected when they are secondary and self-limiting in nature and when.the
} material is ductile."
h
. With r,espect to the design of inserts such as Richmond inserts, the Special Inspection Team found that these components are not governed by the ASME Code nor by any other standard which the NRC has adopted as a regulatory requirement. Thus, the nly applicable regulatory standards are tne requirements of 10 CFR 3 Part 50, Appendix A - General Oesign Criteria For Nuclear Power Plants, Criteria 1 and 2, which require that such comoonents be ~
j capable of performing their intended design function wnica is to carry the imposed loads without failure. . The Special Inspection Team has evaluated the amount of thermal expansion that would result under worst-case LOCA concitions and the available load-displacement data. For the worst-case ' 1 analysis of an eleven foot long memoer, unrestrained thermal
> ; growth resulting from LOCA conditions was computad to be 0.086 inches. The worst case condition wa's established by identifying
{ the longest tube steel member attached to the concrete. This ,
, member was a part of the feedwater system gang hanger located ,
inside containment with an overall span of approximately 30 feet. This gang hanger is anchored to the concrete by the use of 1 1/2-inch diameter Richmond inserts. From the load-disclacement curve of the 1 1/4-inch diameter Ricnmond insert in 3,000 osi concrete, the calculated growth or strain recuired to relieve the applied thermal load reoresents 22 percent of the approximate failure strain of 0.4 inches. This simplified calculation does not consider the bending of the bolt due to the 1-inch washer offset. Sending in the bol woulc nave the effect of lessening the shear force resulting from thermal exoansion due'to LOCA on tne insert. Thus, even for tnis worse case,
, the'LOCA induced thermal expansion strain contribution in tne
. insert would be recucec. The 11/2-inen Richmonc inserts usec in this design would act in a similar fashion. However, there are no deflection test data for 11/2-inch Richmond inserts in shear loading. For the reasons discussed celow tne Special i
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". Inspection Team cencludes that acditional tes: cata is required 'e for 1 1/2-inch Richecnd inserts.
t
] (2) Allowable Loads and Factors of Safety
~
The allowable Richmond anchor tension loads were established by the Applicant based on a factor of safety of two of the ultimate load as determined frem tests (Reference 7) and/or a shear cone analysis made by the Apolicant. The Applicant's analysis consisted of comparing the test ultimate tensien (pullout) Icads with calculated ultimate shear cone loads determined in accordance with Appendix B of the American Concrete Institute's (ACI) "Cece Requirements for Nuclear Safety Relatec Concrete Structures," _ ACI 349-76. The ultimate tension lead was then cefined as the lesser of the two values and the factor of safety of two was j then applied on the lesser value. For the 1-inch insert, the 4 factor of safety of two was based on the shear cene arialysis
- icad. The resultant allowable load when compared to the tes load re'sults in a factor of safety of 2.17. For the 1 1/2-inch
, insert, the factor of safety of two was based en the actual
. tension test results. Allowable shear loads were set ecual to
, the allowable tension leads and for the 1 1/2-inch inser , reducec 4 by a factor equal to the ratio of the manufacturer's allcwable
;; load values (about 0.83). Shear lead allewables for the
.' 11/2-inch insert would have a factor of safety of about 2.4 4 ~
based on the assumption that the shear test ultimate is equal to the tension test ultimate. Although this assumation is
- basically true for the 1-inch and i 1/4-inch inserts, no shear tests have been conducted on the 1 1/2-inch si:e. Published allcwable leads in the Richmond Screw Anchor Ccepany Bulletin A No. 6 are based on a factor of safety of three. As a result of the Applicant's assumptions as to shear Icac capability, the i specified shear load allowables are 50 percent higher for the 2 1 1/2-inch insert than the value reccmmended by the manufacturer.
2 :
. Richmond inserts have been used at scme other nuclear pcwer plants. The Special Inspection Team was able to icentify tha:
one of these plants usec a factor of safety of three, bu: cic not learn the factor of safety usec for Richmenc inser s at :ne other nuclear power plants. The Applican; stated that the manufacturer indicated that a factor of safety of less tnan three has on occasien been recomencec in tne concrete precas: tilt-up industry. Frem a review of the manufacturer's data pubitshec in reference 7, the Special Inspection Team determined that the manufacturer's all.cbable shear values for the 11/2-inch diameter Richmond inser: were extrapolatec frca snear tests en 1 1/4-inch ciameter insert. Althcugh the . published allowable values are theore-ically valid, standard industry practice requires that testing be performed to confirm the values. In addition, even for the shear :es:s
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a il conducted (on 3/4, 1, and 1 1/4-inch) the test data cces not (, fully model the configuration of the ancnor assembly used with a n 1 inch thick washer between the wall anc the support frame. This 1 washer introduces a bending moment in the bolt which is not 1 reflected in the shear test results. i" No combined shear / tensions tests have been performed on Richmonc 9 inserts by the manufacturer or the Applicant. For calculating i the effects of combinec shear and tension, the Appli. cant bas
.i utilized a curve based on an interaction formula given in the Prestressed Concrete Institute handbook. However, the applica-i tion of this formula for the 11/2-insert is based on the use of shear values extrapolatec from the 11/2-inch insert.
The Applicant has stated that ACI 349-80, " Code Requirements for
- Nuclear Safety Related Concrete Structures," an industry standard f not adopted by the NRC as a regulatory recuirement, allows a s.
i factor of safety of two for concrete inserts. The Special
, Inspection Team found that the ACI standard specifies load
., factors and capacity reduction factors and requires considera-tion of the' forces causec by thermal effects under accider.t conditions. In addition, the ACI standard requires a testing
] program far broader than that which has been carried out for the G Richmond inserts. The Special Inspection Team cannot concur
!! that the_ ACI standard allcws a factor of safet) of two to be j used in the manner in which it has been used by the Applicants
; The Applicant's factor of safety of two for the anchorage would
'; be sufficient if based on test data for the size used insice
% containment (11/2-inch) and if it was based on a test in a load-
.. ing mechanism that modeled the actual configuration. The actual configuration, which utilizes a 1-inch thick washer, introduces a bending mcment in the bolt which may influence the load y? displacement characteristics. In addition, the inser s should have been. tested in combined shear and tension if a factor of jl
^'
safety of two is too be considered sufficient. Conversely, the uncertainties introduced by the use of shear values for 1 1/2-incn inserts extracolated from tests on 1 1/2-inch inserts, ne use cf data frem a test that cid not mocal the configuraticn using -he 1-inch tick washer, and the use of generic shear /tensi.on corela-tions in the absence of any shear / tension test for Ricnmond inserts would not be significant if the design loading for the insert, were based on a higher factor of safety,for the anchorage. The encertainties introduced by the test moceling, consicerec together with the limited test data availacle, result in insuf-ficient evidence to accept that the f actor of safety of two for the considered. loads (which disregards loads on the inserts anc bolts resulting from thermal expansion of the at ached su; cort 1
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I anc cending mcments introcucec by the 1-inch thick washer) is j adequate to assure that the Richmond insert assemblies are
; designed with a ample margin for the intended icad carrying 1
i functions. Accordingly, the NRC staff will require that acditicnal testing be conducted to verify the acequacy of the design criteria { utilized. An applicable standard test method is the American j Society for Testing and Materials (ASTM) " Standard Test Metheds j for Strength of Anchors in Concrete and Masonry Elements,"
; ASTM E488-76. This standard delineates an acceptable testing j and reperting procedure which can be applied to the Richmond inser s.
$ (3) Shear and Mcment Desien
~
. . The Special Inspection Team investigated the concerns related to the adequacy of shear and moment design in the actual Richmond
, insert configuration by evaluating Reference 3. This.stucy
's defines the load transformation behavior of a typical inser anc tube steel configuration based on a finite element medel analysis j using the STARDYNE computer code. Tension, shear and moment were calculated for five load cases which represented midspan f axial unit loading along the three principai axes and uni:
q torsional loading transverse to and alcng the axis of the tube
- steel. This study indicates that in all cases the transfer of
; shear frem the tube steel to the Richmond inser: belt occurs
; primarily in the flange of the tube steel nearest the concrete wall. ~ The analysis shewed that the highest value of shear in the flange away from the concrete wall represented 18 percen: of the shear force in the flange near the wall. This behavior, which was verified by the computer analysis, indicates that bolt bending leads to a distributien of shear forces primarily 3 to the tube steel flange near the. concrete surface.
.1 Mr. Doyle's concern that high bending mcments in the bolt result
~
from the shear force being offset from the concrete surface was t evaluated by the Special Inspection Team by calculating the P stresses in the bolt due to the offset. The Special Inspec fen Team found that the Applicant does not calculate the stresses in the bolt in the design of the concrete anchcrage. A1:ncugh :he released mcment and resulting stresses cue to the tuce steel being offset from the concrete by a 1-inch thick washer is neglected during the normal course of design, it was quantified by the Applicant in the STARDYNE analysis. Calculation by the Special Inspection Team of the stresses resulting frc= the snear, tension, and bending moments for the five leading cas,es analy:Ed, indicates that bending stresses in the bolt for the worst-case condition are 15 times larger than the stresses resulting frem'snear. Althcugn cending in the bcit may result in reducing shear on the insert, it imparts an acci f onal cending stress in the bolt which has not been calculated. The Applicant has 4 g9 m,,,e., , e - em = * * = * ' = +- -~ P
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offered some prelicinary calculations indicating that bencing j moments are ind snificant in all but one of 60 cases reviewed.
; It may be that t e effect of such mcments are small in the large majority of cast- While there have been questiens about j whether the bolting is governed by the ASME Code, the NRC staff ; believes that the total stress (including the bending stress) ? in the bolts shculd be evaluated to assure that the value for ? allcwable stress has not been exceeded. The NRC staff requires that this value shall not exceed the ASME Ccde allowable stress -f fcr bolting. +;
Curing the inspection, the Special Inspection Team evaluatec the ability of the Richmond insert / tube steel assembly to resist r , axial torsion. The Special Inspection Team fcunc no concern
. with the Applicant's design guidelines being utili:ed to design E .for this form of l'eading since they are based on valid engineer-
) ing principles for the design of baseplates. Mr. Ocyle's concern 1 . about the eccentricity between the tube steel and the Richmond insert sas also evaluated by the Special Inspection Team. The
.) Applicant's design criteria limits the eccentricity to two times the thickness of the tube steel wall. This criteria is a result 1 of the necessity te establish a maximum alicwaole erection
'5 tolerance which can be accccm.cdated in the factor of safety
. . withcut significantly affecting the design calculation. The philosophy behind specifying minimum factors cf safety for
-. any design results from the need to establish a reserve capa ,
~
bility which will acccunt for the possibilities of overicad anc
, understrength. Such pcssibilities may be due to variations in material dimensions, variatiens in construction procedure implementation, simplifications in calculaticn procedures, i effects of erecticn tolerances, and disregard of secondary
( stresses. , i
! Hil ti-Bolts 4 '
A The Special Inspection Team reviewed the Acolicant's methcd of designing pipe supports utilizing Hilti-bolts to anchor the pipe supccrt to the structural concrete. Due to the high safety margins used for the cesign of anchors using Hiiti-boits, the resulti .g smail load frem LCCA-incuced tnermal expansion woulc be unimpcrtant. The situation for Hilti-bolts is different than for Richmend inserts. Hilti-bolts are ecmonly usec throughcut the nuclear incustry. As a result of NRC Sulletin 79-02, a great deal cf test data has been generated about the perfor:.ance characteristics of Hilti-bolts in the si:es and-configurations used at Comanche peak. The design of the Hilti-bolts utili:es a factor of safety of five. With respect tc d'.illec r Hilti-bolt anchors, the Special Inspection l Team found that the Acplicant's design criteria and installa icn procedures are in accordance with NRC requirements anc will provide
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S . e 23 k i e l acceptable conservatism in tne cesign of pipe supports utilizing , j Hilti-bolts. 1 f This finding is based on the results of the Applicant's testing program conducted on site with the assistance of Hilti Fastening a Systems, Inc. This testing program was conducted to establish the j necessary torque requirements and to provide response to NRC Sulletin 79-02. i, Summary Mr. Doyle's concern that there are large loads on c:ncrete anchors due to LOCA-induced thermal expansion of pipe support tuce steel
,- which are excluded in the design of the anchors is not substantiated.
, Such loads, although not included in the design process, are not
? large enough to result in failure of the anchorage as alleged by 1 Mr. Doyle. Mr. Doy1'e's concern about high bending stresses in the concrete anchorage bolts for Richmond inserts is in part confirmed.
j Such stress'es are not calculated by the Applicant. These stresses
, , should be calculated to assure that they do not exceed the ASME Code allowable s, tress for bolting. On the other hand, the Special Inspection Team considers it unlikely that such stresses will lead 5
to' failure of the bolt. This is an unresolved item (Unresolved Item Nos. 50-445/8226-1 and 50-446/8214-1). In addition, as' discussed i . above, the Special Inspection Team is not satisfied with the sufficiency of the test data supporting the use of the 11/2-inch Richmond inserts. The Applicant's test program in response to the Special Inspection
,' Team findings is an unresolved item (Unresolved Item Nos. 50-445/8225-2 and 50-446/8214-2).
s '
- d. Differential Thermal Exoansion Effects in Pioe Succorts l y Mr. Ooyle expressed a concern that stresses due to a Loss-of-Coolant
,: Accident (LOCA) were not included in the stress analysis for pipe A supports inside containment (Ocyle Deposition pp. 14-21 and 36-63, . '[ " and Attachment E). A similar concern was expressed by Mr. Walsn in Tr. 3109-3145. The concern relates to constraint of cifferential thermal exoansion between the supcort steel and the concrete to which the succort is attached due to temceratures of acoreximately 230*F in structures inside containment during a LOCA (CASE Exhibit 655C).
Both Mr. Ooyle and Mr. Walsh alleged that stresses in the su port steel and loacs on su port anchorage resulting from this constraint were not included in the design and analysis of pipe sucpor*s at 4 CPSES. The stresses and loads referred to in Messrs. Ooyle's and Walsh's testimony were obtained by conservative analyses tnat assumed rigid connections at the pipe support to concrete structure interface. The assumotion of rigid connections is unconservative because it d es not consider tne ability of the support anchor :: deflect when loadec. e --r s--- # - - - -+ e--
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24 f j The decrease in stresses and loads resulting from the inclusion of (. the flexibility characteristics of the connections has been demon-
'j strated by the Applicant (Applicant Exhibit 1420). Factors of safety j between 3 and 71 of the ultimata deflections are reported by the
- Applicant. Deflections rather than stresses and loads were con-j sidered by the Applicant since they are more appropriate for thermally 5 induced, self relieving secondary stresses. Moreover, the Applicant j has stated that the ASME Coce does not require that stresses due to 3 constraint of thermal expansion of supports be consicered in .the C
design of linear type pipe supports. l 4 The Special Inspection Team agrees with the Acolicant that the ASME 1 Code does not require that the differential thermal expansion effects
] r.esulting from LOCA conditions in pipe suoport memcers be included q in the design of linear type pipe supports which are covered by the w ASME Code. Further, the Special Inspection Team concluded that the 3 differential thermal expansion effects resulting from LOCA conditions
' d within pipe support members which are bolted to concrete structures
, will be red'uced due to the flexibility of the anchor connection. The
< Special Inspection Team also concluded that the differential expansion effects in pipe' suoports resulting fecm LOCA conditions does not recre .
, ; sent a safety concern based *primarily upon its analysis of the flexi-
! bility characteristi.cs of tne worst-case support-to-wall connectors 4
as described in Paragraph 3.c.(1). This conciusion is sucject to con-
, firmation of expected deflection / lead characteristics in a shear test
; of the 1 1/2-inch Richmond insert.
With-respect to Messrs. Ocyle's and Walsh's concerns regarding failure j to consider loads and stresses due to differential thermal expansion 4 in pipe support under LOCA conditions, the Special Inspection Tasm
; found that the Applicant does not consider these loads anc stresses.
j The Applicant argues that such loads and stresses need not be consicerec.
! For the reason discussed above, the Specfal Inspection Team agrees with j the Applicant that such loads and stresses need not be considered in y the design of pipe supports. The Special Inspection Team concluces .
ij that this concern does not have a valid technical basis ano considers
- it resolved.
- e. Differential Thermal Excansion Effects in Wall-to-Wall, Ficor-to-
; Ceiline, anc Floor-to-Wall Pice succorts Mr. Walsh expressed a concern regarding LOCA d'ifferential thermal expansion effects in wall-to-wall, floor-to-ceiling and floor-to-wall
'e pipe supports (Tr. 3120-3122, and 3141-3143; Walsh Testimony, p 3, CASE Exhibit 659; Walsh Supplemental Testimony, CASE Exnibit 668).
In particular, concerns about the effects of a 50*F LCCA tempera ure differential on a group of service water floor-to-ceiling pipe suoports were icentified (Tr. 3141-3143). I
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II I Mr. Doyle expressed concerns regarding (1) LOCA differential thermal i expansion, (2) differential seismic displacement, and (3) concrete 3 creep displacement. effects in ceiling-to wall pipe supports and/or 1 anchors and other supports with configurations similar to those y mentioned by Mr. Walsh (Ocyle Decosition, Volume 1, pp. 62-63, y 118-121, 145-151, 214-215, and 307-309; Ocyle Deposition, Volume 2, o pp. 4-7; Doyle Deposition, Attachments 7C-70, 140-14E, 14I-14K, and j" 18). Specifically, the following supports and/or anchors were identified relative to these concerns: (1) floor-to wall service
- water support Nos. SW-1-132-701-Y33R and SW-1-132-703-Y33R (Ooyle i Deposition, Attachment 7C-70); (2) floor-to wall moment restraint '
3 No. MR CPI-CSSSMR-02 shown on Drawing No. 2323-SI-0538-07 (Attach-
'I ment 9Q-95); (3) wall-to-ceiling anchors No. CC-1-057-021-A33A anc CC-1-008-029-533A (Attachment 140-14E and 14I-14K); (4) wall-to-ceiling frame No. RM-1-005-016-C42R (Attachment 13); and (5) wall-to-wall steam generator upper and lower lateral supports.
.s Regarding differential thermal expansion effects, the Special Inspec-tion Team v'erified that the PSE guidelines require that differential thermal expansion be considered when pipe supports span between walls
; or between the floor and ceiling. Based on: 1) the recuirements of t
Items 2 and 8 of Texas Utilities Services, Inc. (TUSI) office memor-andum of March 8,1982, regarding LOCA temperature considerations in pipe support design (CASE Exhibit 659E); and 2) Paragraph 18.0 of ITT-Grinnell Design Guidelines Section IV; the NRC found that the a design procedures are sufficient for the consideration of significant differential thermal expansion effects in wall-to-wall, floor-to-
?
ceiling and other mentioned types of pipe support configurations. In order to verify the adequacy of these design procedures, the Special i Inspection Team reviewed the Applicant's analyses of LOCA thermal
. expansion effects in: (1) the ficor-to-ceiling support No. SW-1-132-701-Y33R, (2) the floor-to wall moment restraint shown on Drawing No.
2323-SI-0538-07, and (3) the wall-to wall steam generator upoer
. lateral restraint. The results of these analyses indicata that LOCA m
thermal expansion effects satisfy Final Safety Analysis Report (FSAR) - commitments. The Special Inspection Team concluded that this concern is resolved. Regarding the effects of differential seismic disclacements, tne Special Inspection Team verified that the PSE guicelines require that
-- - ~ ~
when large frames are necessary to span across a corricor or from floor-to-ceiling, one end connection must be designed as a slip joint. (Paragraphs 2 and 13, TUSI Engineering Guidelines, Sec-tion II). ITT-Grinnell and NPSI guidelines do not have a similar . requirement. However, the Soecial Inscection Team was informed that neither pf these oice support design groups have designed wall-to-wall ot *ficor-to-cefling support frames. In subsequent discussions tne Applicant provided tne Special Inspection Team a cocy of a memo-randum dated January 19, 1983 directing the recipients, specifically ITT-Grinnell and NPSI personnel, to use the same seismic guidelines 4 as those contained in the TUSI Engineering Guidelines in the event se g we .e *
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i l they design these types of support frames. The Special Inspection Team concludes that this matter is resolved. $ The Applicant stated (Tr. 3142) that the designs of the floor-to-
! ceiling service water supports identified by Mr. Walsh had been found j to be inconsistent with the above mentioned PSE engineering guide-i line, and the supports were being evaluated by PSE at that time 1 (Applicant Exhibit 142, p. 25). The inconsistency was identified in late 1981 in the normal process of design review. During the, course j of the inspection, the Applicant informed the Special Irispection Team y that these supports would be unable to withstanc differential seismic displacements and were being redesigned. In sucsequent discussion, L' the Applicant showed the Special Inspection Team component modifica-tion cards (CMC) 46174, Revision 8, and 46730, Revision 4 showing t5at the bottom portions of Item 25 on support SW-1-132-701-Y33R i (Doyle Deposition Attachment 7C) and Item 22 on support SW-1-132-IO3-
- Y33R (Ooyle Deposition Attachment 70) respectively, are to be cut off to eliminate the ficor-to-ceiling columns on the east end of each j
support. The Special Inspection Team concluded that the redesign resolves the concern. The NRC staff will verify that these modi'ff-cations are completed in a follow-on inspection as part of its con-struction inspection pregiam (0 pen Item No. 50-445/3225-3). y In addition, the Special Inspection Team reviewed analyses by the Applicant confirming the adequacy of the floor-to wall moment restraint shown on Drawing No. 2323-SI-0538-07 and the wall-to wall steam generator upper lateral restraint to withstand differentia 1' ~ seismic displacements. Both analyses were'found to be accectable. The Special Inspection Team concluded that PSE guidelines for con-g sidering differential seismic displacements are satisfactory.
. With regard to the effects of concrete creep displacements expressed in Mr. Doyle's concern, the Special Inspection Team determined that these effects would be most severe in wall-to-wall and floor-to-ceiling supports. Accordingly, the Special Inspection Team performed ;
a review of these effects in the floor-to-ceiling service water supports identified by Messrs. Walsh and Ocyle (Ooyle Decosition, Attachment 78). Figure 9.2 of Attachment 73 snows that creec effects are insignificant for sustained loads with durations greater tnan 12 months. Since the length of time from placement of slac to installation of supports is typically a miminum of 12 months, creeo effects are expected to be negligible. (The actual time in this case was 32 months. The concrete placement Number 111-8809-003 for the top slab in the fuel building tunnel was dated Acril 11, 1978, anc the inspection report IRMH 8853 for the Hilti bolts for the support
. was dated January 7, 1981.] The Special Inspection Team concludec on the basis of the acove findings that Mr. Ocyle's concerns regarcing concret'e creep displacements are without merit anc consicers :nis matter resolved.
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I Mr. Walsh's and Mr. Doyle's concerns aceut LOCA thermal expansion loads and about concrete creep displacement effects on wall-to-wall, j floor-to-ceiling and floor-to-wall pipe supports are without tachni-cal merit. Mr. Doyle's concern about seismic displacement effects
't has also been identified in the courJe'of the Applicant's design .: review program. The identified problem has been rectified by a design modificatica. And, procedures are in place to assure that seismic displacement effects will be considered in the casign of . other pipe supports which may be affected. This matter is resolved.
- f. Stability of Pice Succorts Desicned for CPSES Mr. Doyle expressed a concern pertaining to stability of pipe supports (Doyle Deposition, pp. 95-104 and Doyle Deposition, Attachments 4 anc 13). Mr. Doyle alleged that:
- 1. Non-rigid supports, supports which could be characterized as three ,bar linkages, were unstable if gaps between box frames and
. U-bolts and the supported piping will permit rotation of the box frames or U-bolts around the supported piping.
- 2. Supports similar to those described in 1 aoove but with :ero clearance between U-bolts and box frames and the sucported
. piping are potentially unstable because:
(a) Gaps could be created between the U-bolts and supoorted piping due to yielding and permanent deformation in the
~
U-bolts. (b) Friction between the box frame and the supported piping will not be sufficient to prevent rotation of the box frame around the supported piping. Mr. Walsh also expressed a concern relating to unstable supports (Tr. . 3103-3105, and Walsh Supplemental Testimony, dated July 28, 1982, : p.1, CASE Exhibit 649H). The question of whether a particular succort is stable or unstable when standing alone coes not have an imoortant otaring on the func-tional capability of the piping system. Althougn indivioual succorts, when considered by themselves may apoear to be unstable, it is necessary only that the entire piping system and associated succorts be stable when considered as a single mechanical system. Mr. Doyle appears to agree with this concept in his discussion of supcort No. CC-1-043-026-A33R (Ocyle Decosition, Attacnment 13X). This drawing shows a vertical succort utilizing a U-bolt with :ero clear-ance w9f ch is the basis for concern numoer 2 acove. This sucport was judged to be stacle by Mr. Doyle when he stated that: p - even gge euw. e 6Neep6
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28
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"even though the structure below is apparently unstable it ,
takes so little to make it stable that a support horizontally up
' and downstream is sufficient to keep it stable" (Doyle Deposi-tion, p. 210).
The Applicant also appears to be in agreement with the above concept. The Applicant stated that it is not necessary for each pipe support to be stable by itself but that the piping and supports as a system should be stable (App 1tcant Exhibit 142, p. 23). . It is not general industry practice to explicitly address the overall stability of piping systems together with their supports in design guidelines. Rather, it is standard industry design practice to address only the structural integrity of supports in design guide-lines. The Applicant's practica corresponds to this industry prac-tice. Thus, no explicit design guidelines address overall stability. Functional adequacy, including stability, of the overall piping
, system is typically a result of the normal iterative design and review process. Furthermore, industrial experience has shown in the case of non-rigid pipe supports that if the suoport element which attaches to the pipe is prevented from rotating about the axis of the supported pipe at all times, the piping system and its succorts will be a stable mechanical system. Frictional forces are sometimes relied upon to prevent rotation of the suoport element about the axis of the supported pipe; for example, in the case of pipe clamps or -
U-bolts. The use of U-bolts is discussed further in Paragraph 3.g. The Applicant has stated that unstable non-rigid supports have been identified in their review process and corrective actions have been or will be taken where necessary before completion of the design process (Applicant Exhibit 142, p. 27). During the course of this inspection, the Special Inspection Team confirmed. that the Apolicant has begun to assess the stability of non-rigid box frame supports. The Applicant has indicated that all such supoorts will be reassessed for stability. Design modifications under consideration by the ; Applicant are intended to prevent rotation of the box frame around the axis of the supported piping. The'se proposed modifications include: 1) tne use of a U-bolt that is fixed to the box frame and cinched down on the pipe, 2) lugs welded to the pice that will be indexed to the box frame and 3) the addition of stabili:ing struts to the box frame. Since it is the Applicant's practice to cinch down U-bolts on non-rigid supports to prevent rotation, and the second and third procosed modifications provide positive means of preventing rotation of the box frames about the axis of the supcorted pice at all times, stability and hence the functional adequacy of the ciping
. system p4us the supports will be assured. The Special Inscection Team con' c luded these modifications are acceotable. The NRC Staff will,ve'rify that. these modifications are comolated in a follow-on
, inspection as part of its construction inspection program (Coen Item Nos. 50-445/8226-a and 50-446/8214-3).
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f 29 1 i 2 $ Initially, it was not clear that the Applicant had a similar it reassessment program to assure the stability cf non-rigid U-bolt j supports. In subsequent discussions, the Applicant stated that
? U-bolts are cinched down to grip the pipe on this type of suoport.
II Since the U-bolt will not become loose during service life the fj concern about the instability of non-rigid U-bolt supports is
- resolved.
2 The Special Inspection Team concludes that Messrs. Walsh's and 4 Doyle's concern eslating to instability of the pice supports is resolved by the Applicant's stability reassessment program. >i
- g. Use of U-Bolts in Pioe Succort Desicn
~ . Mr. Doyle expressed the following concerns regarding the use of f U-bolts in pipe support designs:
i (1) For rigid supports in which the U-bolts are oriented such that ( their ' principal or ' strong axis is in the direction of the design load, f.e. a one way support, the use of U oolts intecduces: s (a) Constraints on the piping system which are net included in
. the piping stress analysis (Ooyle Deposition, pp. 87-88).
(b) Lateral loads on the U-bolts which are not considered in their design (Doyle Deposition, p. 88). 3 (2) U-bolt deformations are not included in the calculations for support deflections (Doyle Deposition, pp. 195-197). (3) Whert U-bolts are cinched down onto the supported piping: ) (a) Stresses due "to preloading and constraint of differential
. thermal expansion are not considerec in the U-bolt analysis , ,' (Ooyle Deposition, p. 318). ;
3 (b) Local stresses in the supported piping due to constrain of differential thermal expansion are not considered in the piping stress analysis (Ocyle Deposition, p. 313). (c) Pipe supports may become unstable after yielding and
- permanent deformation of the U-bolts nave occurred.
Approximately 30 supports cited in Doyle Decosition Attachment 13 have U-bolts incorporated in their design and are discussed on pages 195-213yfMr.Doyle'sDeposition. Relative to the first of Mr. Ocyle's concerns, tne Scecial Inspection Team determined that Gibbs & Hill identified the same concern during the Applicant's As-Suilt Verification Program. This concerr was
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! addressed by review procedures established in a Gibbs & Hill intar-
] office memorandus dated July 16, 1982.
j The Gibbs & Hill memorandum requires that if the original thermal i expansion pipe stress analysis (which assumes that the piping is
,. unrestrained in the lateral direction) indicated that the piping c thermal movement in the unrestrained direction is greater than 3 1/16 inch, the piping stress analyses be reevaluated as follows:
}
j ' (1) The thermal expansion Code stress evaluations be based o'n the
;; results of a supolemental thermal expansion analysis rather than f the results of the original thermal expansion analysis. The
; piping was assumed to be restrained in the U-bolt lateral 3
~
direction in the supplemental analysis. The support stiffness f in this direction was assumed to be equal to the calculated , g U-bolt lateral stiffness.
'4 (2) The seismic Code stress evaluation be based on the result of the original seismic stress analysis in which the support was 4
assumed to be effective only in the direction of the principal
. axis of the U-bolt and the support stiffness value in this
'. direction was equal to the generic support stiffness value j specified in the Gibbs & Hill Specification MS-200. The
.j restraint offered by the support in the direction of the lateral ij axis of the U-bolt is ignored for the seismic analysis.
Although the Special Inspection Team initially had some concerns ' about the adequacy of this procedure, these were negated by'other y procedures instituted by the Applicant. In particular the Special s Inspection Team was informed in subsequent discussions with the f'; Applicant that all one-way U-bolt supports in which the initial thermal expansion analysis indicated a movement in the U-bolt lataral 3 direction greater than 1/16 inch were modified to accommodate the .
/, calculated lateral movement or were replaced. Example of these 4
9 modifications were reviewed by the Special' Inspection Team. ; 3 Furthermore, in the subsequent discussions with the Apolicant, seismic displacement data at selected one-way U-bolt restraints were presented to the Special Inspection Team. These data indicated that these displacements were less than about 1/32 inch. Loads associatec with these displacements are also negligible.
~
Relative to the neglect of constraint effects when the original thermal expansion analyses indicated piping movements less than i q 1/15 inch in the U-bolt lateral direction, analyses performed by the
. Soecial . Inspection Team indicate that:
0 A (1) Piping stresses due to restraint of up to 1/15 inch of this type of tnermal expansion movement are negligible for all pipe sites. l
- - - + - ,- ,e -
., ,. . .* . . .; a. " a;u m ._.:w .... ... .- . '
..,.: :-. ...s.s u j . .
( ) , , 4j, 31 A (2) Lateral loads on the U-bolts due to thermal expansion movements-
- i. ' of this type of up to 1/16 inch are negligible for all pipe 7, sizes when the relative flexibilities of the pipe and U-bolts '
are considered. f~ Based on the above, the Special Inspection Team concluded that the 2 Applicant's practice of restricting the use of U-bolts on rigid one-way supports to applications where the lateral movement of the pipe determined by the thermal expansion pipe stress analysis is limited to 1/15 inch is acceptable because the resultant forces from ' the thermal expansion and seismic loads (and consequent pipe stresses) are negligible. The concern about the use of U-bolts on one-way supports is resolved.
, 0"uring the course of its revfew of the use of U-bolts as one way restraints, the Special Inspection Team reviewed a related Gibbs &
J Hill concern regarding the use of U-bolts in rigid supports as 9 "two-way" restraints, i.e. , restraints where design loads are speci-3 fied in two" directions simultaneously. In these cases: (1) the support designs show that the U-colts are oriented such that their principal and lateral axis are in the directions of the soecified design loads and (2) the pipe stress analysis assumes that the support stiffness values in the direction of the principal and lateral U-bolt axis are both equal to the generic support stiffness values .in Specification MS-200. In subsequent discussions with the Applicant, the Soecial Inspec. ion Team was informed that U-bolts are not used on two way rigid suptorts
,' for pipe sizes larger than 6 inches. This was verified by the A
Special Inspection Team. The restricted use of U-bolts is a resJ1:
- of the relatively low lateral load cacability of U-bolts for piping a larger than 6 inches.. The Applicant has shown in a Gibbs & Hili
~
study on lateral stiffness in U-bolt attachments that the lateral stiffniss values of U-bolts for pipe sizes 5-inch and under are . ! comparable to the generic stiffness values used in the pipe stress - analysis. Thus, for small size piping, the Applicant's use of the i generic stiffness values in the pipe stress analysis resulted in realistic supcort design loads and pipe stress values. The succort design groups are instructed to verify that the loads on the secoort are within the U-bolt manufacturer's allowable loads during One as-built verification program. For these reasons the Special Inspection Team found the Applicant's design practices for using U-bolts on two-way rigid supports acceptable. This concern is resolved. l { Relative to the second item of Mr. Ooyle's concerns, the Scecial Inspection Team determined by discussion with cognizant engineers that U-Yolt deformations have not t.een included by the Acolicant in its support deflection calculations. As noted above, the lateral loads on two way. rigid supports are verified to be within the U-bolt manufacturer's allowable lateral loads. Similarly, the loads in the l principal direction are verified to be witnin the U-bolt manufac-i
. .. -.. . . ~....._.. -. - - . .
. . . * -. u . ~ . = -- a ... w a . ~... .. .. ... r ... .:..:.:-- u - . i.w v ~:~.5-32 6
*urer's allowable loads during the As-Suilt Verification Program.
4 Therefore, the Special Inspection Team found the failure to include 3 the deflection of U-bolts in the support deflection analysis to be inconsequential since the deflection of the U-bolt is limited to such
! small (elastic) movement that its contribution to the support deflec-tion analysis is minor. This concern is resolved.
.a 3
Regarding the preloading stresses in item 3(a) of Mr. Doyle's concern, the Special Inspection Team determined that the Brown & Root Design Change Notice (DCN) Number 1, dated 10/8/82, to Construction Procedure No. 35-1195-CPM 9.10.Rev. 8 provides additional requirements to para-graph 3. 3. 2, " Threaded Items ," for U-bol ts. It states:
. "When U-bolts are specified on the design document as not having
, any clearances, the U-bolt shall be snug tight so that the .
. U-bolt cannot be moved by hand....
i 3 , Snug tight is defined as the tightness attained by a few impacts
; on an impact wrench or the full effort of a man using an ordinary.
spud wrench." 4 The preloading stresses associated with those procedures are common N to industry use of threaded fasteners (a proven method of holding
) structures together), although difficult to assess. In acdition, paragraph 4.2.6 of the Construction Procedure requires the following
, inspection:
- "The U-bolt shall be visually inspected by QC for cracks, melted spots and excessive deformation. Any one of these conditions
, shall be cause for rejection. This inspection shall be included
. in the final inspection of the hanger."
a
? The Special Inspection Team found this inspection procedure to be
[ sufficient to insure that preloading stresses are within acceptacle limits. In subsequent discussions, the Applicant informed the .' Special Inspection Team that the U-bolts will be field verified to confirm that they are properly tightaned. Further, that the walkdown inspection conducted prior to preoperational testing routinely checks for the proper installation of U-colts. With respect to t*4 constraint of differential thermal expansion aspects of items 3(a) and 3(b) in the third of Mr. Ocyle's concerns, the Special Inspection Team wocld note that differential thermal expansion effects are limited to the case of uninsula'ted oiping. In the case of insulated piping (e.g., main stea'm, feedwater, and
. residual-heat removal piping), the temperature differences between the U-bo1 and the pipe will ce negligible because the U-ooit is in thermai contact with the oipe and the insulation is installed over l
~
both the U-bolt anc the pipe. A review of :ne design temoeratures and pipe sizes of uninsulated piping by the Special Inspection Team indicated that the maximum radial growth of the piping is expected to 7..
,2;. . .. ;. .
_.- ._-....a..--....-:. ,_
- _ . _: --- w ::nc y.
q . j 33 4 a
. be less than 1/32 inch. Since the -U-bolt is in contact witn the pipe
; and will heat up to some extent, a maximum differential radial growth between the U-bolt and the pipe of about 1/64 inch seems reasonable.
Assuming the U-bolt is as stiff as the pipe, the effective maximum
- radial constraint will be in the order of 1/128 inch (0.008 inch). ,
Since the U-bolt stresses and pipe stresses associated with 1/128 inch l 0 - radial constraint are negligible, differential thermal expansion 1
'y effects in uninsulated piping are negligible. Alternately, since the !
maximum temperature differential between the U-bolt and pipe in
. uninsulated piping is expected to be less than 50 degrees Fahrenheit, )
calculations performed by the Special Inspection Team indicated that j
.d the associated secondary stresses and loads are negligible enlative *
, to ASME Code allowables. Further, the U-bolt is nomally provided 4 with a 1/16 inch diametrical gap on the pipe to facilitate its installation. Even after cinching down, there is not full cir-i cumferential contact between the U-bolt and the pipe. This will also j alleviate differential thermal expansion effects. The Special
- . Inspection Team concluded that the differential thermal expansion aspects of Mr. Ooyle's concern are resolved.
The Special Inspection Team would note that Mr. Doyle also expressed
, a similar concern regarding differential thermal expansion effects in
} box frame supports with zero clearances. Relative to this concern
# the Special Inspection Team understood through initial discussions with relevant cognizant engineers that it was an uncocumented Gibbs &
Hill design recomeendation, where U-bolts or box frames are in direct
~
? contact with the supported pipe, that clearances be provided between
, the U-bolts or box frames and the supported pipe only if the diametrical
~
growth of the pipe exceeds 1/32 inch at design temperatures. The
? Special Inspection Team determined that of the three pipe support design groups (PSE, ITT-Grinnell and NPSI) only ITT-Grinnell has
. documented guidelines which incorporate the Gibbs & Hill recommenda-
*/
tion. The Spedial Inspection Team was informed that the two remain-ing groups also follow the ITT-Grinnell guidelines. In sucsequent discussions with the Applicant, the Special Inspection Team was informed that the above 1/32-inch design guideline was amplicable I only to box frames. It was not applied to U-bolts. For the reasons discussed above tne Soecial Inspection Team agrees that such a design guideline is not needed for U-bolts. With respect to tne box frames, the Applicant statad that box frames were used only on low temperature systems (e.g., service water, com-ponent cooling water). This was verified by the Special Inspectio.n Team. Because of this, the diametrical expansions of the pipes are
~
generally of the order of 1/64 inch. Assuming zero clearance, since
- the pipe-and the frame are of equal stiffness, the deflection (bow) in the frame would be approximately 1/128 inch (0.008 inen). The pipe. wall would also be pusned inward an equal amount. Similar to the case of uninsulated piping discussed above, stresses in the box frame and pipe due to constraints of 1/128 inch of differential l
thermal expansion are negligible.
. :; :; p : ... :. .-...,. - . ; ,.. ~ c.. : __ - - a. . -. w.~ w.. . . . ..a . . . . , . . . .. ..u.;. 4 . j . 4 . 1
'! 34 k
q i The Applicant also stated that although the design may specify a zero clearance (no gap) for a box frame, construction techniques often result in a diametrical gap of up to 1/32 inch. Thus, in practice b)$ ~ the frame often provides a gap which further alleviates the con-3 straint. Based on the above, th,e Special Inspection Team concludes q- that Mr. Doyle's concerns about the pipe stresses caused by con-4 straint of the thermal expansion of the pipe by box frames is without d foundation. This concern is resolved. l '
- With respect to ,the instability aspects, item 3(c), in the third of 3 Mr. Doyle's concerns, since the stresses due to preloading. and f differential thermal expansion effects are expected to be negligible, F loosening of the U-bolts due to thermal cycling will be precluded.
The Special Inspection Team concludes this concern is resolved. 7
'i Mr. Ocyle's concern about the . restraint by U-bolts of lateral mov'e-3 ment of the pipe due to thermal expansion at one way restraint . , points, and his concern about the preloading stresses have also been I identified 'in the course of the Applicant's normal review program and f these problems have been rectified. Mr. Doyle's other concerns about 1 the use of U-bolts have been found to be without a valid technical basis. . 's .
4, h. Leading Due to the Seismic Acceleration of the Pice Succort Structure r Mr. Walsh expressed a concern regarding the inclusion of seismic' acceleration of th'e pipe supports in the STRUOL analysis. (Walsh'
~
1 . Supplemental Testimony, page 1, Case Exhibit 659H). Mr. Ooyle
; expressed a similar concern regarding the effect on pipe stressas of
't. the loads imposed by the pipe supports during a seismic event (Ooyle Deposition, Attachment 12A). In addition, Mr. Walsh made assumptions regarding the natural frequency of some supports and concluded that 3 . the supports would fail. (Walsh Supplemental Testimony, Tr. 3100) y In response to the expressed concerns, the, Special Inspection Team , q reviewed the following items. : (1) Review of Analvses to Oetermine the Effect of Seismic Acceleration Loacs All small bore piping supports were designed by PSE. There are over 8,000 such hangers and supports in Unit 1 and about 7,000 in Unit 2. The large bore piping supports were primarily designed by ITT-Grinnell and NPSI. There are over 15,000 such supports , and hangers in Unit 1 and over 11,000 in Unit 2. Seismic
. accelerations are considered.by PSE in the design of the small bore piping. For large bore pioing suoports the seismic acceler-atfon load of the sucports tnemselves were consicered by NPSI
- and ITT-Grinnell to be relatively low in comparision to tne design loads imposed by the piping. Therefore, the seismic acceleration load of the support was not included in the pice sucport design process. To confirm and validate that assump-
- .. -- . .. - -- ---. ~- -- _.
/;. _l w - s..: . xw ::.12Q5NO3%
- Q -- - - --- : h: .* = AE "
..' ., 3 1
- h
- f. .
1 j 35 1 I q tion, the Applicant randomly selected approximately 400 supports. 1 4 From this random sample, which included designs by ITT-Grinnell, i NPSI, and PSE, a selection of 23 " worst case" supports was made for detailed analysis by the Applicant. These 23 supports were
- those in which the seismic acceleration loads were likely to be j most significant due to the configuration of these supports.
2- They included unbraced cantilevers, large frames braced in one direction, and large structures with relatively small pipe loads. These 23 supports were reanalyzed in detail cy the pipe y support design groups to consider the effect of seismic acceler-J ation loads on the support design. It was found that the
') stresses in the most highly stressed members of the supports I were well within allowable limits of the ASME Code Section III
_ and in a majority of cases, the additional loads imoosed by the seismic acceleration of the support frames were negligible.
. A separate, reanalysis was performed by NPSI on 13 su'pports in I
. which the seismic acceleration lands had been neglected by NPSI 1 in the original design. These supports are considered to be
[ worst cases from the standpoint of seismic acceleration loads.
. The conclusions from this study are essentially the same as stated above, i.e., the seismic acceleration loads'are
'{
, negligible.
$ The Special Inspection Team evaluated the calculations performed f by both PSE and NPSI in detail. The review included the mode'-
ing techniques, design criteria, analytical assumptions, computar
))
, programs, and hand calculations. Discussions were held with
',; individuals in. the PSE, ITT-Grinnell, and NPSI design groups wne e routinely performed the calculations and were involved in the i design process. On the basis of the Soecial Inspection Team's
_ ,.! review of the Applicant's reanalysis 'of pipe support designs, y the Special Inspection Team concurs with the Applicant's con- , .. clusions that: (1) in a majority of cases additional loads . d resulting from seismic acceleration of support frames are J negligible, and (2) in no case will the inclusion of the leads
-+'~- --
due to seismic acceleration of the support structure result in overstressing the supoort structure. The Special Inspection
-Team consicers this concern to be resolved.
i (2) Review of Desicn Criteria to Maintain Rioidity of the Succorts i Mr. Walsh assumed that the support would be.excitad at the frequency corresponding to the peak acceleration of the floor i response spectra (Walsn cross-examination Tr. 3100). This is an erfeneous assumption and results in unrealistic seismic acceler-
; atton load predictions. Actually, tne natural frequency of the
[ - support is much higher than the frequency at which the peak acceleration occurs in the floor response spectrum. When the seismic acceleration of the support is determined at its correct
,. - 7 w
f.:...a. a .*.K jE G & M M ".:! % ' - * . -- - - 2.. - "% s . M - r.M a ' *W
- I 9 *.*
= ,
3 . 1 4, 36 .
]< natural frequency, the resulting. stresses are founc to be within ASME Code allowables.
C The NRC Inspector reviewed the design criteria adopted by PSE,
]j ITT-Grinnell and NPSI to ensure rigidity of the supports. All '
j three design groups limit the deflection of the support to 1/16 inch under service level B loading condition. This limitation i, should ensure a rigid design. The Applicant is providing a . i study demonstrating the adequacy of its guidelines to as,sure a J rigid support design (see Paragrapn 3.j). For rigic frames the d seismic acceleration loads would remain low. O
] In addition to the 1/16-inch caflection criterion, NPSI provices out-of plane bracing to ensure staoility and rigidity. Vertical bracing of the members is provided if the member or frame ,
,1 overhangs more than 5 feet 6 inches in a horizonal direction.
g Similarly, horizontal bracing is provided if the member or frame
, extends more than 8 feet in elevation. The PSE and ITT-Grinnell d design groups do not have specific guidelines but provide j- bracing for rigidity based on engineering judgement in their normal design process. The Special Inspection Team's review of 100 randomly selected supports determined that: (1) the NPSI j . criteria have been adhered to; (2) deflection limits have been y maintained in the support design; and (3) the seismic accelera-
?
tion loads of the supports are likely to remain negligible. The
. Special Inspection Team concluded that this concern does not present a safety issue and considers the concern resolved. ' 'i (3) Effect of the Succort Loads On the Pice Stressas Ourinc a .j Seismic Event a
is - 0 Mr. Ooyle expressed a concern that certain types of supports are t 'i attached to the wall in such a manner that the weight of the l
! support would transmit some load to the pipe, and that the '; effects of this additional loading ha've not been adequately ~
- .3 considered in the stress analysis of the piping. -(Ocyle Deposi-tion Attachment 12A). The weight of these supports acts at some
~
l distance from the axis of the pipe. This eccentricity in the support weight may introduce some torsional stresses in the pipe in addition to the stresses due to the support deacweight. Although not explicitly stated by Mr. Ooyle in his concerns, this torsional effect was also evaluated by the Soecial Inspec-tion Team and is discussed in the following paragraphs. Pipe restraints of the type cited by Mr. Doyle function by restraining pipe motion along the axis of the snubbers (Ocyle Depostion Attachments 12E througn 12N). The .<eight of the support structure may provide some accitional loading on tne pipe at the location of the restraint. The Special Inspection Team reviewed the analysis of these suoports anc the procedures used by the Applicant to include tne effect of tne support load
* * = , e emep *. ,.******& -**,9**-* *" * ' ' "
_ , = o *U M *
. . . . . ~ - . . . . . . . .. .. ~- - . - . . .- .
,:: n d k r % i *
~.'.' ;-..:.... a - - a.. u=w. me, . . .. a : a k~ --a . - . ! - a M .*"
i' . ;
/
u 37 5 0 i' 3 on the pipe. The Special Inspection Team found that an assess-I ment of this contribution to the piping load is made on a k case-by-case basis by botn the Gibos & Hill and Westinghouse 6 pipe stress analysis groues, and it is,added to the piping loads 1 during the stress analysis of the piping run, if the contribu-
$ tion is. considered significant. The following table summarizes L the treatment of the weight of the support in the piping stress
] analysis for the specific examples cited by Mr. Ocyle.
4 Organization Support Support Lead .
$ Pipe Responsible Load on in Piping y Succort No. Dia. for Analysis pice Analysis A
y . SI-1-120-004-C52K 10" Westinghouse 130 Neglected 4 .
}
-)
SI-1-104-008-C52K 10" Westinghouse 100 Neglected 9 . SI-1-031-704-A32R 12" Gibbs & Hill 30 Neglected M k MS-1-003-013-C72K 32" Gibbs & Hill 974 Considered 3 d MS-1-003-009-C72X 32" Gibbs & Hill 2015 Considered II a y, she Special Inspection Team also. investigated torsional effect j of the eccentric support load on the pipe stresses. Support j Nos. MS-1-003-013-C72K, MS-1-003-009-C72K, CC-1-043-015-A43K , 5 (Ooyle exhibit 13LL) were selected for tnis purpose. These are , ih _ considered to be worst case configurations from tne standpoint 3 of torsional effects. Calculations performed by the Special
; Inspection Team indicate that the increase in pipe stresses due y to the torsional loading was less than one percent. Based on
~~~
y~~~ this investigation, the Special Inspection Team found the increase S in the pipe stresses due to torsional effects of eccentric
,y support loads to be negligible. ; - Q)
The Special Inspection Team fcund that in, practice the Acplicant includes the weight of the support in the pipe stress analysis if the support weight exceeds a small percentage of the succort pipe weight. On that basis, the Special Inspection Team found that the Applicant's procedure of adding the weight of the suo-port to the piping weight, on a case-by-case basis, to account for the effect of the support. load on the pipe stresses is
'. acceptable and does not represent any safety concern. In addi-il tien, based upon the negligible increase in pipe stresses due 1 -
to ' torsional effects, the Special Inspection Team concluced tnat torsional effects do not reoresent any safety concern. Mr. Ooyle's concern is considered to be resolved. f a me - e, 3 - 2^- %** ** '
' '7r , ::.:.. ::~ .
.a."'" :- - -.:.: . % .;.:-.. J n .~. ~ . . :.; ~ - u - m- -. -. . . - - *~ di "
h . f ~ 3 . .
}
r 38 1
.* Mr. Walsh's concern regarding a need to include seismic accelerations
% in the pipe support design analysis and Mr. Walsh's analysis project-4 ing failure of the supports under seismic loads are without valid y technical bases. Mr. Doyle's concern that the pipe stress analysis
. ;] did not adequately consider the added weight of the support was also
', without a valid technical basis.
- f. Moment Restraints and local Pioe Stress Oue to Welded Stanchiens on
;$. Pipes Mr. Walsh and Mr. Doyle expressed a concern that the effects due to *
<j welded stanchions on main steam, containment spray and feedwater
( piping have not been included in the as-built piping stress analysis. 3 These effects are: (1) moment restraints introduced in the piping
~
1 system and (2) local stresses in the pipe wall. Examples identified 1 by Mr. Walsh are supports CT-1-024-004-522K and FW-1-096-704-C62K d (Walsh Supplemental Testimony, CASE Exhibit 668, p. 1; Item.2, CASE
?
i Exhibit 668A). Examples identified by Mr. Doyle are supports CT-1-008-006-522K and MS-1-003-009-C72K (Ooyle Deposition, Attach-i ments 11LL-11NN and 12N-12P, respectively). h The Appifcant has stated that, regarditig welding of stanchions to
] pipes by NPSI, ITT-Grinnell and PSE, th6 final.as-built piping and
'g support verification program will assure that the actual support con-figurations will be taken into account (Applicant's Exhibit 142,
, Pages 25-25). The Applicant has further stated that y " stresses at pipe /we,lded attachment interfaces will be qualified q by Gibbs & Hill to the as-built loads during the as-built stress 3
analysis." (0.M. Rencher to ITT, NPSI TSDRE's, Stress Analysis of Weided Attachments, TS8R #V92, April 7, 1982). l.
; The Special Inspection Team conducted a review of the Applicant's M As-Built Verification Program at Gibbs & Hill in New York on October 27, 9
A 1982 to verify the adequacy of the program in addressing the concern j expressed by Mr. Walsh and Mr. Doyle. The review showed that the l-' verification program requires consideration of: (1) restraint i characteristics of "as-installed" (as-built) suoports; and l (2) stresses due to welded attachments in combination with aopec-l priate AOLPIPE computer program piping stresses. To confirm that the Applicant's As-Built Verification program is , being adequately implemented in this regard, the Soecial Inspection i Team selected for detailed review Gibbs & Hill. Stress Proolem AB-1-03
,- dated August 23, 1982, for the main steamline No. 3 inside contain-
. ment. The Special Inspection Team selected this problem because the main steamline is the most critical of the three lines identified oy Mes s rse.' Walsh and Doyle. The review showed that stancnions are l welded to the 32-inch diameter main steam line pipe at the following snubber supports'and that the pipe stress problem included the
, analysis of local pipe stresses due to these stanchions:
6
, - - - . , .~ ,~.- . - ~ ~ . ~ - - - - -' ~~
_~3
x 2
. . .: . , .; = .. . _ . . .u ' . > . ew ' "
. L .. .&-~x . a. .. <. w: a.ws. ~au--a a~- =~2.. :.
i -
- r -
7
- 7 ,.
- 3 39 ,
.) , .
] .s -
r MS-1-03-005-C72K s j MS-1-03-00?-C72K y , G MS-1-03-009-C72K (Doyle Exhibit 12N-12P) x. F; MS-1-03-010-C72K .
'~
i MS-1-03-014-C72K > a . . _' f" All= but the' first of these ? supports utilize dual snubber designs with
~
. . _ distances of approximately 5 feet between snubbers.i With' respect to W . moment restraints:on the main steam piping, tho Special-Inspe,ction 5 Team founfthat no moment restraints were considered at'these supports j in tne . Applicant's piping stress anitlysis. ,Tne Special Inspection Team perfarmed calculations base 4 on the snubber or translational stiffness of Table 3.1-1 of the Gibbs & Hill Specification 2323-MS-200
, and the 5-feet distance mentioned above. These calculations gave
~
,] stiffnesses of rotational restraint of the same order of magnitudp (2 4
to 7 times) as the generic rotational stiffnesses used by Gibbs & [ Hill in the pipe stress analyses. The Special Inspection Team con-O cluded that the rotational stiffness associated with these designs should have'been included in the piping stress analysis. Subsequent '
.i discussions with the' Applicant indicated that this rotational restraint
. had. also been ifontified during the Applicant's normal design review t and that the pipe stress ar;tlys1s was being modified to consider this
..- rotational restraint. The,Special Inspection Team reviewed the
). proposed method of analysis (" Minutes of discussion at the Meeting o between G&H and NPSI on March 17, 1982") and concluded that the
# ~
method of modeling the rotational restraint and the attendant loads
.' on the snubbers was acceptable. .Since the Applicant is including' 4e this rotational restraint in the pipe stress analysis, the Special Inspection Team found the concern on mcment restraints introduced in 9 7 the piping system to be resolved.
1 With respect to local stresse's, the Special Inspection Team found t- that the Applicant evaluated local pipe stress effects in their . As-Built Verification Program where applicable, due to radial and , f . shear loads and moments. The Applicant utilized for its local stress ;-
$ evaluations the CYLN0Z 2 computer program. The CYLN0Z 2 computer program was developed by Franklin Institute on the basis of Welding i' '
Research Council (WRC) Sulletin No.107, " Local Stresses in Spherical and Cylindrical Shells due to External Leadings," August 1975. The Special Inspection Team concluded that the use of the CYLN0Z 2 , computer program is an acceptable method of analyzing local stresses. The Special Inspection Team also determined that the Applicant's
' calculated local pipe stresses'were comoined with internal pressure and' AOLPIPE bending stresses at these sucport locations in accorcance s -
with the criteria in. Equations S, 9, and 11 of NC-3650 of tne ASME ( Code Section III, Subsection NC. The criteria were satisfied at all five support locations on tne main steam pipe. L 4 5 e9e - _ ease e .- .
.e j. =s. ok .
~a-:.-l. ... v . ;.::::n;..: .z..
. . : .u s: _ .- ~......_-... .... . . - . . . . . . . . . . - . . .
u t .- '. a ,
~
, 40 b~
is 3 The Special Inspection Team noted however that differential thermal
') expansion effects between the insulated main steam pipe and the $ uninsulated structural steel support structure were not included in d the local stress evaluations for support No. MS-1-003-009-C72K. '1: These differential thermal expansion effects should be considered.
j In subsequent discussions, the Applicant stated that it had in fact i} considered this effect and determined that the resultant stresses 3 were acceptable. The Applicant has agreed to provide the Special d Inspection Team with its analysis. The Special Inspection Team will
@y verify the acceptaoility of this analysis (0 pen Itam No. 50-445/8225-5).
On this basis, the Special Inspection Team concludes that the concern
% expressed by Mr. Walsh and Mr. Ooyle is being adequately addressed by p the Applicant's As-Built Verification Program.
4
- j. Deflections and 1.ocal Stresses in pioe Sucoort Structures
_Q n Q Mr. Doyle has expressed concerns about excessive deflections and
? -
uncalculated local stresses at locations where brackets are attached to plates and other members of the pipe support structure. Mr. Ocyle 3 alleges that bracket loads cause local deflections which are. not
, included in the total displacement calculations for the support
.) hanger. He also alleges that localized stresses resulting from the lq bracket are not considered in the stress analysis of the support hanger. (Ooyle Deposition, pp. 159-172; Doyle Deposition, Attachment
'j 11A). Specifically, the following supports were identified as being
(.; examples of these concerns:
$q (1) Support CC-2-008-709-A43K (Ooyle Deposition, Attachments 11FF thru 1111).
11 j (2) Support CC-1-023-034-533R (Ooyle Deposition, Attachment 4G-4H).
'Y 4
y (3) Support CC-1-107-008-E23R (Ooyle Deposition, Attachment 11TT).
?
3 (4) Support CS-1-239-007-A42R (Ooyle Deposition, Attachment 1300
- 4 thru 13GG).
In computing the response of a piping system to comclex loading ccm-binations such as those which include a seismic event, it is important to assure that piping supports are sufficiently stiff so that they do not adversely affect the response of the piping system. The Apoli-cant uses generic stiffness values in its calculations of oiping system response. The use of generic stiffness values is common
, practice and is acceptable provided that the generic stiffnesses adequate]y represent the stiffness of the installed supports. The Aoplicant and its piping analyst, Gibos & Hill, indicated that they believe that the use of their overall ceflection guideline of 1/15 inch' maximum deflection under service B condition loads will result in supports wnose stiffness is adequately conformed to the generic values used in the piping stress analysis. In discussions with the
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e 41
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4 a Applicant, the Special Inspection Team noted that in the aosence of 4 review of the particular supports, it was unclear that the 1/16 inch j deflection guideline in fact results in support stiffness comparable j to the generic stiffness used in the piping stress analysis. The , l f" Applicant agreed to provide a study demonstrating that supports designed in acco: dance with Applicant's criteria and guidelines have j sufficient stiffness to assure that they do not adversely affect the j response of the piping systam. This matter remains unresolved.
; For the Component Cooling " Water support No. CC-1-107-008-E23is (Doyle j Deposition, Attachment 11TT), Mr. Ooyle alleged that the displacement i of the support will exceed the design guideline of 1/16 inch because i a 1-inch plate will allow rotation which has not been computad. This
_: support has been " vendor certified" by ITT-Grinnell to satisfy the W deflection guideline of 1/16-inch maximum deflection. A review
- by the Special Inspection Team of the original design calculation's
, showed that the deflection calculation did not include the potential y rotation of the plate as alleged by Mr. Ocyle. Subsequently, the
, Applicant tested the support to determine the actual deflection under service level B loads. The actual deflection was found to be less than 1/16 inch. The test has shown that the potential rotational effect alleged by Mr. Doyle does not result in excessive deflection.
However, since the actual stiffness of this support was founc to be about 1/8 of the generic value used in the piping analysis, the Applicant has been requested to rerun the piping stress problem with the actual stiffness value and to provide a report of its results. j . This matter remains unresolved. In component cooling water support No. CC-2-008-709-A43K a 16-inch diameter stub pipe is welded at the elbow of a 24-inch diameter component cooling water piping. (Ocyle Deposition, Attachments 11F g thru 11II). A 1/2-inch thick circular cap is welded at the end of
- the 16-inch diameter stub pipe which in turn has a bracket welded to it. Mr. Doyle alleges that the displacement and local stresses in :
, the 1/2-inch thick plate exceed allowables under application of the 11.9-kilopounds service level C icad. He further alleges that the 3/16-inch weld attaching the bracket to the center of the plate is overstressed. Although this support had not been vendor certif'ed at the time of the inscection, the Special Inspection Team made ci.lcula-tions which indicated that the maximum deflection may exceed the Applicant's 1/16-inch maximum deflection guideline, but the 3/16-inch weld was not found to be overstressed. The Special Inspection Team reviewed some preliminary calculations provided by the Aoplicant covering the displacements and local stresses for this sucport. A
, numerical error was uncovered by the Soecial Inspection Team in the Apolicant's preliminary calculations, which resulted in the under-estimation of bending stress in one memoer of this support and could result in an overstress condition. In discussions the Applicant indicated that its subsequent review has also identified an over-stress condition in this support which would be rectified as part of
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y . > Q . 1 3 42 i 1 [ its normal design iteration process. The Applicant will provide a y status report on the status of this supcort design. The corrective 2 action will be verified by the NRC staff in a follow-on inspection as y part of its construction inspection program (0 pen Item No. 50-446/3214-4). 2 i For the Component Cooling Water Support No. CC-1-028-034-533R, Mr. Doyle alleged that the stress in the web of the W6 x 12 beam 3 d exceeds the ultimate stress for the material of the beam (Ooyle Deposition, Attachment 4G-4H and 113). The Special Insoection Team 2 found that the calculations performed by Mr. Doyle are in error. In
, determining the stress in the web of the W6 x 12 beam of this support, Mr. Ooyle has erroneously used the cube of the web thickness instead
; of the square of the thickness, res.ulting in unrealistic stress
'- values. Mr. Doyle's concern about this support is considered resolvec.
2 In chemical volume and control system support No. CS-1-239-007-A42R, j Mr. Doyle alleges that the deflection guideline of 1/16 inch is ?. exceeded. The Special Inspection Team determined that the plate thickness in this support was initially specified to be 1 inch. A Component Modification Card (CMC No. 58004) dated June 11, 1982 was 1 issued to revise the plate thickness to 1.5 inches. With the revised
- thickness, the 1/16-inch maximum deflection guideline and the ASME I _,
Section III Subsection NF Code requirements are satisfied. The .
~-
concerns relative to this support are considered resolved. Ouring its inspection, the Special Inspection Team noted that there did not appear to be clear guidelines for specifically considering local stresses resulting from bracket leads, nor do they appear to have clear guidelines for considering deflection contributions from
. localized effects. In discussions with the Special Inspection Team,
' the Applicant stated that even though there are no explicit guide-lines, it is routine practice for the support reviewers of all three pipe support design groups to censider these local effects in the
- design of pipe supports and in the review of such designs. The .
. Special Inspection Team has examined examples of cases in which local ;
effects have been considered in its inspection of vendor certified supports described in Paragraph 4. In summary, Mr. Coyle's concerns about excessive dsfisction3 in cer-tain supports had in two instances also been identified by the Appli-cant's design review program. In one. case the problem has alreacy been rectified and in the other the preolem is to be rectified by redesign. Mr. Ocyle's concerns in two other instances have not been
.' substantiated. Thus, the concer.1s raised by Mr. Ooyle are resolved.
The foll,owing two additional studies discussed above relating to succort. stiffness which the Applicant has agreed to provice remai n i unresolved. J h e e e a 4 -M &M P . 3 e
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y 1 a) A study providing assurance that the Applicant's design critaria i
!.' and guidelines provide sufficient stiffness to the supports ; $j (Unresolved Item Nos. 50-445/8226-6 and 50-446/8214-5).
7 b) A pipe stress analysis providing assurance that support i No. CC-1-107-008-E23R has sufficient stiffness to perform
$ satisfactorily (Unresolved Item No. 50-445/8226-7).
- k. Consideration of Friction loads Mr. Walsh expressed a concern regarding the consideration of fric-
< tional loads between the pipe steel and supports during thermal '.' expansion. He stated that when considering the coefficient of friction in thermal expansion ITT-Grinnell uses 30*. coefficient of friction of the deadicad plus thermal load of the pipe, while NPSI uses 45 percant of the deadicad plus thermal plus CBE load, and PSE "only considers it (frictional loads) when they want to." (Walsh l Supplemental Testimony, p.3, CASE Exhibit 659E).
1 The Special Inspection Team has reviewed the guidelines used by the three design groups (PSE, ITT-Grinnell and NPSI). The NPSI guide-lines specify the following values of the coefficient of friction. Coefficient of Friction Condition 0.33 For steel contact in each of two - directions. - O.45 For steel contact in any one
, direction to simplify calculations.
The frictional force is defined by NPSI as the frictional coefficient
! multiplied by the deadweight plus thermal plus CBE loads. If the pipe movement, ap, is larger than the def1'ection of the structure, ;
as, due to the full friction force (P), the full value of the fric-tion force on the support structure is utilited in designing One support. If Ao is less than as (due to the full friction force, (P)] a reduced value of the friction force is used on the succort structure. This reduced value of the friction force is (lo/as) x P. ITT-Grinnell guidelines specify the following coefficient of friction: Coefficient of Friction Condition 0.33 Steel to stael ITT-Grin'nell defines the frictional force as the frictional coeffi-cient multiplied by the deadload plus thermal expansion loads and this is to be considered only for pipe movements in excess of 1/15 a inch. If displacement of the structure 'under full frictional loads
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" exceeds thermal displacement, the friction force may be derated by the structure's spring rate multiplied by the thermal displacement.
3 J PSE guidelines specify the'following coefficient of friction: I Coefficient of Friction Condition
.;; 0.33 Steel to steel The PSE guidelines state that friction loads shall be calculated for thermal and deadweight loads, and applied in the direction in wnich the thermal movement of the pipe is unrestrained on rigid frames. -
For thermal movements of 1/16 inch or less, frictional loading is to be ig'ored. n Ail three support design groups provide adequate bracing to maintain
- ' rigidity and structural integrity of the supports during a seismic event. Therefore, it is not necessary to consider seismic loads in the determination of frictional loads. On this basis, the Special Inspection Team found the load combinations (deacweight plus thermal)
; used by ITT-Grinnell and PSE for computing frictional loads to be acceptable. The inclusion of the CBE in these load combinations by NPSI results in a more conservative estimate of the frictional loads and is acceptable. .
------~~
In addition to' a differe'ce n in the load ccmbinations used by the _ three support design groups, there is a difference in the coefficient of friction used by the three support design groups. The Special Inspection Team found the use of a coefficient of friction value of 0.33 for steel to steel contact a commonly used value and its use by PSE and ITT-Grinnell is acceptable for steel to steel contact. As to the difference in friction coefficient, the higher friction value used by NPSI is more conservative and is also acceptable. ~
-c l?ie ' seismic response of the piping system is highly insensitive to ~
~ variations in frictional loads. Therefore, any differences in the ;
calculatad frictional loads arising out of the use of differing
- friction parameters by the three suoport design groups will not have a significant effect on the pipe stress analysis.
The Special Inspection Team concludes that the frictional load design parameters utilited by the Applicant are acceptable. Mr. Walsh's concern is resolved.
,_ l. Consideration of Kick-Leads .
, Mr. Ooyls expressed a concern that the Applicant was not considering
" kick-loads" in the design of the plant piping (Doyle Decosition Attachment 11RR).
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i C 45 0 o A 4 3 The Special Inspection Team found that Gibbs & Hill included the i directions of supports and hence the " kick-load" force component in the pipe stress problem for the main steam and feedwater ifnes if the 3 a support as-built misalignment was 5 degrees or more. The Special 1 Inspection Team found- that acceptable for the main steam and feed-l water lines. In subsequent discussions with the Appifcant, the
; Special Inspection Team was informed that a similar procedure was
- j employed for all other Class 2 and 3 piping. This procedure was
; verified by the .Special Inspection Team's review of a stress problem for the baron recycle system. This concern is resolved.- Mr. Coyle's t concern was found to be incorrect.
I j m. Modelinq of Wide Flange Members As Infinitely Riqid In Torsion
..; Mr. Doyle expressed a concern with respect to the ITT-Grinnell modeling of wide flange members using large torsional rigidity vaiues 3; in ITT-Grinnell Procedure No. RP-2, "STRUOL Modeling for Structures i Subjected to Web Bending (Ocyle Deposition, pp.180-81, CASE Exhibit 6698). Although.Mr. Doyle's concern was with the use of a torsional constant of 10,000 inches 4 for wide flange momeers, he states, "I don't recall what that was used for." The Applicant responded to Mr. Doyle's concern by stating that large torsional rigidity values ~
are intended to maximize the torsional moment which in turn is utilized to perform a conservative evaluation of torsional stresses. (Applicant's Exhibit 142F, " Supplemental Testimony of Kenneth L. Scheppele, Roger F. Reedy, Peter S.Y. Chang, John C. Finneran, Jr. , and Gary Krishnan Regarding Doyle Allegations"). - The NRC Special Inspection Team reviewed ITT-Grinnell Procedure RP-2. The review determined that the procedure provides guidelines for STRUOL modeling when investigating web bending in wide flange struc-tural members subject to certain well-defined support configurations. l The guidelines require that the torsional constant (polar moment of inertia) of the wide flange members in which web bending is to be
- evaluated be assumed to be 10,000 inches 4.- Torsional stresses in the .
members are subsequently calculated from the results of,the moment ' analyses based upon the above assumption. On page 11 of the proce-dure it statas, "In evaluating stresses one must be careful to reali:e which output values are real and wnich are fictitious cecause of the way that properties were assigned." The basic torsional analysis of a wide flange member initially
, involves calculation of the torsional moment to be resistad. This
. moment calculation is determined by. multiplying the rigidity of the-4 . member with the maxir" angle of rotation to be resisted. Torsional shear st,resses are subsequently comoutej by dividing the product of the torgional moment and the member thickness by the polar moment of i ne rti a. Since both calculations are independent of eacn other, tne initial calculation of torsional moment may assume a large value of torsional rigidity and thus result in a corresponding large moment.
The subsequent calculation of torsional shear stresses will enerefore e , a.e e po eme.g e ahpulmag gquespee =
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O d fj result in a more conservative result. The Special Inspection Team G found that the torsional moments evaluated on the bases of tne
$ torsional constant of 10,000 inches 4 are conservative. Subsequent M calculations of shearing stresses utilize the correct torsional y constant values published by the American Institute of Steel Con-3 struction (AISC). The Special Inspection Team considers this design 3 approach to be valid and conservative.
A J Mr. Doyle's concern that ITT-Grinnell erroneously utilized large y n- torsional rigidity values in tn mocaling of wide flange memoers has y no technical merit. Large torsional rigidity values were employed by 5 ITT-Grinnell to maximize tarsional moments of wide flange members. Di Use of large torsional moment values in calculating the torsional 3 shear results in conservative stresses. ITT-Grinnell used correct j AJSC torsional rigidity values in the torsional shear calculations. S Mr. Doyle's concern is incorrect. This concern is resolved. d l 2j n. Effect of Cold-Formino On The Ductility of Tube Steel
,h During Mr.'Walsh's cross-examination of the Applicant's rebuttal witnesses, he presented a concern with respect to,the effect of cold-forming on the ductility of tube. steel (Tr. 5078). The Special
.. Inspection Team has addressed this general concern by performing a
' .4 literature review and identifying the results of tests conductec to
's .
quantify the effects of cold-forming. Although the cold-forming of structural steels will increase the
,j yield and ultimate strengths of the material, the relative magnitude ... .of the increases are not the same and therefore result in recuction d in the spread between the yield and ultimate strengths. This reduc- ; tion results in a decrease in the elongation capability.or ductility.
Ductility in a material is a desirable quality which represents its
,- ability to undergo plastic deformation prior to rupture. This Z ductility reduces the effects of stress concentrations and helps to ,
g achieve uniform load distribution by guaranteeing plastic stress .
- p redistribution. This plastic deformation mechanism is reifed upon in ;
, .a the design process to take into account any detrimental effects
._ resulting from secondary stresses. Ductility therefore provides relief of secondary stresses prior to the material reacning f ailure
~
i strain. The tube steel utilized in the design of pipe succorts is I designated as American Society for Testing and Materials (ASTM) A500-Grade B steel having a minimum ductility requirement excressed as minimum elongation in a 2-inch length of 23 percent. t i . The following papers, published in the American Society of Civil Engineer's Journal of the Structural .0ivision, dealing with tne effects 'of cold-forming on steel were reviewed as part of this j inspecyi'on:
- 1. " Structural .8ehavior of Thick Cold-Formed Steel Members," ty l W. W. Yu, V. A. S. Liu and W. M. McKinney, November, 197a.
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( l i 2. " Suggested Steel Ductility Requirements," by A. K. Dhalla anc G. Winter, j Feb ruary,1974
.:i
- 3. " Steel Ductility Measurements," by A. K. Dhalla, and G. Winter,
{ . February,1974.
.1 1 4. " Corner Properties of Cold-Fonned Steel Shapes," by K. W. Karren, j.
February,1967. [] 5. " Effects of Cold-Straining On Structural Sheet Steels," by A. Cha,jes, a S. J. Britvec, and G. Winter, April 1963. Q
] 6. " Effects of Cold-Forming On t.ight-Gage Steel Members," by K. W. Karren, y and G. Winter, February 1967. 'l i Ohalla and Winter have demonstrated that even if the ductility in a -
material is reduced to 3 percent, this value still would result in an
}: . acceptable level of ductility for structural perfcrmance. In their 1 paper (Reference.2), Dhalla and Winter stated, "An analytical study 4 of perforated and notched' plates in tension indicated that a unifor= ;; elongation greater than or equal to 3 percent appears ryecessary to plastify the critical cross section of members with such stress
- j concentrations and to achieve full net section strength."
2 Karren (Reference 4) has oetermined that the percentage clangation in - f 2 inches for corner specimens representative of cold-formed A 500 ( Grade B steel varies from 6 to 19 percent, de.;erding on the ratio of - , 9 the corner radius to thickness. Karren concludes that even though
.i the corners of cold-formed structural shapes will see a loss of ductility, the structural shape will remain functional. He states, 4 $j "The reduction in percentage alcngation as compared to that of the ;; virgin material varies from 20 percent to as much as 90 percent, but j--
permanent elongation even for the sharpest corners tested was in the
- j range of 5 to 10 percent, indicating considerable remaining ductility."
s .
) Other ASNE approved 'eaterials such as A513 Grade 1015CW,.have minimum -
- j
. elongations of less than 12 percent and are acceptable NF materials 9 for pipe supports. This fact also indicates that the reduction of ductility in the corners of A500 - Grade 3 material, although less than the specified 23 percent, does not necessarily rencer the tube
. steel shape nonductile.
Mr. Walsh's cencern that cold forming of A500-Grade B tube steel adversely affects its ductility has not been substantiated. A500-Urade B cold-4 formed tube steel'is sufficiently ductile to perform its design intent.
, The concern is. resolved.
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- o. Ooerating Condition loads Accear To Be In Error 4
Mr. Doyle stated a concern that emergency operating condition loads were smaller.than normal and upset operating condition loads on pipe supports. Support No. CS-1-235-067-C41K (Ooyle Attachment 87-8U) was identified on p. 130 of the Doyle Deposition in this regard. The drawing for that support indicates that the emergency operating condition load is 1030 lbs. and the normal and upset operating condition load is 1070 lbs. In general, normal and uoset operating condition loads are usually smaller than emergency operating condi-tion loads, which in turn are usually smaller than faulted operating condition loads. Thus, the Special Inspection Team interpreted Mr. Doyle's concern to relate to suspected computational errors in the Applicant's analysis. Examination of the drawings for support No. CS-1-235-067-C41X snow
, that: (1) the support is a seismic east-west restraint utiliting .a 3 .
mechanical snubber, and (2) the faulted operating condition load of 4 1040 lbs. is smaller than the normal and upset operating condition load of 1070 lbs. The Special Inspection Team reviewed the Westing-house seismic piping analysas using codes ADAYAPQ and ADAYAPS for Stress Problem 1-41, which includes support No. CS-1-235-067-C41K. This review verified these support load values were correctly cal-
~ ~
culated and that both the emergency and faulted operating conditions loads were smaller than the normal and upset operating condition load. 2 The Special Inspection, Team also found that the seismic analysis inputs for the upset operating condition were the Operating Sasis Earthquake (OBE) response spectra, and the corresponding inputs for 3 both the emergency and faulted operating condition were the Safe
.i Shutdown Earthquake (SSE) response spectra. Comparison of both of ; . these spectra show that: (1) for some periods, accelerations for the OBE horizontal spectra are greater than the corresponding SSE accel-erations; and (2) for all periods, the SSE accelerations are less :
than twice the corresponding OBE accelerations. The Special Inspec-tion Team concluded that these response spectra characteristics, together with the fact that the SSE damping value of 4 percent is twice the CBE damping value of 2 percent, lead to the condition expressed in Mr. Doyle's concern. This concition is not unexpected when seismic analyses are performed where actual " pense spectra and differing damping values for the 08E and SSE are used as analyses
. inputs.
Mr. Doyle's concern that the emergency condition loads are smaller than norgal and upset loads, in particular on sucport No. CS-1-235-067-C41K, is without a valid technical basis. The Aeolicant correctly used the CBE and SSE response spectra and damping values in tne seismic analysis, and the loadings for support No. CS-1-235-067-C41K are correct as shown on the drawings. The fact that the emergency and/or faulted operating condition loads are smaller than tne normai
, .f u.- -F =:.ww
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>a A w W n.;
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,, and upset operating condition load is not indicative of an error in
~ the analysis. This concern is resolved.
't
. p. Welded Stepped Connections, Fillet Welds and Skewed Welds
. Mr. Doyle raised concerns about welded stepped connections, under-sized fillet welds and the use of skewed T-joint welds (Doyle Deposi-tion, Tr. 3742-3749). Mr. Doyle referenced the requirements of the c.
American Institute of Steel Construction (AISC) and the American Welding Society (AWS). Doyle Deposition Attacnments 68 throu'gn 6F provided specific references to requirements for circular tubular
, joints, minimum fillet weld sizes, and multiplying factors for skewed T-joint fillet welds. Deposition Attachment 6A identified supoorts .a Nos. CC-1-045-026-A33R, SI-1-031-704-A32R and MS-1-029-039-563R as
- examples where welds are undersized by 1/16 inch. The Attachment Q also identified support No. AF-1-008-003-533R as being in violati'on
,] of AWS Code require.ments for the diameter ratio of circular branen g and main members.
y - Welded Stecoed Conr. actions
$ Welded stepped connections are' perpendicular joints between pipes or
- $ tubes of different sizes. With respect to support No. AF-1-008-003-j S33R, the referenced AWS requirements for stepped pipe-to pipe geome-
- ) tric parameters do not apply since there are no pipe-to pipe welds at this support. Nevertheless, the Special Inspection Team reviewed the 3
PSE design guidelines being utilized for the design of integral - stanchions on pipes. Mr. Doyle's AWS reference aoplies to the design of architectural tubular structures which are not intanded to serve as pressure piping. The design of integral attachments on pressure j piping is governed by the ASME Code, not the AWS or AISC code.
; Design guidelines being implemented for pipe to pipe attachments tallow fillet welds to be used when the ratio of the diametar of the stanchion over the diameter of the pressure boundary pipe is less than or equal to 1/3. A combination bevel' and fillet partial pene- .
~
tration weld is specified when the ratio is greater than 1/3 but less than or equal to 2/3. Ratios greater than 2/3 are treated as soecial cases requiring analysis of actual effective weld throats. Addi-tionally, local effects due to integral attacnments are analyzed during Gibbs & Hill's pipe stress analysis to verify that localized pipe wall stresses do not exceed ASME Code allowables. This analysis is performed utilizing the CYLN0Z 2 computer code. Representative examples were reviewed by the Special Inspection Team during the
. inspection at Gibbs & Hill. The Sp~ecial Inspection Team found that the Gibbs & Hill stress analysis techniques are acceptable (See
. Paragranh 3.1).
a Althougn Mr. Doyle's exnibit (Ooyle Decosition Attachment SS) only
;, related to circular tubular joints, the Special Inspection Team considered tne adequacy of the design of perpendicular tube-to-tuce welded connections by reviewing the results of analytical evaluations l m
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of such connections. The following paper, published in the American j Society of Civil Engineer's Journal of the Structural Division, was reviewed as part of this inspection:
. _ _ _ . . Finite Element Analysis of RHS CRectangular Hollow Section]
T-Joints," by R. A. Koral and F. A. Mirza, September 1982. U This paper presents the results of RHS T-Joint modeling performed to determine ultimate and working strengths and to determine the sensi-tivity of joints to different geometric parameters. Punching shear
; and rotational stiffness results were also analyzed in this reference. .
*. The model utilized in this reference takes into account strain
?, hardening and the rounded corners inherent in tuce steels. Korol and Mirza found with respect to ultimate strength, that joints having member width ratios of about 0.4 and less are weak in resisting j branch moments and punching shear without reinforcing. Supports at Comanche Peak do not fall into this category since the lowest width
'b] ratio utilized is 0.67. For joints with member width ratios greater
-3 .
than 0.6, Koral and Mirza found that RHS T-joint connections are much
- stronger. The authors also found that ultimate axial loads and moments were typically five times higher than the corresponding yield
, load and moment. Since the designs for RHS T-Joint designs reviewed i
at CPSES are similar to designs shown by Koral and Mirza to be in conformance with sound engineering practice, the Special Inspection 2 Teai found that the tube-to-tube joint designs utilized by the Applicant represent connections wnich will perform the design intent, and their use is acceptable. Fillet Welds The drawings of the three Jupports with alleged undersized welds were evaluated by the Special Inspection Team to determine if an under-sized weld condition was specified. The Special Inspection Team
, determined that support No. CC-1-045-026-A33R is a support numoer which has never been issued and there is no other indication that
, such a support ever existed. Design drawings of the two other sup- !
ports (suoport Nos. SI-1-031-704-A32R and MS-1-029-039-S63R) were evaluated with respect to minimum fillet weld requirements of Appen-dix XVII of the ASME Code. All the reviewed welds were found to be in accordance with Code requirements. Prior to the concerns of Messrs. Walsh and Doyle, representatives of the NRC Region IV Engineering Section and the vendor Programs Branch conducted an inspection at Nuclear Power Services, Inc., (NPSI) on i November 17-20, 1981, (Inspection No. 99900531/81-01). During that inspection,15 support drawings for the CPSES which specified fillet welds thht were not in accordance with ASME Appendix XVII require-ments ,we're identified by the NRC. As a result of this findings, NPSI performed an internal design audit to define the extant of noncon-forming fillet welds. The internal review identified 382 supports which did not meet the requirements of the ASME Code for minimum
-s .
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.N, 51 J ~2 i fillet weld size. Component Modification Cards (CMC's) were sub-1 sequently issued to modify all welds not meeting code requirements.
j As part of the current inspection, an independent review was per-j formed by the Special Inspection Team of a representative sample of the documentation which defines and resolves the undersized welds j identified during the inspection at NPSI. This review indicates that g , the affected supports are now in compliance with the ASME Code. In addition, during the Special Inspection Team's independent design
$ review of the one-hundred supports referenced in Paragraph 4 of this ,j report, all specified welds were evaluated for adequate size 'and no 'i discrepancies were identified. '1 Skewed Welds i
f.! Procedures utilized by the three pipe support design groups for the
; design of fillet welds at skewed joints (skewed welds) were revieked q by the Special Inspection Team during.this inspection. Skewed welds H are those welds joining two structural members that are other than in j
the same plane and are not perpendicular to each other. A typical example is two members joined at an angle of 45 degrees with a weld at the joint toe of 135 degrees and another at the heel of 45 degrees. j Weld angles between 60 and 135 degrees are being analyzed by the
, Appif cant by determining the effective throat of the weld. (The j- effective throat of a weld is the minimum distance from the root of a 1
, weld to its face). Weld angles of less than 60 degrees but greater l y' than 45 degrees are considered to be groove welds, not fillet welds, and are thus considered to be a form of penetration weld. Groove-F welds are allowed by the ASME Code. This weld type is also allowed I by'the AWS Code, which Mr. Doyle erroneously identified as the e controlling code (Ooyle Desposition Attachment 6). The Special
$. Inspection Team concluded that the design procedures being utilized 7 by the three pipe support design groups for skewed joints are based
. 7, on sound engineering practice.
9 g An additional related matter not raised by Mr. Doyle was earlier :
'i identified with respect to the adequacy of the Applicant's quality
- control inspection criteria for skewed welds (NRC Inspection Report No. 50-445/82-14, Unresolved Item No. 8214-02). In response to this item, the Applicant has begun a reinspection program of skewed welds in supports utilizing newly developed inspection criteria. This pro-gram, when completed, will provide information on whether skewed welds have been constructed to the required sizes. This item is still under review by the NRC Region IV staff.
, Summarv Mr. Ooyl.e's concerns about welded steeped connections in circular tubular joints, undersized fillet welds, and skewed T-joint welds have'not been substantiated. The concern is resolved.
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}1 .q. Section Procerty Values Utilized by PSE le p Mr. Walsh raised a concern about the use of two different member -+
properties for tube steel sections (Walsh Testimony, p. 5, CASE
.: Exhibit 659). Mr. Walsh stated that because of the variation in the 2, properties, " reactions and deflections could be off by a much as
; 25 percent" (Walsh Testimony, p. 5).
d i Prior to January 1982, Pipe Support Engineering (pSE) used the member
,7 property values of the 7th Edition of the American Institute of Steel ;j Construction (AISC). From January 1981 to January 1982, PSE also .
4 used the values listed in the Welded Steel Tube Institute's Manual of Cold Form Welded Structural Steel Tubing (1974). Subsequent to January 1982, PSE used the values listed in the 8th Edition of the AISC Manual. Calculations of stiffness and stress were performed by
,; the Special Inspection Team on a cantilever beam to assess the true c generic impact of the section variations. The maximum relative ;f difference in stiffness in a 6 X 6 section of tube steel due to' the i -
use of differing tube steel member properties was found to be 4.6 f[; . percent. For a 4 X 4 section the relative difference was 7.5 percent. The relative differences in stress were found to be 4.2 percent for the 6 X 6 tube steel section, and 7.1 percent for the 4 X 4 size. These values are based on the maximum difference between the 8th N. Edition of the AISC Manual, and the two previously used member 9 property tables.
'?
. Since all large bore and Class 1 small bore pipe support designs are g being re-examined by the Applicant using the member property values
;. in the 8th Edition of the AISC manual, only small bore Class 2 and 3 u supports are affected by the variations. As discussed above, the
, actual variations in stiffness and stresses are all less than 8 percent.
? Accordingly, the Special Inspection Team found that any impact on tne 9
affected supports resulting from the section property variations are
't minor and will not result in unanticipated behavior of the supcort .
y due to gross errors in stiffness. Since the actual variations in the ;, q stress levels will not exceed 8 percent, it is not expected that
- 7, unforereen detrimental stresses will occur.
The Special Inspection Team concluded that Mr. Walsh's concern about different tube steel section property values utilized by the PSE pipe succort design group is resolved. The Applicant is currently reanaly-l zing all large core and Class 1 small bore pipe suoport designs using consistent memoer property values. The differences in section property values for small bore Class 2 and 3 supports are less than 8
, percent, and will not result in unanticipated suoport behavior. This
, concern ,is resolved.-
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i i j r. Sucoort Pads Welded Over Pice Girth Welds i j Mr. Doyle expressed a concern about the welding of support pads over j pipe girth welds (Doyle Deposition, p. 82). Two pipe supports are 1 specifically identified, Nos. CT-1-137-701 and CT-1-137-702. Mr. 4 Doyle stated that both supports are ASME Code Class 2 supports for j ] the Component Cooling System. He further stated that ASME Code 4 Section XI will be violated in that inservice inspection will become j i impossible to perform on the covered girth welds. Page 5 of ,0epost-
) tion Attachment No. 2 is sketch of a detail on the suoports wnich are i identified as support Nos. CT-1-137-701-522R and CT-1-137-702-522R.
y' The support numbers designate these two supports as Class 2 supports. The sketch appears to be a rough representation of a proposed support
, orientation and does not contain a drawing number.
S Upon review of the controlled Brown & Root support drawings and
.i associated Component Modification Cards, the Special Inspection Team 1 '
determined that the supports identified by Mr. Doyle are actually Class 5 supports for the containment Spray System. The correct 1' support numbers are CT-1-137-701-525R and CT-1-137-702-525R. The Special Inspection Team concluded that these supports are correctly i .. designated Class 5, because they support that portion of the Contain-ment Spray System.which forms the by pass test loop and are not . f relcted to the functional safety-related portion, which is Class 2. q The Special Inspection Team notes that Class 5 pipe welds are not . j . included in the ASME Section XI inspection program. f*
. Additionally, the Special Inspection Team determined that both 4
1 supports employ removable clamps and do not have any parts perma-J nently covering a pipe girth weld. The existing designs for these j, two supports do not preclude the inspection of the girth welds. 3 , f Mr. Doyle's concern that pipe support pads on Class 2 pipe supports ,. j were welded over pipe girth welds is not correct. The supports '
;d identified by Mr. Ooyle are Class 5, not Class 2 supports and there- ;
fore. are not included in the ASME Section XI Inspectiert program. Second, the supports employ removable clamos which do not preclude inspection of the girth welds. This concern is resolved.
- s. Damace to Pioe Suecort Ourine Hydrostatic Testine i
Mr. Ooyle alleged that pipe support No. CC-1-115-028-F43R (Ooyle , Oeposition Exhibit 11W-11XX) failed during hydrostatic pressure l testing of the component cooling water system by excessive local yielding of the support tuce steel wall (Ocyle Deposition, pp. 72-73,
. 181-182).
l 6 i The Applicant stated that tne deformation due to local yielding in l this support: (1) had occurred during installation and adjustment of l the piping system, and (2) had been identified by a design review ( engineer during- a field walk-down prior to the hydrostatic pressure i . . s.. . g. m . . ...
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i - k 54 4 1 1 j testing'; i.e. , long before Mr. Doyle made his allegation. The
- Applicant further stated that the support is being modified to 4 accommodate increased piping loads resulting from an as-built piping i analysis (Applicant's Supplemental Testimony, p. 6, Applicant Exhibit ^
l 142F). 3
$ The Special Inspection Team reviewed the Applicant's QA Nonconform-i; ance Report (NCR) No. M-2531 against support No. CC-1-116-038-F43R, j dated October 9,1980, which indicated that the adjustable eye red of j '
the support sway strut was bent approximately 15 degree in the 1 threaded portion. Sending was attributed to " movement of the pipe - d without disconnecting the sway strut" (NCR No, M-2531). Although NCR , M M-2531 makes no mention of damage to the 1/4 inch x8 inch x8 inch d structural tubing to which the 1 inch x 2 inch clevis of the support
?. sfvay strut is welded, it is reasonable to expect that the structural g tubing was damaged at the time of issuance of the NCR in view of the it severity of damage to the eye rod. The NCR disposition required only d that the bent eye rod be scrapped. Replacement of the bent eye rod 4 was documented in the NCR follow-up inspection report dated November 25,
( 1980. The Applican' t 's hydrostatic pressure test Data Sheet No.
.: ICC-014-1101 and Flow Diagram No. MI-0230, R-6 show that the hydro-N static pressure test was started on May 13, 1982, and completed on 4 May 14, 1982. Since no other NCR's were issued on this support between
'j November 26, 1980, and May 13, 1982, it can be assumed that the damaged.
.i tube steel was still in place during hydrostatic pressure testing.
M Additionally, the results of 'the Gibbs & Hill analysis for stress
'i problem 1-64F show that the 158 lb. load incurred during hydrostatic j, testing was much smaller than the 4897 lb design load for normal l and upset operating conditions. Since the hydrostatic load is only a 9 small percentage of the design load, the Special Inspection Team concluded that the support could not have been damaged during hydro-
]; static testing. In subsequent discussions, the Applicant showec j ,
the Special Inspection Team a copy of Component Modification Card
,) (CMC) 81948, Revision 3, dated October 28, 1982. This CMC showed that the original tube steel alleged to be damaged was to be replaced ;
d)= with a new piece of 3/8 inch x 6 inch x 8 inch structural tubing.
- j The NRC Staff verified that the damaged tube steel was replaced.
~
Mr. Ooyle's allegation regarding the cause of the damage to suoport No. CC-1-116-038-F43R was incorrect. The support was damaged before
+
the hydrostatic test. The damaged tube steel was replaced. The concern is resolved.
- 4. Inspection of Vendor Certified Suecorts b In its tastimony regarding the concerns raised by Messrs. Walsh and Coyle, the Applicant, stated that Messrs. Walsh and Doyle did not see final approved designs because of their position within the organi:stion, and tnat the examples provided by Mr. Doyle were not final approved (vender-certified) designs, but were interim designs in the pipe support design evolution process. .The Special Inspection Team reviewed the design status d
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L-1 t i 55 . G i 3 of each of the pipe supports identified by Messrs. Walsh and Doyle and 3 found that only one had been vendor-certified (Support No. CC-1-107-008-
't E23R).
l$ g The Applicant also presented testimony that the final design review i procedures and practices would have detected and eliminated any of the 5 significant shortcomings identified by Messrs. Walsh and Doyle, had all 3 steps in the design evaluation of the pipe supports taken place. In order 1 to determine the validity of the Applicant's statements, the Special
$ Inspection Team evaluated the implementation of the design review
- process j by reviewing a sample of the pipe support designs which had completed the design evaluation process; f.e., the pipe support design drawing had been
;q marked " vendor-certified."
'b>
3 Theref5re, in order to assess the overall design process, the Special *
@ Inspection Team performed an inspection of 100 pipe supports which O have been vendor certified. (Such certification indicates that the pipe M support design organization has reviewed the support and certified that it 9 is capable of resisting the design loads as determined from the Gibbs &
~y Hill pipe stress analysis.) As of the data (December 1 and 2, 1982) of 1 this inspection, 1,264 supports have been vendor-certifitd. Of this total number, 1,161 (92 percent) were ITT-Grinnell designs and 103 (8 percent)
]hj watre NFSI designs. No PSE designs had been vendor certified as of the date of this inspection. Utilizing Military Standard 1050-63, " Sampling 5 Procedures and Tables for Inspection By Attributes," as a basis for i determining an appropriate sample size, 80 ITT-Grinnell supports and 20
'.i NPSI supports were randomly selected from the 1,264 vendor-certifled *
- f. supports for independent design verification. Each support was reviewed
% for adequacy of design with respect to the following attributas: stability; i compliance with the Gibbs & Hill 1/32-inch temperature / clearance recommen-N dation for box frames; fillet and angular weld sizes; floor-to-ceiling and U well-to-wall constraints; Richmond anchor offset frem centerline; inversion -Q of design load condition magnitudes; potential for unaccounted torsional D loads; local stress effects; cinching of U-bolts on non-rigid supports; $ use of U-bolts as two-way restraints; out-of-p1'ane bracing for seismic :
1 loads; static and dynamic effect of support mass on pipe; kick load poten-N tial; out-of-plane friction loads; and satisfaction of the Apolicant's 1/16 inch deflection guideline. The vendor certified supports which were ~ selected by the Special Inspection Team for independent design verifica-tion are listed below: ITT-Grinnell Succorts Sucoort No. Revision succort No. Revision i i -i - AF-1-001-032 Y33R 3 AF-1-003-004-533R S
/
AF-1-009-019-533K 3 AF-1-027-005-533K - ! ) 9{
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' $ AF-1-100-019-533R 5 BR-1-016-003-553R. 2
-ir j :
BR-X-106-056-G43R 2 CC-1-007-021-A43R 3 U CC-1-009-003-A33R 6 CC-1-011-018-A43K- 4
,I (1 CC-1-030-007-533K 3 CC-1-033-003-533R 6 -
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CC-1-110-004-A43R 4 CC-1-125-012-F33R 4
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.4 a;> CC-1-155-004-543R 2 CC-1-156-015-A73R 3
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y CC-1-167-004-543R 2 CC-1-196-002-552R 2
," CC-1-203-003-552R 2 CC-1-250-002-SS2R 4 CC-1-250-005-552R 1 CC-2-011-004-A73R 4 CC-X-006-003-A43R 3 CC-X-021-002-A43R 3 l CC-X-031-002-A43R 4 CC-X-038-006-F43R 2
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I CC-X-041-006-F43R 5 CS-1-014-016-552R 5 5 C5-1-063-013-522R 4 g-CS-1-074-042-542R 2
..... . . CS-1-155-025-542R 4 C5-1-158-008-542R 4 .
CS-1-158-040-542R 3 CS-1-217-002-A42K 2 .
?
. . CS-1-454-010-552R 3 CS-1-911-007-552K 4
, 00-1-012-018-Y33 R 5 00-1-012-043-Y33R 2 FW-1-103-002-562R 2 FW-1-103-005-562R 2 FW-1-104-011-562R 3 EW-1-114-004-562R 2 MS-1-025-007-572K 2 MS-1-027-019-543R S MS-1-027-039-333R 5 MS-1-028-006-563R 5
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MS-1-028-046-553R 2 MS-1-073-002-552R 4 o
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MS-1-075-003-552X 5 MS-1-416-004-533R 2 RH-1-013-008-532X 3 RH-1-014-010-532X 5 5 3 RH-1-025-004-522R 2 RH-1-064-008-522X 3 1 d-: SI-1-031-020-Y32R 4 .SI-1-031-038-Y32R 3
} SI-1-031-064-532R 3 SI-1-038-009-522K 4 i; SI-1-039-027-532R 4 SI-1-039-036-542R 3 .y .' , SI-1-052-003-542R 2 SI-1-093-006-532R 4
- s d SW-1-007-010-J03R 3 SW-1-027-003-J03R 4 5
f.j SW-1-102-080-543R 2 SW-1-129-067-543R 2 r
-l ,W-1-132-042-A43R 2 SW-1-132-068-543R 4 .] -
SW-1-173-049-543X 3 SW-2-004-008-A33R 2. il - SW-2-102-006-A33R 5 SW-2-132-013-A43R 3 . 3 . M SW-X-007-001-J03R 3 SW-X-007-002-J03R . 4.
~
- . NPSI Succorts 5 - :'i . - (j -' - Sucoort No. Revision Succort No. R'a vision
.1 - $$ h b -
CC-1-197-013-C42R 3 CC-1-197-038-C42R 2 CC-1-199-002-552R 4 CC-1-208-002-C53R 3 CC-1-208-006-C53R 3 CC-1-208-008-C53R 2 CC-1-212-007-C53S 2 CC-1-215-032-C53R 2 CC-1-256-006-S535 3 CC-1-254-003-C53R 2* p CC-1-272-006[C53R 1 CS-1-012-003-C42R 2
./
CS-1-012-005-C22X 3 CS-1-077-013-Ca2R 4 b
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) CT-1-039-431-C42R 2 CT-1-051-416-C72R 2 3 ) FW-1-098-008-CS25 2 FW-1-099-006-C62K 3 ?.
e i, MS-1-151-005-C52R 2 MS-1-151-012-C52R 3 4 v
^3 'j -
As a result of this independent design review, the Special Inspection Team has determined that all supports reviewed in the random sample satisfy tne
'l Applicant's applicable d'esign criteria for the attributes reviewed. The * ) review did not disclose any discrepancies in the random sample which would 2
indicate a failure of the Applicant's design verification program to
.? identify and correct supports to assure compliance with applicable design
- l. criteria. .
Y
- a. 5. Exit Interview -
,v
.Ej - An exit interview was conducted on February 8,1983, with the Applicant onsite. Each of the unresolved and open items were discussed. The following NRC personnel were present: 5. J. T. Collins
. G. L. Mcdsen
.. S. 8. Burwell
- 7. F. Westerman
; 0. L. Kelley
- a S. McCrory 1.[
I ' ll 1 4 . 5 F
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NUCLEAR REGULATORY COMMISSION A
~7, ~ $ wasmcTow. o. c. 2 ossa
- [ [#
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$1 d -
kf MEMORANDUM FOR: Harold R. Denton, Director, NRR
-/h
. Richard C. DeYoung. Director, IE -
- Robert B. Minogue, Director, RES
{ Thomas E. Murley, Regional Administrator, R.I James P. O'Reilly, Regional Administrator R-II N f'f James G. Xeppler, Regional Administrator, R.III 4 > -John T. Collins, Regional Administrator, R-IV _y_ . Learned W. Barry, Director &: Controller, RM William J. Dircks FROM, Executive Director for Operations
SUBJECT:
3MPLEMENTATION OF THE COMANCHE PEAK PLAN FOR THE
' COMPLETION OF OUTSTANDING REGULATORY ACTIONS i 1
l Recently, Tom Ippolito, Project Director for Comanche Peak, circulated for ! review and- discussed with a number of us, the Comanche Peak Plan for the ; completion of Outstanding Regulatory Actions (Enclosure 1). This plan has. : been approved and I am now requesti.ng actions by each of you, as specified below, to implement the plan. These actions are consistent with sqy discussions wil.h Regional Admini'strators during the mar.agement. meeting on May 17-1.8. 1984, and my memorandum of May 25, 1984 Th2 Plan describes regulator
-- : inspection, and allegations)yand actions in four identifies themajor areas resource (licens'ing, (including requirements hearing.
3 funding) needed to complete these regulatory actions. All activities are ; geared toward completion by October 1,1984, the Applicant's projected fuel load date. , 1 Most of the resources needed are to form a Technical Review Team (TRT) to evaluate and resolve a large nuder of. technical issues, including allegations. l presently identified. The TRT staffing and technical groups breakdown is ) shown in Enclosure 2. Approximately two-thirds of the TRT reviewers will be drawn from contracted resources; however, NRC staff resources will be used to supervise each technical group and to integrate their activities with ongoing Region IV act<1vities'. The resources will be distributed between the various offices to minimize the impact on any single organization. Where supervisory personnel are required for the lead of eachr of the TRT technicil groups, I expect you to nominate individuals who can manage effectively in a demanding and visible environment. All personnel should. be fully qual.ified for the requested function and should have a proven capability of balancing a nuder of views to reach a conclusion. In adjusting interna.lly for resource allocation, you should ensure that operating reactors are adequately covered. l .. --- .. c'?g REGl0NAL ADMINISTRATOR
- I pid)[ A DggyTY ADMINISTRATOR gg po ADMINISTRATOR
..n... ... DmcI9R.N
I . v- ; .' .~* , ' ; . . ~ I -- ~~~ [ Resource needs, by Office, are identified below. To the extent possible, re-sources identified should include individuals already familiar with the Comanche
. - . _ . . Peak profect or those familiar with the regulatory actions completion program at Waterford.
- 1. NRR 2 project. Managers to coordinate additional licensitiv octions
~
Tedhnical, Reviewer to coordinate O! activities with the TRT Program
. Manage
- 2. E
' Assistant'(Deputy) to the Program Manager of the TRT (Jim Gagliardo has already been designated to this function for both Waterford and
. Comanche Peak).
- 3. _IE,'NRR, RES and Regions I .II, and III Gmup Leader to manage the Electrical / Instrumentation group of the TRT.
. Group, Leader.to manage a joint Civil Engineering and Mechanical Ingineering group of the TRT.
- Group: Leader.to manage the Test Program Group of the~ TRT.
- Group' Leader to manage the QA/QC Records Group of the TRT. The
' assignment of an individual who previously participated in the f" Comanche Peak special review held in early April 1984, would be desirable.
Group Leader to manage the Coatings Group of the TRT.
- 4. Region IV On 'an;as needed basis and upon request of the Project Director, Region IV'will provide 2-3 inspectors to support the TRT effort.
._-..-. - .5, g.
Financial resoure.es (as previously identified in separate correspondence from H. Denton) to support contracted activities. These needs include:
- approximately 2/3 of the staff of the TRT inspection personnel to. augment Region IV inspection effort technical , expertise for coatings issues e
f
9 . . 5*r, s .'. . . . {.,.. :
}
- 3-Candidates P are to be identified directly to J. Roe by June 14, 1984 Designated individuals should be available for commencement of their assigned tasks on about June 18, 1984
~*~~~'If there are any questions regarding this memorandum please contact me promptly.
V-w - t b illiam J. Dircks
- y E xecutive Director for Operations
Enclosures:
As Stated - cc: J. 8.'Ma'rtin, RV B.' Hayes. O! P. Norry., ADM G. Cunningham,. ELD
- T. *Ippol,ito, DL/NRR
, D. Eis'anhut. DL/NRR J. Dayis, NMSS
- m. M. M.
6 WEEEEBe ee se .e 6
- V M
I . 8
* " * " '- l 8
.. .4 . . 4 O
e . o O s
;. EMCLOSURE 1 W .- . ..
. '. y
. \
COMANCHE PEAX Pt.AN FOR
'. . THE COMPLETION OF Oui. STANDING REGULATORY'ACTIO'NS l
; MAY 1984
.;- Aeproval:
~
. E.DA-s T t % = t. W m '
. c. 'c 'ss R. DeYoung'. Dj, rector. Ii nate
'1 W, .n'bL'- l N .5'L '
. H.' T. Denton , Di rect.or, NRR N te '
i R. L. . . .e L\ Tit s ,- % w. T,%9, ; *, yc
. . J. T.' Col Mns , Mministra *.or Date Region IV 4 .
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- I ! COMANCM M/.Y P.f. rCP THE COMPLET!0ft OF OCST*.*:r'I? / o.E90LATORY ACTI0t!S t -- - --
- t. 'pv9905E MIC Scope On t' arch 12,19Pa, the EDO d' rec M "R4 in r anace al' rec =ssarv t'er actiert
'leadinn 'tc licensinn dec'sicas 'ae 0.'ranche Peak and lle*arford. A copy c#
that direc*4ve is included as U We'ont 1. This plan as'e14shes +.be -i progran for Comanche Peak. The purpose of this plan is to assi.r* the overall coordire*4on and integration of the outstanding eacuir*.ory actions regarddac Cemanene Peak, and, achievino their resolutinr cr er to a licensing decision. This plan i eacomoasses all licensing, ir.sacc**on, bearino, and allecation issues.
~~ '" " Further,' this clan addresses tha cecce of the work readed, srecifies the critical' path issues, identifies ine responsible line ornani ation. to setieddle for completien, and (wrece applicable) the need for addi*incel resources 'c meet the schedule. .
l'he planned comoletion date for all reculatory actions is usumed *n be October 1,1084, and resnurce needs are predicated on *. hat assumotion. A' status.rannrt will be issued to management every two week 5 s ar+.iro two weekstafter the approval of the clan.
!!. BACKGROUND -
Comanche Peak Steam Electric Station Unit 1 is in the final staces of the
. cperating license review process. The Construction Permits for Unit 1 ~
. and 2 were issued on December 10, 1974 Texas Utilities docketee their aaplicatinn f.or operatino licenses on April 25, 1978. The rire' Envirnemaats' Statement was issued September PA, 1981. The Safety Evatun t.iun ocer'. (SEO was! issued en July 14, 1981. Because of'the large number e# outsfencing issues ideati.*ied in the SER, the s*a " reconsnanded r deiryino the ACRS rev aw.
d S E.8c Supolement No. I was issued on Octcher 16, 1981, and the ACSS meet. ire _ . _ _ _ _ _ . was; held on Mcvember 13, 1981. The ACRS, by letter dated ?!cvecoer C, '"F'.. sucoorted issusnce o' an ooerating 1-icense. The lates* SE9 suco1**tr* was issued on Novenber 23, 1983.
'lomanche Peak has been in a heavil." certested hearino fnr ove- *wo .vaws.
All' but one centention have been dismissed. The rerainino centention cuesti.ons the ability of the apolicant's Ouality Assurance /nuality Control Drogram to prevent deficiencies in t'he design and construc*ien n d *he rier*,
~ ~ - " -
- The..i.itensing 90 erd has admitted many allegations of desion and crestructior def1ciencies into the hearing as relevant to this contentlan.
. Th'e Apolicants ara current 1v orn.iecting a fuel Inad dr*e #ne Unit 1 to be in late Septenher 10P4 The basis for this prefection was crevided te the staf* on tdav 7, 198a. This fuel load date acomers achiavable but allows no fle.xibility for unexpected avents in a Very t'oht schadule.
The nurter of, hearing issues and urcertainty regardinc the +1minn of 'he
..... . . 1.icensind Roard's initial decision may imoact the fue1 load.
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h 2 II, inAN F00 WE C0fiPLET!CF "e SiiTST/"Nt r, REGULATORY ACt0NS This plan describes the net $cd ir which coordinated reculatory actions
' 'are to be *:aken by the sta ## te he ready to support an NRC decis ion
,regarding Comanche Peak 'icensiag. As stated in the Purposo, +he n1n encomcasses all licensino hearing, inspection, and allecation issen.
. This surinar.v addresses the score o":wcrk. needed, identi'ics the r'esconsible line neganizatior, tha schedule for completion, anti the
. rescuece needs to meet the schedule.
The management organizational arrangeent responsible for directing the overall e# fort and coordinating actions by the varicus invnived offices is shown graphically in the enclosure to the EDO memorandum c' March '12,19M (Attachment 1). The management is under the overall tirecticn o# T. A. Ippolito, WFo ,eports r to the' Director of the
. Ohlsion cf Licensing. The managers responsible for implementing ~-~ ~ C ~ , and directirg this organization are the following individuals:
Project Director (T. A. Ipoolito)
- -t0 ELD Contact (.1. ScintnH
- - - - - -i OI Contact (B. Hayes 4
. PtRR Action Region !Y Actions TE Actiens (T. Novak) (R. Bangart) (R. DeYoung)
The line offices will centinue to manage their own eesconsibili+.ies
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recard/ng Comanche Peak in accordance with the schedule arc object'ves 6f.'this plan. Line office activities are to be ccordinated with tha program manacement organizetion via their representativo as identi#dec'
. . . _ . _ . above. Additional rescurces are expected to be necessary to sunoor+
licensing, hearing, and inspection issues, and substantial resnucces p Comanche Peak. arc necessary to respond to the approximetely 400 allegations This plan ornposes the fonnation of an Technical Rev4ew Team (TRT)
' to.gvaluate and resolve a nunter of technical issues, including
. :)$ allegations, presently identified. A proposed orcanizational chart of the';TRJ f.s shewn below. The groups identified will be assigned to evaluate and resolve technical issues and allegations that have been
_ _ _ . _ . _ . . Grouped into five technical areas s OA/ CC , El ectri cal / I ns t rume n ta t i on
'Civ.11/Hechanical, Ceatings, and Test Programs. The groups will be comprised of a group leader and reviewers that are specialists in the carticular technical area.
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. Corucche Peak Techrical Review Team ITRTT Prn,4ect DirPCtor n eputy Pro,4ect Director w
- L---- Office of investications GROUP LEADERS rNnC Electrical / Civil / Mechanical Coatings Test Pregrar Records : 'hsteurenta tion s arc
, Coe*a'.i n nai Reaciress
- The stp* finn of these groups will be. drawn frori the var *ous Nec of#8ces andler: contractors as arranged between the Project 111 rector and lir.e
.. .. .. reanecertent. The TRT may be called together for a specified period t of time, dispersed back to the indiv* dual's parent office. and *. hen - -. : reconstituted in whole or in part as needed to corolete resolut on "
d Fof like issues. ,
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*The TRT will b,1 tMer the direct supervision of the Prc.dect Director.
.tr! recordance with the EDO memorandum of tiny ?5,1984, the 'RT -
l organiyationisscheduledtoteinplaceandfunctineingby@ne a. I?Pa.
"iDetailed guidance will be issued by the Pen,4ect Direc cr y the Techedcal
* - ' - .ReviewiTeam and other participants ih this effort. This ruidance wil!
. ' address the *ollowing:
I .. Method and appecach for identification and
. . disposition of allegations -
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'- Tiacking System l -. Preparation of Documenta. tion and Records
. Protection of Individuals i-
* *l.tnitiation of Special NRC actions, such as
.Confinnation of Action Letters or 50.54(fl
. ' letters
- Manpower acccuntina
' The basic uoon which the schedules and resource es**ma'.es have ben : -
' developed is that the Comanche Peak fuel load date is October 1. !984 Siaure'l .is an overall schedule and Figures 2 throuch 5 ar* 4ndividual
'scheduTee for the resolution ad Licens tre, llearing, l'espectiusi. and "A1: legations Regulatory Actions, respectively.
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-- ne r9a,*nr issues, schedules, and rescurc? es+4etes needed to meet
'the. schedules are suffr'arized as #011 wt.
,A., t.ieersino Reculatorv Actiens Licensing Actions are these thirgs resultino fecm the desice review of the FSAR. NPR is responsible for the resciutier
- ~ of these actier items.
. ra The total number of outstancting action items is 27.
..- Fnur of these action items are cons.idered to have the potential
..- 'or impacting the schedule. These ' items relete *.o U the acecuacv of the TDI diesel generators, 2) the 93 Applicants' exemp?irr
. request for relief from-GOC-4, 3) review o' the Cygna Report rf
.i . an indeoendent assessment of desion and constructicn, ard 0
' electrical equipment environmental cualification.
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.'!RR experience with othar facilities involved in complex licensiec - - - - - - - ;' ; reviews (Diablo Canyon, Seabrook, and Shoreham) indica +es that additional project management resources are necessary. Two
- t ' additional pro,iect managers for the' period from .1une-Septerbe" will be needed, fnr a total of 8 man-months of sdditiona? e## ort.
-i The technical resources present.ly assioned by NRR to evaluate and
. ' , . resolve the remaining open licensing actions are sufficient to . leet
- -- - . . ;i the schedule shown in Fioure 2. Additional IE resources ar* art expected to be recuired as the GT inspection is complete anni
.: CA/0C reviews and emergency preparedness reviews are essentia"y Z. , , . .
complete. f.,WaringReculatorvActions
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. [ , . llearing Actions are those issues in, content 4nn before the ASLP.
l 7ere are three major issues each with a number c# sub. issues. 'he
- - . - - .- ehree major issues are Desion Adeovacy and Oua11ty Assurence.
. Construction Adecuacy, and Construction ruality Assursece.
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..There are two critical oath actions; Desien Affenu no Constmetinn
._._. ) Adecuacy. The desion adecuacy action concerns an TOVp being oerforced by the Applicants ai: the staff's request. CYf1NA is per#nmine the review for the Aeolicants. This is current?y uncer review by the sta#f. Cyona personnel actions may have contributed *o be creantiff-
,' catinn of insoection areas to App 11 cart 0A/0C personnel. Rasolution
- - * - ~ * -
- . nf this concern na" make it necessary to recuest additiceal l ; independent assessment activiti'es.
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.; The critical path issue concern'ing Qonstruction Adeouacy is containment liner coating (painting).
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-- -- -- ' .: . The resources oresertly availah'e n' e su'# icier * *e resolve sit hearinc actices with the exceptina o' the critical ::ath issues. It is es*1 rated that 10 man-mor* s are rec.uired to resolve the i Desien Adeouacy Actien, and 6 rwo.nths to evalua+a *Me Corst-ucen-i Adecuacy Action (paintine). The design adeouacy review wil'. recuire
,' a tean composed of IE and NRR per:;onnel, similer to the Cygra !"VP cffor*. . .-
t -
- Coordination of Fearing ac'fvities is expected ta *e extensive and involve intenrating *.Me activitiet of MRR, OELO and Decinn ?
.- with tra Technical Review Taam. An additional senier canace" 'SES-
- 1 leveli is needed to man' age this effort as it is a.xpected ' hat the
': Pec.4ect Director will deierte full time effort to carace-ent of
! the technical review ted.n 4ctions 9,onenencinoG_une a, 19P C) il These estimates assume that the reviews will cceclude that the
- j'. existing circumstances are acceptable to the staff ane/cc nn mader ccreective Actions are recuired of the Apolicants. Should *his.':reva otherwise, additional resources will be recuired for resolutden.
- r. , See l'igure 3 for Hearing Testimony Completion Schedule.
_____ i. i: C . Insoections Pecuistorv Actions'
.r I: Inspection actions are those that issure that adecuate cercletion
; ;. .of plant constructier. and the readiness of the Applicants te noerste
.the olant. These actions are the responsibility of Region IV.
.a The total number of outstanding act, ion items is 377. These may i be grouped as follows:
i- - SER verification: 30 actions
~4nutine construction inspections, precoefat d neal test t- program and operationel read.iness inspectiert and startup test program: 121 actions
-- - - doerating Licensing: ?O actions
..#. 0cen items inspections (unresolved items, violatices, 50.55(e) items, inspectar follcw-up items and
-',' Part 21 items: ? actfons . . y
!. - Ronm inspections: TBD de 4 oWeow d .
CAT follow-up: 5 acticns _ _ . . . _ _ . / __ . ' All the inspection items requiro resolution orinr to OL issuance.
- ]. . , ; Pany recuire appitcant actions prior tn inspaceden ne relate to
. . g' . hearing dssues. Particularl.v sionificant is t%, retest inspection
' effort as the applicant plans to re-run approximately 25 preoperational
- 8. - -
- tests to confinn system readiness subseouent tn various modificatices
. '. and design changes. Many of these tests will be witressed by
. the NRC and test results wil.1 be evaluated as apornoriate. Svstaa's
.. involved include sefecuards systems, reactor protective system.
'- service water, component cooling water, and the diesel genera *.nr.
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,, The number could impactoffuel inspection load. items represents a sizeable effert that
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Scre edditional resources will .be recuired to crrplete 'ha reu*ine
' inspection pergran and resolve :the r'any coen iter"S. 't is excee**d
' p that *.his crea coulc! reiluire approximately 46 man-conths. Consdrerf aq the number of items and based on Peterford experience, 'h* Deqine
.,, estinates that much of this 7tffort can be handled with ex's-4-r rescurces but that aoproxir.ately 18 man-Ponths erditional res urces will be recuired. See Figure a for the Inspections Schedule.
1
. D..,.Alleoations peculator _v Actions
.._ _. '.i The Allecation ' Actions are thoie concerns reported by Wrieus i' ' . .ir.dividuals, intervenors and. action groups recerding the safety
'1., of construction of the plant. . Concerns regarrding wrongdoings, intimidatien, etc. are not included in the techetcal review *.eam L
;i effort but are referred to OI or nIA ,as appropriate. -
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' ~To date the number of individua1 actions 1s accroxinately 400.
! These actions are grouped into spec'ific categories to facildes*e their resolution. Resolution of these actiers will involva the
.ij Technical Review Team, NRR, OI, and;Degion IV.
.:' '. .The organizational group with primary resconsibility for resolut4er
' of *.hese actions is the Technical Review Team (TRT). The reseurces
, required to resolve these actions are identified balew accercing tc
! the Team functional grouo: .
l . Rescurce a ! Functional Grouc No. of Allecations Es t.faaae l'i'an-r ca*55 '
' ,! . CA/0C R*coros 7 1205 Q u)o&O Electrical /instrum. 4 Civil /Fechanical 07 17
. Coatings .
11 4
. Test Programs
- 14 ?
I Estimated Totals 757 V l The TRT effort is expected to recuire additional administrat ve i
+ - - - -
support (secretarial) of aporoximately 3 man-conths. Perce.
. - . - . _ .. ;. the tctal TRT resource needs are 45 man-months.
.; The to*al program for resolvino.the allegations .1ctinns is a
- .. .. !. critical patti item. See Figure 5 for the schecule for conoletten
;,3 of the review of these al1egations.
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7 e - - .! In addition, 97 allegations will be hardled by the #ellowinc d ' ices:
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Responsibt e
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Furetienal Groue No. of Allecations Head O"'ea Intinidation 30 CI NRR
!~t Design Pipe / Pipe Supports 19 18 NRR/IE Vendor / Generic
<> Independent Assessment Program 7 NRR
__._ _ R:V Miscellaneous 23
- - - - ~ ~ 2' . Design of pioe and pipe supports, and the Independent Assessmart Program allegations will be dispositioned by NRR personnel that
. . are handling these issues for the hearings. Irtimidation allegations will require additional OI resources, as discussed
. later in this section. Existing resources in the Venoor Inspection Branch, IE ard NRR will disposition the vendor / generic allegations. Existing resources in Region IV will be resconsible for tne miscellaneous allegations.
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E. Office of Investigation Actions.
. O! actions are those actions necessary to succort the resolution
.i- of allegations. They involve issues where wrong-doing, intimidatf on.
~ . - - . - - -
or harrassment may be involved. m_. ;
.; ' It is clear that with t!1e present resources assigned to the Cemanche Peak investigation (one; investigator) the senedule for resolving the allegations and wrongdoing issues will not be met. We estimate p several fremadditional investigaters June through September, will .for be required a total on ful? time of 12 man-months of bas's effort.
.. .. .. q i During this 4-month period OI will recuire the full-time succort of one individual with a technical background, as many aIIetetions are
]
a ccmbination of technical and wrong-doing issues. fnr a total of 4
. ;: , man-months.
T'h e kPC staff effort to complete the actinns tn the licensing, hearings,
----' inspections' and allegations areas will be, substantial and the impe_c3._ 4 71 ha
~ ~ ~ ' " - " feltiby several Offices. The f ni ng w=e e lists a total of(B21 spearate) ~
actions requiring approximatel 00 man-month of effort above the existing (budgeted) resources. Personne rm A n:o this effor+ "411 he ebtained fremicontractors. It is estimated. that approximately dmi1TiecM'l be nece$sary to furd contractor assistance in support of Comanche Peak reviews during 'the remainder of FY 1984 The estimates are somewhat fragile and assure thatino major new issues are raised, that-the :Apolicants meet their projected scht4ule, and that sta'f review of' the identified issues will conclude that the existing circumstances, or the resolution, is ' acceptable. 1 I
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- Attachrest
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UNITED STATES
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i"~ ~f * + NUCLEAM REGULATORY COMMISSION
. - -3, j = =, eton.a.c. asses d "4. ./ Muut n im4 "PO ORAN00M FOR: John T. Collins, Regional Administrator Region IV
- Harold R. Denton; Director .; 'Tt Office of Nuclear. Reactor Regulation
'. Richard C. DeYoung, Director Office of Inspection 1. Enforcement
! Fit 0M: William J. Dircks .
Executive Director for Operations _ _ . _ _ . _ . SD5 JECT: CINPLETION OF OUTSTANDING REGULATORY ACTICMS ON CDMANCHE PEAK AND WATERFORD
. Construction of the . Comanche Peak ,and Waterfoni faciTities is nearing completion. There resin a number of issues that need to be resolved before the! staff can make its licensing decisions. The issues remaining for these Staats an ouite complex and span more than one Office. In order to assure
.--. -- the.overall coordination / integration of these . issues and to assure issues are. resolved on a schedule to satisfy hearing and licensing decision needs, I:em' directing MRR to manage all necessary MRC acticas leading to prompt licensing decisions. Darrell Etsenhut. Director. Division of Licensing, NRR is being atsigned the lead responsibility for this activity. He will c4 ordinate the efforts of MRR, IE,. and Region IV, and will coordinate this activity with 0! and M1.D. Prior to any of the affected Officas uncertaking major activities (e.g., inspections) or meking' decisions on these plants.
.that activity should be concurmd 'in by NRR.
iddi are presently in the process of assigning a. dedicated senior manager to assist Mr. .Eisenhut in the management of these activities. The$first phase of. this program will be the identification of issues needed to be resolved for each plant prior to hearing, and licensing decisicas.
-Once the issues have been identified a Program. Plan for resolution of each item should be develooed and isolamented. ' The' Program Plan should address the iscope of the wert needed. the ldentification of the resconsible line
! organization, and the schedule for completion. In principle, this effort I wil:lt therefore be similar to the effort undertaken regardino the allegation riivfew on Diablo Canyon excent thai: this' effort should encompass all licinsing, inspection, hearing, and allegation; issues. , e
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4 2 Each affected of" ice will assign a full time senior manager to work with NRR to deffne , schedule and complete.the issues. , All affected offices shouldI expect these
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identified by each of you within a few days. provide ' dedicated resources and ghe their full support to this effor., to assure that all existing issues an exceditiously handled and all new issues In
--- -- -- are peceptly provided to NRA so as not to delay the licensing decisions.
addition, copies of all infonention, documents, depositions, etc. should be promptly provided to NRR to ensure a coordinated approach. T' anticipate that the approach utili:ed here mill be necessary for a number of upconing 01. projects, and as directing NRA to take the leed for carrying
.,eut this activity.
- - : . . :~Z: . .
Wil (e. J. Dircks , Executive' Director for Operations
~cc: 8. Cunningham, ELD .
. 3. Hayes, Of .
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- E Review of construction g (CAT)
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!ssues (Req IV) -
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. . COMANCHE.PFAK Major Regulatory Actions Schedule August. Sepiceber _ 0,ctober REQlLATORY Hay ,. June .luly ACTlatts ,
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--------------------------------X Licensing (SERs) lieerJnq Testimony SuMaittals to A51.8 X--------------------------------------------------I
---------X Inspections X- - -------- ----- ------------------- ------------- .
Ancoations x-------------------------------------------I .
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Cl)MANCHf PfAK SCliflRE INSPfCTinti AND REGIONAL- RESPONSIBillilfS' August September Oc totirr
.lul_y April
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- - - - - - - - - - --X Coxstruction Inspec. X----------- --------------------------
HC 7512 .
--------- ---------------- ------------- -----------X Construction inspec. X----- ----
Items Followup 50.55(e) Items, X----------- - - - - - - - - - - - - - ------------- ----------,X Part ?! Items fcliowup Primarily edures Revfew -------------------XTest X -------- Primarily Test Witnessing
-------- ------ 2 Proc -------------
Preoperation Results inspections HC 7513 Rev lew Preoperational -------.------------) Inspec. Items . . . . . . . . . . X-----------
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Fnt1'nwup . Procedares R eview----------------x X Post-OL-----------X X----------- tillnessing Startup Test . Program inspec. MC 2514 g x (final Rpt) x Operational Readiness Report per Module 94300 Items from 55fR 5 x ----------- ---- -- t x-Itens --- --- froei SSE.R..'s 1 -4 x 5ER Verifitatinn -- -- .------- if Bulletins Inspec. X------------ .------------ .---------_-- H!scellaneous x----------_- __----------- _ ------_ 1-- . ' --------- .--------x Allersat iuns Resolution Opi.ra t or 1. ir er.?. ing X--------. -------x i
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April May . lune .hely '- August
- Septe nber I I i ~ l- I l'~ ' I I I i i l I I T- 1 T-i i-Ortnbcri i Plan Approval X 1.
P. Personnel X-- ---X Arrangement
- 3. Obtain 1.ogistic X------ - -- X Support
- 4. Develop Review X---------- --- X Packages . .
- 5. Task Force X Artefing &
Assignments , 6 First Sito X--1 Review Period ,
- 7. Second Site X-- ---X -
Review Period 8 ~. Third Site X-----) , Review Period ,
- 9. Prepare final 1----X Dra f t S5f R
- 10. tianagement X---X Review
- 11. Final Report X----X
- 17. Suleitt Nrport. -
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( . p ~,.. n: . u .1- . . .u a ' .m .. . a. - t r . ;L , - iftlWICAl. RIVIDI TfAH-(TRT) STAFFINr. i . Prograci Manager i 01 Assistant Tech. Reviewer }
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, , i Tlectrical/
7Af0f.~ Civil /rechanical Cnatings Test Program ,
- i Records Instrumentation I
ieader f.eader i.ejder Leader Leader 6esedesv 7 Reviewers 2 Reviewers ? Review'!rs 7 Reviewers 2 Reviewers Mechnical ingr (7) Paint A Coatings Prcoperal i( eal Design Change Contrni ficctrical Engineer Test Enqineers IAC Engineer Piping & Pipe Engineer
./ Configuration Houmnt Merhnical Engineer OA Audt-Procedures Supports Engineer QA Audit-Dogument Welding 8 NDF Fagr
- Control / Records (2)
Civil [ngfaeer Concrete A Rebar QA Audit-Dualif & Engineer Training Structural Engineer QC-Weld HDE Exonr QC-Concrete beinr (? Sessions)* * (1Sessinn' (3 Sessions ) a (1Sessi.sn) (3 Sessions)
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- Each onsite " session" is a mininum of I? days
**In addition f n firnokhaven National l atviratory seview currently in prottress E
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; [, .n.,3% UNITED STATES -27 je y y; e f[ j NUCLEAR REGULATORY CCMMISSICN 3-
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- - 'l: Mr. R. J. Garf, I.tecu:ive 'lica ?residen:
and General .*anager
- C'. Iryan Tcwer
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. S'.iJECT: C:ns: uction A: nisai Ins:ection 50 425/33-12, 50 445/53-12 W s e#e-s : :ne c:nstructio.n a r;f sai ins:ecti:r ::rcu,e:e.c :y - e Off':a :#
.ns:ec- cn anc En7cecemen: (.m :n sanuary o 3
. e:raar/ , .::: anc e:r;ary
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. '4ar:n 3,1333, a :ne C:manene ? ear Steam Elec rt: 5:stier ane y:ur Jai'as
- r c nta ef#ica. E1e C:nstr;ction A::raisai Taam (CAT' was ::m: sac cf eecers of II and a numcer of ::nsuItants. The ins:ecticn ::verec ::nstr;ction activi-tas au:ncri:ec ty NRC Constr;cticn :erni- :.::R-125/'.27.
This ins:ecti:n 's the sec:nc of a series of c:nstruc-icn accraisal ins:ec ' ns
- e4.; ;1annec :y :ne Of'ica :f ',ns:ecticn anc Inf:r:ement. E e esuits O'
- nese ins:ec;i:ns will be usac :: evaluate inciemen ati:n Of managemen: ::ntr:1
- cf nstr;ction activi-ies anc :ne :uality cf ::nstr;cti:n a nu: aar Oiants. -
, . e eac;: sac recor- identi#ies tne areas examiaec :urd g :ne irs:ecticn. ! .si:nin :nese areas, ne ef#ce: c:nsistec cf :e sitec ins:ecti:n :f selectac ar ware su:sacuen :: Cuality Centrol ins:ecti:ns , a ::m:renensive review :#
ycur Cuality Assuranca Pr: gram, examination cf Or:Cacures and esc:rcs, :tsarta-i:n if acrx activities anc inte /iews ,,ita anagement anc ::ner rerscrnei. J.:cercix ; :: nis Ie :ar is an E.tect:1ve St. mary Of ne resul s :f -he ins:ec-
- 1:n anc Of c:nciusi:ns tacnec y -his Offica. E.(:a : fer ne area Of *ne nesting, ventila-icn anc air' c:ncitioning (H'/AC) system, ceficiencies nc ac in instailec har: ware cic not incica:a :ertasive d ailures :: meet c:nstr;cti:n
, installation recuirements. :n :ne H'/AC system, a creaxcewn in wert anc :cali y l ::ntr:1 was icentifiec. NRC Regicn I'I nas Oiscussac nis matter wita /cu anc i: is cur uncerstancing :na: nis ma::er receivec i mecia:a acti:n :y jeu anc l
ycur c:n rac: Ors :: evalua:a anc c:r sc: :nese ::nci- :ns. MC :eg'an :'l wi:-
- n.inue :ursue --is issue wi n you. :r:::: Tansgemen 5: anti:n : :ne esciuti:n # : ner :e si ec :efdciencies icen:id'ac :ur ng ne ins:ecti:n s l neecec.
1 . ! :n ::ntras- : :ne '/AC sys am ndNRCCAT'ns:ec:es#:unc#ew:e#4'encies in'is.insec-'cn:fsafetysyIa:m:i: ng. :SPE :::e -aciegra:ns f:r nis
- icing anc sameles ins:ectec in :nis area sn wec evicence Of gece .crxmansni:.
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v . i, . 4 AMI 11 :333 . 3 Texas Utfitties Generating Ccmaany . A::encix 3 :: nis le::ar c:ntains a lis of ;o antial enf:r: ament. ac-icns
- asec en ?!RC CAT ins;;ec or ceservaticos. These nave ceen rafarrec to -he Region IV Cffice for review anc recessary acticn.
In ac: rdance 41:n 10 CFR 2.790(a), a cecy of nis ie ter anc :ne encicsures will te placec in ne llRC Public Cocument Accm unless you notify :his Offica, by teleonone, witnin 10 days of the date of this letter and su:mi ri: an acclication to wi*.nnold infomation con:ained herein within 20 days of :ne ca:a af his le: er. Suen acciications mus: se c:nsis:an: with ne recairemen s Of 10 CFR 2.7?C(b)(1).
?io reply :: tais tet:ar is requirec a: :nis time. .1RC Regien IV will accress ne ;otantial enf:rcement findings at a iatar ca:a and sny recuirec res:cese will be adcressed a: : hat time.
Shcu!c ycu have any cuestions c:ncarning :nis ins:ecti:n, : lease ::n:ac: us cr
- ne ilRC Region IV Of!ica.
Sincerely, 1 Ricnarc C. DeYoung, Direc::r Cffice of Ins:ecticn and Enforcamen: Inclosures: i- 1. Accencix A - E.tecutive Summary
- 1. A:cencix 3 - ?ctantial Enforcamen:
Fincings
- 3. Inscaction ;ecor 50 125/83-18 50 426/33-12 Distributien
.i f ie IE R/F RCPS R/F CASIP R/F
- 3. 3each R. Heishman J. Tay1cr J. Snie:ek R. Cefoung IIOi f r n ,\
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SUMMARY
an announcac Construction Accraisai Team (CA7', ins:ec:icn was cerf rmec a: :ne
- manene Fesk 5:eam Electric Station curing ne :eri:c .'anuary 21 - Fecruary 2 i *g83 anc Fe:ruary la - Marcn 3,1983.
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- a. m . ..w l- 's tne ;csi-icn Of -he :nstructicn Accraisal Team :na: the resul s of nis ins:ection indicate several ::nstruction program weaknesses. NRC Regicn !7 nas
- een race aware of these weaknesses and is ursuing :nem with licensee manage-ent. The licansee is initiating c rrective acti:n anc/cr con-inuing nis e'#:r :: aescive the identifiec ::ncarns. P :::: anageman: 1::en;i:n ::
- rer ceficiencies icentifiec curing :ne inscection is aeeced. In acci:icn, re 'i C CAT team f:und few deficiencies in f s 'rs:ec:::n of safety system
; ::ng anc ASME C:ce raciegracny for :nis :t::ng. : s:ec-icn :.' sam:les in nts area sneaed evidence Of gcod workmansni .
4
*he identified ::nstructicn =r: gram weaknessas are as felicws:
".. *esul s of ne ins:ection incica ac a trea<::wn in face'ca-icn, 41stai-iation, and ins:ection in ne neating, ventila:icn, anc air c:nci-icning
'hVAC) systams.
- 2. A nuccer of examcles were icentifisc Of failure mee: cr : aria 3:r separation of safety-reia ec cacies ' rem tecnanical structures anc :i:ing, anc sacaretion of recuncant trains of safety sys:ams. Tsis was :ue in
- ar; :: :ne licansee decision not ins:ect ins al'.stiens for recuirec sacaration until installatien is essentially ccm:leta. The NRC CAT ins:ectors are c:ncarnec wnetner; (1) ne ins:ections can :e ef'ec;*vely ,
- ncuc:ac af ar instaliation, and (2) whe:ner acecua:a correc-icn actions can be sc: molisneo sitar installa:icn is :: mole:ec. Correcti:n Of ca:le secaration eficiencies at a later data c:ulc require resea:ing ::rtiens Of system tasting. (Note: This mattar is :eing :ursuec with One iicansee by Region IV, IE and NRR througn site review anc evaluaticn.)
- 3. The licensee's quality assuranca pr: gram cic not ensure :na: certain nanger, su; crt, electrical anc teenanicai ecui: ment was ins allec ne Tates cesign cccuments, anc commensurttely :na: an accro:ria a ins:ecti:n ads C ncuC:ac :: :ne Ia:ast esign Occument3.
1 Fincings als incica:a a num:er of instancas enere enc:nf:rming
- nct:icns were identiff ec; newever, vardeus re nccs (e.;. , :uncni ts 3, ins ection recer:3, vertal, anc etner infcreal retaces; eere asec :
accress anc resolve nese ncnc:nf:rmances. hese me:nces :: no: :::::y dita recuirements :: icentify ncnt:nforming :enci; ions anc revica c:rrective actions 7 Oreven; recurrenca.
- 5. he it:ensee's Oua'ity Assurtnce acci :r: gram sneuic ~aave :een cre effective in cate<. ting anc cc:aining ::rrecticn of de#f ciencies 'n A-1
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8 .5 . s t ,- 1 1 safety-rela:ed werk; such as these in :ne HVAC system, mechanical I ecui: ment, anc electrical c:m;cnents.
- n summary, ne icentified weakr. esses require increasec cecica-icn by manage-en: at all levels := assure completed installati:ns mee design recuiremen s and :nat ins:ection cccumen:stion reflects na :ne :: mole ac installa:icns have been acequa aly ins:ected :: -ne la as: cesign :ccument.
.. AREAS INSPECTEC AND RESULTS Eiectrical and !ns rtmentation Construction: u ui;i:le instances :? cev'a-icns
- Om recutrements were ccservec reia ive :: electrical anc electrical /
?.ecnanical se: ara:icn cri aria. The review cf r:cecures anc rec:r:s asso-c11ted witn electrical inspection activi-ies incici:ac inacecua:a pr:cacures :
assure reins;ection of ecdified, previcusly acca::ec Class lE cemecnents. 9eenanical C:nstructien: Ceviations free cesign recuirements were ocserved in v, ac:e::ac :::e succor:s/ restraints, anc HVAC instaliariens. The existing ~
- r: gram # r :i;e su::cr / restraints does nc accear acecua a : Orc erly verify na: finti as-cuilt harcware meets tne final cesign recuirements. CC ins ections of One HVAC system sue cr dimensicnal features ,ere no: cerf:rmed anc cuc:/accessary ins ection activity c:n:r:Is were n:: cetinec in procacures.
These deficiencies were reflected in :ne as-. built c:nci-icns wni:n cid not c:nform :: design recuirements.
- d:ing installati:n ::nditiens a::earec 2:acuata anc ex:ansive ;r:blems wi:n
; mecnanical equi: ment were cat evicent.
Weldinc/1cndestructive Examina:icn: The welcir; anc NCE a:: ears : te in ac::rcance wt n recu1rements excas for :ne HVAC area. The HVAC welcing activi-ties reveal significant deficiencies. ?r :lems were icentifiec wi n uncersirec
.. elds in structural su::cr:s, inacecua:aly :rainec ins:ecticn :erscnnel, imer:cerly cualifiec weicing ;r:cecures, and inacecua:a welcing ::cumentati:n.
The review of raciegra:nic film of a:Oreximately 30 field weics, ne eview cf NCE procacures, anc :ne intarview cf NCE ;erscnnel, inciucing cemenstra:d:n of NCE technicue, incica as an acequate NCE Oregram for safety system pi:ing in ac:Orcance wita recuirements.
. Civil and Structural Constructien: Cccumenta icn reviewee in -his area snewec
- nat in generai, acrx was :er :rmec in ac::rdanca wi n si a precacural recuire-ments anc :ne licensee's c mmitments.
3 *ccu remen t. 5 : race. and .Matarial Tracea:ili v: ?recurement, On-si a s: rzge in warencuses, inc ma ar ai racanc 11 y were f:unc : te ac:aotamie. Sam les of s:: rage in.:u sica lay-cewn areas anc in-clace s : rage of safs y-ret a:ac ecui: men: reveaiec na s me ecui: ment was nc: ;r tac:ac as recuirec,.
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nree cualf y ::n r:1 ins:ac :rs were t..arvi a .. ......~.. .. a. . .sa...ent or intimica:icn aere ei ner r:cer:y resclved by TUGCC Or aere referrec'to NRC Region **/ :: de incluced in an
- ngcing investiga:icn. :ntarriews anc car-ifica-icn reviews reveaiec na: scme s
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luali v assurance: ne licensee's :uait y assurance Oregram sncui: have :een cre e Hective te conit:rdng anc c:n:rciling safety relatac activi-ies, as exemelifisc by NRC CA-' icentifiec ceficiencies in - e .mW C anc elec rical se: ara-icn areas.
- esi:n Chance Cen rcis anc Corrective ac -ion Systems: In :ne area of cartain angers , su:corcs , e e ectricai anc recnanicai scu:: ent, m
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..........a . 2 .. iii y :: nave an a eca .. ,,i.y.am in-clace a: :ne time Of
- nis ins:ection :: ansurt that field installaticns were ::nstruc:ac to the 1 as cesJ.gn cccumen anc :na an accrecriate :ual' y ir's:ec i:n aas ccm-
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- !a ac. Ir accition One larce numeer f ac:rcx'ea =l e '"Cs anc '.5 :CC There were ins ances wnere Scnconfor'.ances were icentifiac, :u: vari cus retnces f s: e infoma!) were usec :s accress anc c solve :nese cenc:cfor-'ances ra:ner
- nan - e me:r.cc s:ecifiac :y sita :r:cacures. Evi:ence .:f s: rc:ria e correc-
-ive acti:n 1.cc ecnnical jus-'fication c:ule 10: :e te e minec in scre cases.
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AP:EN0!X 3 PC I'IT:.at ENFORC5 VENT :DDI iGS I as a resul Of :ne Construction Accraisal Team (CA7) ins:e :icn of January 11 - Pecruary 2,1983 anc Fecruary 14 - Maren 3,1983, ne foi!cwing i:ams have :een
. Pe#errec :: llRC Region I'l as ;ctential enforcemen fincings (Secticn 1 eferences are :: :ne detailec cr:icn of the Ins:ection Rescrt).
A E*ec:rical nc *ns rumenta-icn Construc:ica
.. Centrary :c 10 CFR 50, Accencix 3, Cri arien X anc FSAR Section 17.".10, h certain inspection activities were not executac :: verify installation ecnfermance witn prececures inclucing caele s: acing in trays, cable tend racit, cacie fill, :acle su::cr:3 anc tray instalia-icn nar: ware (Sec:icns II.3.1.a. :. c, anc !!.3.1.:(1)).
- 5. C n rary :: 10 CF7 50, Ac:encix 3, Critarien X'!: arc 25AR Section 17.1.15,
.- . :he estaclisced ins:ecti:n =r: gram cic no: ;revice acecuata ::ntrols ::
assure that caviations fr:m electrical and electricalimeenanical secara-
- icn criteria as cefinec in :he .:5AR were cr: met j 1:en-ifiec anc ::r-rected (Sections I:.3.1.f. II.3.2.a. anc II.3.2.c(2)).
- 5. Centrary :: 10 CFR 50, A::encix 3, Cri arien 't, F5AR 5ecticn 17.1.5, acc .
IEEE Standard 250, pr:cadures to imclement ins:ec:1cn ac:ivities relative
- :ar:ain assects Of :a::ary maintenanca nave n.c: teen :eveiccec ce tmolementec (Section I:.3.3.c.).
Meenanical C:nstrue:icn 1.'
- entrary : 10 CF3 50, .2ccencix 5, Cri erien 7, :5AR Secti:n 17.1.5, and Q:-CAF-il.1-ES, :artain CC accectac ASFE ;i:e succor:s/ restraints are nc: installed in ac::rcance wi:n the design cccumen :: wnicn -he pt;e succor:s/ rtstraints nere inscected (Section III.3.2).
- 2. Centrary :: 10 CFR 50, Accendix 3, Critaria X and XVI*, and FSAR 5ections 17.1.10 anc 17.1.17, an inspecticn ;r: gram nas no: :een estaclishec to verify and document installation c:nformance :: crawing recuirements in regarc : ;i:e sue:cr'.s/ restraints anc mec.canical ecut:-
ment installations (Section III.3.2 anc 3). -
- 5. Contrary : *0 . CFR 50, a ccendix 3, Criteria 7 anc X , an: .5AR 5ections
'.7.1.5 anc 17.1.10, instatied and GC ac:ectec neating ventila:icn anc a+r
- nci-tening (HVAC) :uc;, su::cr:s anc ecui: ment :: no c:nf:rm : :esi;n recuirements. :n acci-icn, ins:ection Or:cacures nave cct teen esta:;isrec
-Or executec :: verify c:nformance of MVAC su :cris :: :esign Orawings (Section III.3.1 ).
, _ . . _ _. . __ . 2_ _ .
g . -- .- .- . .. ._ .. _ . . . . __a.v; - . . . - 2 Welcine anc Ncncestructive Examina icn
~:n:rary : '.0 CFR 50, Ac;endix 3, Criterien IX are :5AR Section 17.1.9, car atn scaciai :r: cesses reia-ive :: :ne HVAC system were nc: acecua:ely
- n:rcliec, inclucing imcre:erly cualified :recacures; imer::erly cualified ins:ec :rs; im:r::er certifica:icn cf NCE :ersennel (Section IV.3.3).
Civil anc Structural C:nstruction C:ntrary :: 10 CFR 50, Ac;endix 3, Criterien V and FSAR Secticn 15.1.5, civil
- nstructicn :ss :rececures were inacecuate :: ensure na mixer unif:rti y
*ests as recuirec by P.e ASME-AC;-359 C:ce were erf rmec a- ne :rescribed 7 scuency (Sec:icn V.3.2).
- recurement. Stortce and Material Treceability Con:rary 10 CFR 50, Ac:encix 3, Cei;erien 1:::, .:5AR 5ecti:n '.7.1.13,
~P-CAP-8.1, Rev. 5, CF-CFM-8.1, Rev. 1, anc "CP *.0, Rev. ', s :Page Of ce-tain safety-reia ec equi: ment in cu:sice lay-c:wn areas an ins:allec in ne ;ian; was no: cr:ceriy ::r:r lled (Secti:n '/I.S.2).
.'uality Centrol !ns:ec Or Effectiveness
".. Contrary :: 10 CFR 50, Accencix 3, Criterien II and F5AR Secticn 3.3, indivicuais were certified :: leveis of :a: ability witncu ne recuisi a ex:erience cascribed in Regulat:ry Guice '. 58 (Section V:~.3.2.a.(2} }.
- 2. Centrary 10 CFR 50, Ac;encix 3, Criterien X anc F5AR Secti:n 2.3, ins:ecti n rec:rcs were cre ared anc ac:e :sc by L-I ins ec: Ors as ne
ins:ector of record" ra ner : nan :ne recuirec L-II " ins:ec::r of rec:rc" recuirec Oy AN5* NAS.Z.5 (Secticn VII.3.2.:(1)).
Cuality as surance
- 1. Contrary :: 10 CFR 50, Accencix 3, Criterien XVIII and FSAR Sec:1:n 17.1.13Mye not been c:ccuctec i a frecuency er a suf*icien:
ce :n 0.fcentify and correct significan creolems in varicus areas of
- nstructicn; i .e. , HVAC and electrical se araticn (Sec-icn
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Accencix .2, c.r,.ter,. cn X,/. .t anc .. . R .
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. lated maintenance instructions 1cen:1-,section .
- s. .
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'981 anc 1982 were not resolved in a -imely manner (Section 'l!!:.3.2.3.(5)(c)) .
a.
.:ntrary : 10 4..s :.C, a,ccencix .:, Cr*. erien ,1,. anc .: r;.a :ec-i:n .s .,..:.
- rawings ui n cu:-of-ca e revisions anc drawings i n :amacee :r unreac-acie ti-!e :lecxs =ere :resen- in ::nstruction wort areas 1 5ecticn
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. . -:n rary :o 10 CR 50, /.ccencix 3, Cri erien 'I anc :5AR 5ecticn 17.1.5,
- recacures were no: acecua e := assure :esi;n :. anges were :recerly ransmi :ec to :ne Oualf y Centrol organi:a:ica sucn na an accrecriate ins:ecticn could ta :erformed (Sections IX.3.a,1.0 and IX.3.1.c).
... ventrary :3 0 s..R.- :.0, a.ccencix .s , c.ri e r,. a .... an: t.i, anc ...
- en ,ecticns
'.7.1.2 and 17.1.15, nonconforming c:nditicns icentifiec rela-ive *. scce safety-related nareware installations are cc :eing :rocerif accurentec, evaluated, anc cis:csitionec inreugn :ne Carrec-ive Action ?regram.
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-t U!.ITED STATE 5 NUCLEAR R'EGULATCRY CCPM!S35*CM .
CFFICE OF INSPECTION AND ENFCRCIMENT
- '!*5:0M CF CUALITI ASSURANCE', SAFEGUARDS, AhC INSFECT:CN FRCGRANS REACTCR CCNSTRUCTICM PRCGRAMS 5 RANCH Fe:c- Nc.: 50 425/83-13, 50 226/83-12
. :ccket 1cs.: 50 425, 50 446
. ':ansee: Texas Uti!i-ies Genera-ing C:m:any 20013rjan Tcwer Callas, Texas 75201 racili y Nace: C canche :eak Steam Electric Station Uni s '. and 2
- 1s:ec-i:n A : C::anche Feak 5:aam Electric 5 a-icn, Gian :se, Texas and Texas Utilities Generating C ::any, Ca!'as, Texas
- ns:ec-ion Concuctad: 'anuar/ 2a - Fecr;ar/ A, 1933 anc Fear;ar/ la - Maren 2, *.983
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I. 'NSFECT!CH SCOPE AND C3*ECTIVES
~he cojective of this ins:ection was :: evaluate the adecuacy of :nstruc-6 -icn a: the C:manche Feax 5:aam Electri 5:2:1:n (C?SES). This cojective das ac::molishec nrcugn review of :ne construction :r: gram, One quality . assurance program, anc :ne cesign :nange er: gram, wi:n an:hasis en :ne installed har ware in the fiele.
n Wi:nin Ine areas examinec, :ne inspection consistad of a :etailec examina-tica of seiected hareware sucsecuent :: licansee tuality centrol ins:ec-
- icns, a selective examina:1cn of pr0cedures anc recresentative rec:rcs, and ceservation of in-crecess work. Interviews were concuctec with cesigna:ad sita canagers, cuali y ::ntrol ins:ection :ers:nrel, Inc craft
- ersonnel en a routine : asis.
For eacn of the areas inscacted, tne fclicwing was caterminec:
. Is the hardware installed in ac::rdance witn he 1::revec casign?
. :c incivicuals di:n assigned res:ensibi'i;tes in a s:eci#ic area uncerstanc :neir casigna:ec res;cns;;ilities?
. Art cuality verifica:1cns :erformac curing :ne ::nstrue:1cn crecess witn acclicaole hcid cints and are cuali y verifica:1cns c:nductac adecuate inscection accactarce criteria?
. Oc ;ersonnel invcived witn ;uality Assurance /Cuali y C:n rel have ne organi:aticnal freecem :: ;erf rm :neir : asks at:ncu: narassmen: Or intimidation?
. Are management ::ntrols estaolisnec and imclementec :: ::n:rci
'i activi-ies in :he sucject area? l The areas in unich a saiectac sam: ling inscection nas ::ncuc:ac incluce:
. Electrical anc Ins;rumentation Construe:icn
. Mecnanical Construction
. Welding anc Ncncestrue ive Examina:icn
. Civil and Structural Construction
. Procurement, Storage anc lda arial Traceacility
. CC Ins:ec:Or Effectiveness
. Quality Assuranca
. Cesign Change Centrols anc Corrective Acti:n Systams 9
I-i
- .1 i WW93 .
-- - _ - _ - - a'N
.~-
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!!. ELECTRICAL AND INS'AUMENTAT!ON CCNSTRUCT!CN A. 09dECTP/E
> The cojec tve of tne assessmen; in :nis area was := ce armine wnetner safety-relatec elec rical and instrumentaticn c:m:enents were :eing installec and ins ec:ec in ac::rcanca wi:n a :rovec engineering esigns, regula: cry requirements, anc :ne acclicant's F5AR c:mmi ments.
Accitional ecjectives were :: determine wnerner pr:cecures, instructions, and crawings usec := ac:Omplisn c:nstrue:icn activities
.ere adecua:e anc wnether cuality-relatec rec:rcs ac:ura aly ref'.ec the c:moleted wert anc ne ins:ec:ec activi:ies.
- 3. CISCUSSICN
- 1. Electrical Caele Installation The NRC Ccnstructicn aceraisal Team (CAT} irs:ect:rs selec ec a samcle of installed electrical cacie runs :na: nac been ins:ectec y Quali y Control (OC) ins:ect:rs. he sam:le inciucee ;cwer, ::n r:1 inc instrument camies. For eacn of :nese a:ies, :nysical inspection of cacle was mace to ascertain ::meliance 41 n acclica:le design and installaticn critaria relative si:e, yce, Iccaticn/ routing. enc racius, pre action, se:aratten, icentification, ;nysical leading anc succer:s.
The fellcwing ;cwer caoles, ::aling a;cr:xima ely 1,500 feet, aere
< selected fr:m cifferent systems, electrical : rains, :nysical ic:ations anc sizes.
Ca:! e 'le. Tv:e Fe m _Tc E01001513 1/c 750 MC:4 W-iC8 !?!WE301-G2 E?vCE307 -G5 EG100387A '/c
. 750 MC:4 W-208 EP5WE3CO-07 E7MCE3CA-01 E01C0 10 3/c 6 AWG W-120 EFMCE301-09 T3XCSAFSA01 EG10C037 1/ c a/o AWG W-206 E?SWEA02-1* C?!CTAFCSCa AG100385 1/c TRI? TEX W-612 E?SWE301-09 Pane:ratten Eli The felicwing instrument caoles totaling a:creximately 1,CCO feet were selec ec and ins;ectec :: c:nfirm :r.e revicusly sta:ac installa:1cn attributes.
Cacle No. Tvee Frem _7: E0'.25235 '. SHC 74 : air Cen- 5:~v ?ce C2 3CF Ins: :NL 0 *. 15 AWG W-165 EG 135252 W-270 Eia- :aaa--a:1cn Te-m Sox I-ia f:r '.LiaT3'. A samoling of a:Or0xima eiy 1,200 #ee: Of instailec electri:ai
- ntrol ca:la was selectac ?-tm vart:us areas Of ne :lant. The ins ecticn was cerformec :y examination of cacie segments aita*n selectec caele tray sections instailec in :ne Safeguarcs, auxil i a ry ,
C n:rol anc Centairment :uilcings. I sncuic :e notec :na- nts metncc of samolic; :ic cc ::nfirm ne s:ecific u-ing Of :acles. II-i 4.
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j r 4 1 A detailed explanation of this matter is covered in the folicwing paragra;hs under Cable Identificatien. The NRC CAT inspectors o: served a: proxima:eiy -GO feet of in-peccess cacie pull activities. This figura eteresents :ncte c:ntrcl ca:le ;ulis, e:utac in varicus
, areas of :ne plant. These were cacie nu cers EC021923 N/M, E0117573
- anc E0221388.
~
. The ocserved in-process caele pull ac-ivities were :erf:rmec in acc:rtance witn procecural recuirements.
J a. Cacle Scacine in Trav The cesign basis 3:ecifies na accact:tes f:r ca:1e instal-lec in trays require cera-ing basec on neir instal ~ec c:nfigura-tien. Gibbs & Hill 5:ecificatien 2323-ES-100, Rev. 2, Secticn
. 4.2.1.4 states in : art.
- cwer cables run in caole tray snall have maintained cable s: acing acere sc indicated in ne ca:ie anc
~
raceway schedule anc ca:!e uil carcs. Maintainec s; acing between cacles (edge :: ecge) shall :e a minimum of cne uar:ar of :ne ciameter of ne lar as: ca:le*. Texas Utilities Generating Ccccany (TUGCO) Orccacura QI-QP-11.3-26.6, Section 3.1.4, sta:as in : art, d a minimum
, secaraticn of one quartar of the cable diameter snail be main-
- ained between siderail of cable tray and adjacent cacled . The NRC CAT inspect:rs noted tne folicwing cable trays c:ntained i :recerly spaced cedium voltage ;ener ca:ies:
- Tr3v 'lo.
.7123ABF27 m..en.e n0 ce.
T122 ASP 71 7113EA323 7120A8810 7120SECC6 712GABFla 7113SABC6
. b. Cable Send Radius Gibbs & Hill specification 2323-ES-lCO, Rev. 2, Sectica 1.2.2.3, states, "Cacies shall be trained so nat ne minimum :encing racius for ;ertinent :lant cacle training is not exceecec*.
TUGC0 Pr:cacure CI-CP-11.3-25.5, Revision '.6, Secticn 3.'. 1, states in : art, "The QC ins:ec:Or snail cetermine ne minimum tend raciusd for :ne escle being installec or recovec, anc saal' incluce minimum : enc racius ins:ection curing ne surveillanca. II-2
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l The NRC CAT inspectors notac electrical cables that were installed witn less than a!!cwable minimum bene radius in :te
+ folicwing 1ccatiens:
Cable No. er 7vre i:ca:icn
; E012C532 3attery Charger No. SC1EDI EG102592 Sa :ary Charger No. 5C1EC2.2 (3) Train Type W-215 T'.2GC3F32 E0102534 T12CC3031, T12030560 2 1/c Type W-2C6 711GEA337, C1'.G05112
- c. Caole Tray Fili The Ccmanene Feak Statten (C?SES) F5AR, See:1:n 3.3.2.1, states
-' in part, " Cable tray fill criteria generally limit the summation of the cross-sectienal areas of con:rcl cables and pcwer cables
- a maximum cf a0 and 20 :er:an:, res:ectively, of One useacle cross section of :ne tray. "cwever. nese :ercentages may :e exceeced creviced the fcilewing : ccitions are satisfied:
(1) Cacles Oc not ex:anc a:cve ne sica raIs :f :ne taale tray. (2) For Ocwer cable the thermal rating of :ne cacle is not exceedec.
.(3) Cable tray su::ce cesign is idacua:e..."
. Curing tne ins:ecticn of installed electrical cable, :he NRC CAT instec:crs icentified cables nat ex:ancec a:cve ne sice -aiis of Cable Tray Nos. T'.2GS8G22 arc T13GAC31a.
- d. Cable Sucecres Gibbs & Mill 5:ecification 2322-E5-1CO, : age 3 5, caragracn (:. '
states in art, "Where succor:s f:r camles in vertical cacle tray and c:ncui are not snewn en ne crawings, :::-of-ciser succor:s... and acciticnal succor:s if recuirec f:r long vertica' risers snall :e provicec sy :ne contrac :r :: mee: :ne cli: wing requirements: T (1) Sucports shall be Kallem's mesn grips or engineer accrevec ecual... (2) One cacle succer: shall :e crevidec 1: ne :: cf vertica! t raceway er as c!csa :s :ne :o as rac:ical. a su::cri sna : te revicec for eacn accitional in ervai as s:ecifisc in tre f:licwing acle: ... The NRC CAT inspec :rs icentifisc several runs :f 750 :cm 5.3 <1 SHLJ cacle -instailec in vertical riser Tray No. 71'GSA801 inc T11GSABa5 cf :ver 100 feet, ni:ncut s:ecifiec su::cr s. s 3 e ow m. o + e emme. ==me 4
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n-R e. Cable Identification a 2 The CPSES F5AR recuires :na: all Class !E ca:les be identified y J! a nine alphanumeric cnarac:er :ag.
- )'
Gibes & Hill Scecifica:1cn 2323-ES-1CO further requires na: identificati:n tags te placed at ne :ermina-icn poin: Of ne cable; f:r example, in an ecut; men housing or a a erminal tox. i 4 , Curing the inspecticn of ne selectec sample of Class lE centrol t cables, it became evicent that witncut a more liberal use of 1 identification tags, it neuld :e extremely difficul :: Orace -he . pnysical reuting of the samcies seitc ed. This das :ue :: :ne a' cuantity of cacie installed in ne : ray, wnien in scme :ases c:meletely buried tne selected cacies for extenceo distances. 4 It was decided by the NRC CAT inspect:rs :nat One con:r01 cacle - samole wculd be ;;ade by an examination of ca 1e ins:alled in selec ec tray segments frcm varicus areas of ne :iant. 'ahi l e a :nis methcc cf sampling c:uld not c:nfirm ne ::11 P:uting Of c:ntrol caoles, it did ;revice acecuate assurance ina: Otner cuality at:ributes of caole installa-icn nac been met. -k ,1* Discussions with the c:ntrac: Ors electrical CC gr:uo revealed that further difficulties with cable icentificaticn are en-T1 : untered when installed class II c:ntrol caoles sustain in-j sulation damage. The repair :recacure recuires :na :ne caele :: N :e recaired or replacec mus te icentified. 7: ac::mciish nis, the QC ins:ector may often attamet :s trace the caole o its
, terminaticn =cint by .ne " hand-over-hand" rethec. Where nis is not possible, ex:ensive evaluati:n of raceway senecules are mace
- nreugn the process of elimination :: determine wnica of :ne
- acies c:ntained in :na: : ray are cf :ne type, si:e anc reel fcctage as :ne one whicn is damaged. This peccess has :een f0unc
! :: te time-censuming and not always accurate. An exam:le was . given of one damagec caole wnicn was identified anc culled Out.
cnly Oc finc :na: it was no: :ne caole it was :ncugn: 0 :e. A1:ncugn current caole icentification metnods are in ::nf:rmance
'. with aoolicant c mmt:ments, the practice of placing c4 icentifica icn tags only at cable ends has hincerec :ne installation and inspection effort.
n
; f. Electrical Secaration The C?SE3 F5AR, Secticn 3.2.'.1, s:::es in :ar , 'The minimum se:aration c1 stance :e: Ween recuncan: Class *.E acui: men; anc circuits internal :: :ne main con:r:1 tearcs is maintainec at six i nc.9 e s . . .
During the ins:ecticn of cacia ermina-icns wi nin ::n:rci bearts, tne NRC CAT ins:ect:rs icentified mul:icie instances wnere ne six inen separatien be: ween recuncant rain diring nac not been maintained. Scme Of :nese accearec : have :een :ausec by imcrecer training of cacles witnin ne :arel, c:ners aere a II 1
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result of the location of the terminal :cin: en a device anc its
- roximity to a device of the recundant : rain.
The electrical CC cegani:ation has accurentec many of nese c:ncitions by use of :ne Ncnc:nformance P.ecor: (NCR)0r as a
- unenlist item. Mcwever, there accear : be instances wnicn have not been acdressac anc/cr correc ac.
Gibos 1 Hill Scecification 2323-23-1CO, Sec-icn 4.2.2.3, s ates in part, "In the event that the accve separa-icn cistances are not maintained, :arriers snail ce installed between recuncan: Class 1E wirit.g. Fce main con rel cearcs, Service air Cemeany stainless stael flexiole c:ncui: tyce 5353 snali :e used as a
- a rri e r. "
The NRC CAT inscectors cbservec rac installed carriers wecse c:nfiguration =rovideo inacequate prc:acticn :e: ween recuncant wiring as aeil as barriers ins allec ir :ne rain cen:rci bearcs Of a type c her than Service Air Cem;acy stainless steel flext:?e concui: type 5553 as recuirsc. -
- 2. Electrical Cable Termination An inscection of a sei~ectac samole of electrical cacle ene terminations was performec Oc determine c:moliance wi n :ne acclic3 Die requirements. Inscection a tributes incluced rerifica:icn of precer lug material anc size, crecer Icun:ing narcware, ac:ura:e location and icentifica:1cn of tarminal :locxs anc :cints, :recer crimo and crim ing tecl, verification of the calibra:icn status of Of ca:le
- 0o1 and
- ncuc: rs.instruments The MRC CATusec, anc ;r:cer inspect:rs ins:erminating!e ec:ec :a- er.c :erminaticas On ne f licwing cacies.
Cable No. Tvee Locatien EG112132 7/c C?l-ECPRTC-C5 EG015013 9/c C?l-EC?RTC-05 EG127547 3/: CPI-ECFRTC-C5- . 50109846 3/c CP1-ECPRCR-03 E0112367 12/c CP1-ECPRCR-03 E01C9811 2/c C?l-EC?RCR-03 E0110070 12/c CP1-ECPRPR-03 E0139235 12/: CP1-EC?RTC-01 A0113460 2/c C?l-EC?RTC-0; 20123795 3/: CP1-EC?RTC-01 EG113353 12/: CPX-EC? RC3-C'. EG113355 12/c CPX-EC?RC3-01 EG1130901 2/c CPX-EC?RC3-01 E0113318 12/c CPX-ECP9C3-01 AC015325 2/c CPX-EC?RC3-01 E0113331 12/: CPX-ECPRC3-01 . E0133882 3/c C?X-EC?RC3-01 EG113351- 7/c CP1-EFwCE3-02 II-5
y :.. .. . . . . . . a . u . :.. .u o-.: . . - - . . . . L. .. :. j . . 4 . .
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t EG113367 7/c CP1-E?MCES-02 EG113268 2/c CP1-EPMCE3-02
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. EGIC0474 3/c CPI-E?MCE3-C2
; E011335; 7/c C71-E7MCE3-01
( E0113365 2/c CF1-E?MCE3-01 1 E0113366 2/c CP1-E7MCE3-01 i E0113235 2/c C71-EPMCE3-01 E0100414 3/c CP1-E?MCE3-0 .' E01150s, </c CP1-EPMCE3-09
- ^^
- n
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. 4 1,' r. L.5 '. F..: S.-7 L .2. -9 2 E011,3542 2/c CP1-EPMCES-09 c ..35.- S o:- :i.- - e:1.. :u. e.ra.uo.
- w. .. .
E0100890 3/c CP1-EFMCE3-09 E0125664 3/c CP1-EPMCE3-09 EGIC0701 3/c CP1-EPMCE3-08 EG109259 2/c CF1-E?MCE3-08 E0106198 2/c C?l-EPMCE3-07 '
. E0109253A 7/c C?l-EF"CEE-07
- a.1-5 asco- . ., c :..: s . -. . ;. ,
n., The NRC CAT inscec:crs also coservec :he in-crecess termina-icn of tcur
,' Nuclear :nstrumen: System (3:5) Triax connect:rs. 7hese were c:m:leted f.- varicus c.hannels of : e Raac e ?r ection Sys em (R75) sys em as fellcws:
Caole No. Channel EYlic790, IV
- ett
- ta59' 5kSIO5 l
EY110794 IV Two instances of imerecerly termina:ec c:ncuc:ces were no ec in ne folicwing loca:icns, anc were sucsecuently c0cumen ac :y CC :erscnnel
~ '
en an NCR: Caole No. Lecation A0123795 CP1-EC7RTC-01 734-94, 96 (Terminal lugs are not cre:erly -ightenec) E013331 C?X-EC?RC3-01 (Insula:icn camage s green c:ncuc :r 1: erminal :oin: 5 5 ',
- 3. Electriesl Ecui: ment ins:sila:icn A saccle of -hirty-five sieces .cf ins allec electrical ecui: men-
. wert ins:ected. Samoies were selectec ' rem tonn Uni: ; inc 2, casac On system func-icn and safety classifica:1cn. Cemeccents selec ec ircludec the folicwing:
r?..Q
r . .
- st
. ~~
- a. Mc: Ors The installation of four mo::rs and assccia:ec narcware was f es:ec:ad f:r sucn 1:ams as 1cca-ion, anenerirg, ;rcunding, ican-ifica icn anc re: action.
do: r :dentifica:icn . Cemeenent Ccoling Watar Pume .v otor Mc. C?'.-CCAPCC-CIM RHR Pump Mot:r No. 73X-RHAPRH-01M RHR Pumo Mot:r No. 73X-RHAPRH-GEM Safety Injecticn Pum: Motor No. 73X-5:APS*-02M Curing ne inspection of :nese i ams, i: aas ec:ac :y ne NRC CAT inspec: Ors :nat in several instances ne to: r Or ;tmo had no-been groundec. Discussions with ne elec rical GC grouc revealec that ;rcuncing was net regarcac as a :ar; Of Class 15 scui: men-at CPSES anc :neref:re cces not eaca!ve :C ins:ection. The .*lRC CAT ins:ect:rs c:nciucac tna- ne installation activities etlative :: :ne accve alec rical ecui:cen: vere :arformec in ac::rdance wita recacural recuirements. Grouncing f:r mot:r er
- uco casings, wnica is required f:r :ersonnel safety inc cr tec-
-ico, nad not teen perf rmed in severa* ins:ancas.
;eins ecticn cf electrical acuicment is ciscussac in :aragracn 3 of tais section.
- . :enetrations
~~e # ilewing ins allac ::ntainmen: :ere ra :en assam: lies aere ins:ectac:
Number Elevation, Fee: 1Ea9 310 lEla 870 1Ea6 343 2E33 2E6 - The lccation, ty:e, mcunting and identifica icn were cemoarec with :ne installation crawings. CC records associatac ni:n inscection of :nesa items were als reviewec. Activities ecserved anc cccumentation reviewec incicatac acrt :erf rmec in nis area das in ac::rcanca wita *ecuirements.
. Motor C:ntrei Cantars i The felicwing 48C 7 me cr c:ntr:1 :antars OdCCa; in ne Auxi'iary anc Safeguares buticings aere ::::arec : instalia-ion Crawings relative :: Icca:icn, counting anc icantification:
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38 4 MCC Identification No. 1E33-2, C?l-E?MCE3-05 N0. 1I54-2 C?!-E?MCES-C6 No. 2531-! C?l-E7MCE3-01 The ins:alla:icns reviewed incica:ac wcrx <as :erformec in ac::r-cance wi:n recuirements,
- d. Swit:heear .
The fellcwing 6.9KV sw1::ngear was ins:ec ac anc ::meared :: installation crawings reia-ive ; i ca:icn, ::unting, an: identification. Switcheear !:entifica icn Mc. 'EA1, C?l-E?SWEA-G No. 1EA2, C?l-E?SkEA-02 No. 2EA1, C?2-EPS%EA-01 . Mc. 2EA2, C?2-EFI%EA-02
- nstallation activi-fes relative to inese swi engear were ;er-formec in accordance ai-n recuirements.
- e. Statien Satteries The Unit *.125V vital attary recms were ins:ectac, inclucing :ne installed batteries, sattery racks ano associa ec ecui: ment. The '
location, mcunting, anc environmental contr:1 f:r ins alia:icn No. 37-1E01, CP!-E?BTEC-01 anc No. ST-1ED2, C?!-E?S37EC-02 ere ccm:ared with acclica le re:uire. tents anc CC ins:ec:f on rec:rcs. Curing the ins:ection of :nese i ems, ne NRC CAT ins:ec: Ors identifed tnat :nere was a c nsiceracie amount of activity in :ne
- attery recms. This was cue :: prt:araticn f:r relocatien rf ne vital Oa :ery calls so nat rewert of :ne seismic ta::ery racks coulc be ace:molisnec. The ins:ee: rs ::servec :na: ai:neugn :ne batteries were cnarged anc in a stata of eceratien, there were no signs ;cstad :: pr nibit smcKing or Ocer flames. Acci-icnally, concreta chipoing activities in :ne wali : cove One ca:: aries nad left cecesits of c:ncreta en :ne cells :nemselves. Ciscussions with :ne electrical GC gr ue revealec .na: ne 1257 =a::eries rac
- een uncer :ne c:ntrol of TUGC0 since '97?, ar: cea :een cvec
- wice since initial installation. Curing nese acves, rescensibili y f:r ins:ecticn of attributes ass: cia:ac wi--
tancling,.tcunting, cre:actica, anc re-energt:2-icn 4as given :: a contract:rs CC grou:.
~
.The NRC CAT ins:ect:rs ceservec ne activities ass: cia:ac 4i n relecation Of :ne vital :a;;ary calls. Curing nis ac-iv4 y , :we call casings were damaged and recevec fr:= seriica.
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A review of records ass: cia:cd with maintenance of -he vital batteries was made to assure that all attributes of Me main-taranca cr: gram nad been ace:me'.isnec and dccumented. Items such as :leanliness, cali voltages, s:ecifi: ;ravity anc electrolyte
. level nac :een insoec:ec and cccumentac on a regular basis.
- Mcwever, :ne NRC CAT ins ec:ces coservec na: :ne C?SES F5AR, in
;' Section 3.3.2.1, uncer " Testing anc Ins action,' states in part, s
*?ericcic inscacticn and testing of DC systams are cerfor:ec to mcnitor tne ccnditicn of :ne ecui: ment to ensure reliacle
, cceration. . . Ali maintenance anc as-ing crececures anc criteria
, . for reolacemen are in ace rcance wi n !EEE 250-1975, and Reg.
- Guide 1.129. 'IEEE Standar: 450-1975, Secti:n 3.2.3 recuires a yearly eneck anc recorc cf:
8 (1{ Call Cendition (2, Call-to-Call (CetailecDetail and Terminal Visual Ins:acti:n)Resistanca Connection (3) Integrity of :ne Sa: ary Racks
~ ' The NRC CAT ins:ec; rs f:unc :na :ne :r:cadure used :e r'O rm survatilanca of :nese staticn ba:: arias (.10. EU4-701, Rev. 0; dic no- im:lemen: ne recuiremants f:r yearly ins: action of cell-: -call cetailec arminal resistance, nor aas :nere c cu-mented evicence that -hese attributas nac caen ins:ectac cr verifiac.
Ciscussicns witn the TUGC0 Electrical Maintenance Grou: revealec na :ne Electricai :4aintanance ?rocacare 1c. EU1-715 was in :ne recass of being issuee :s c:ntrol ac-ivi-tes associa:ac witn inscection of call-to-cell detail tarminal resistance ;erfor ec
- data.
- f. 125 Ycle OC Svstam Ecuicment asscciatec witn tne c: era:icn of :ne 125V OC System was ins:ectac : verify ::m:liance wi 2 clicaale 5:ecifica:1cns anc crawings. The NRC CAT ins:ectors selec ac :ne f:llcwing sam:le of equicment.
Icentification Nc. Ecuiement 3C-1E01-1, CP1-E.:BCED-01 3attery Charger 3C-1E03-1, CP1-EPSCED-05* Ba::ary Chargar 3C-1E02-1, CP1-EFSCEU-02 3attery Charger SC-1 eda-1, CP1-E?SCED-06" 5attery Charger 17'EC-1, C?1-EC'7EC-01 5:s ic Inver:ar 17IEC-2, C?l-EC17EC-02 5:stic *nvertar 59/1ED1 Cvervoi age Reiay 59/1502 Cvervcitage Ret ay 270C/1501 Uncervei age Reity 27DC/ ~.EC2 Uncervel age Relay 54/1ED1 Gr:unc ?rc acticn Relay 54/!ED2 Gr unc Protaction Relay
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'These items were Originally Uni 2 acuismen: transferred by Permanent Equipment Transfer (P.E.T.) to Ucit 1
*ns allation activi-fes revtewec incica ec work was :erfor:ec ir ac: rdance with recuirements.
;. Emercenev Diesel Generater The electrical ascec:: cf ne Uni: ~1 emergency ciesel genera::r (A and 3), including control ca:inet wiring, were ins:ected for icca:fon, acunting, separa,-icn, ;r :acticn anc identifica:icn.
7hese reviewec as;ects incica ec work nas :er#:rece in ac::rcance w1:3 installation recuiremen s. n.~Merce 0:ersted valves A samele of f:ur ec:Or :: era:ec valves (MCVs' aas setec:ec anc ins:ectec for items sucn as 10ca:f =n, cun :ng, gr uncing,
;re:ac:f on, anc ;re:er wiring. The MOVs selec ec were:
MOV !centificatien No. 1HV4776/28537 do.10611/650 No.1HYS58/BF270925 No. 1HV4709/15142 Ins:ection of work ;erfor:ec in :nis area incicated :ne wcrs was
. :erf:rred in ac rcance wt:n.recuirements.
3 Electrical Cencuit and Cacle Tety Installatien
- a. Electrical Conduit he NRC CAT ins:ectors ecserved 26 runs of installed electricai conduit, associatec :ull coxes, fi::ings, anc :ne ass:ciatec concuit sucocris. Total fcotage of enese c:ncuit runs nas a =rcximately 1,3C0 fee . The ins:ecticn revealed several discrecancies in the area of elec rical c:ncuit installati:n, including many instances wnere c vers :: junctien/:ull texes, anc c:nculets wert not installec as recuired.
Discrecancies in .he icentification nuccers te: ween instalied c:ncuits anc Me CC ins:ec:1cn re orts in :ne CA vaul 4ere ciscoverec. Further ciscussien Of nis ma er is #:uce in
- aragracn 7 of nis section uncer F-ccecure. Acci icnaliy, acrx
;erfcreec may not have :een inscec:ec : ne accrt:riate cest;n dccucent; s ecifically, c:ncui: su::cr instalia-icns. Fur:ner ciscussicn of this matter is f:une in ;a.ragra;n 3 Of nis secticn uncer Ins:ection Prcescures.
Relative :: NIS c:ncui installation, :Me recuirec secart:icn fr m flucrescent lign: fixtures was no: maintainec. The "?!E3
!!-10
s, . . { FSAR, Section S.3.1.4 sta:en in part, dall nuclear instrumenta-ion system (NIS) cables are reu ed in conduit ac::retng :o their cnannel assignment... Aisc, a minimum clear 11- se: ara:icn of 7.vo fee: is maintainec fr:m ::ncui- : electrical ncise scurces sucn as :cwer anc rec ::n:r:1 : soles." TUGC3 ?recacure Q! bF-11.3-29.1, Rev. 3, Section 3.1.2, states, in part, "...a minimum sacaraticn of 2'-C" mus :e main sinec
. :etween NIS c:nduit sys:ams anc flucrescant lign:ing fixtures and lignting systam c:ncuits exceo: for c:nduits crossing at an angle of more than 15 degrees. The NRC CAT tas:ect:rs ican-i'iec One 311cwing installed Class 1E 'l!! c:nduits anich :: not maintain
- ne recuired sacaratien 3r:m fiucrescent lign: ::ures. These were as felicws:
Cenduit No. CliSiG045 1 . C15Y10039
. CSIA-915Y C15Y1CCal i 0. Electrical Caole Trav Eigh runs of installed cacle : ray, ccm: rising 134 tray segments wi:n an aggregata length cf abcut 1,500 feet, were ins;ectac relative :: sue:cr: location, se: ara:icn, :r : action anc ;nysica; leading. Samoles were selectec fr:m ne Reac:Or, Auxiliary, Control and Safeguards buildings. A randem inscection of an accittena! 5C0 feet of :a le ray was alsc c:mele:ac.
(1) Cao'e 7ety At acnments
, The NRC CAT ins ec crs caservec several ins ancas wnere caole tray segments were not cl:ac cge .9er or procerly attacnec
- associatec seismic succce:3.
. These instances were identifiec curing :.9e ins:ecticn of ne follcwing cacie trays: .
Caole Ir3V Nc. Tid 6sCLG6 GR 713.r.CL,O,9 l a
- J4.~ 4.3 T13GCFC20 713GACE94 (2) Cacie Tetv Secarttien The NRC CAT 'ns:ec::rs icenti'ied instances f :acie :rsy secaration :nat cic not mee: ne C?SE3 .:3AR ::mm1:ments.
These instancas reflect a conci:1cns :nat exis in otner . areas of ne clant. E.tamoles Of :evia-icns fr:m recuirements aere icentiac in
-ne areas of recuncan: train secart icn, intarnal ::ntr:i
;anel wiring secars en anc electrical mecnanical sacara:icn.
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The NRC CAT inspectors notac :nat actions taken o correc: separation deficiencies have yet t be imolemented. It is unders:ced :nat :ne use of caole tray c vers anc c ner accac:acle fire barriers will alieviate a numcer of :nese :r:blems; hcwever, Stere accear to be instancas, car-tcularly in :ne catagery of electrical mechanical secart:icn, wnien may act be ::rrectable witncut a significant amount of rewort :s ins alled ccmcenents. The CPSE3 F5AR, Section 3.3.1.a. sta:as in ; art, "The :able and raceway secaration criteria are based on ;reservatica of independence of redundant systems... Cacles of reduncan Class IE circuits are run in sacarata caole trays, c:ncuits, cucts and
- enetn:icns."
"The raceways of one train are se:arated fr:m ncsa of :ne c:ner train by locating them in secarate structures er en cc:csite sicas cf large r: cms or s; aces. Where this is 90: ::ssibie, sacara:1cn is maintainec as :escricec :eicw, Or :y :r:vtsing barriers. The Class II cacles are r utec sucn :na: any single failure in ene train sys:am c es not :ause a failure in anc ner train system."
The CPSE3 F5AR, Section 3.3.1.a ::ntinues: "In ;1 ant areas wnien are free frcm potential ha:ards such as missiles, ex arnal fires, and pi:e wnip, the minimum separt icn ce: ween recundant. cable trays is three feet between trays sacars:sc heri:entally g anc five feet between : ray se:aratad vertically." The .'IRC CAT inscec crs identi'ied :ne fclicwing insta11ec Class IE cacie tray segments wnica cic not maintain :ne ncdiced secaration be .veen reduncant trains: Train "A" Train *3"
! 7130ACG51 Frcm 713GACZ79 713CACG5a Fr:m T'.3GACZ71 713CACG63 From T13GCFCC8 3 T'.2CA5823 Fr:m 723GAC85
- T12CABA42 Frem 723GAC85 I; Gibos & Hill Scecification 2323-ES-100, Rev. 2, Section a.3.2
- states, "... cable tray sna11 not be placec witnin 5 inenes :f C' ass I cr II ciping or c:mecnen: welcs unless c:ner vise a::revec by :ne wners fiele recresentative (act.1:icnal ali:wanca snail :e made for ; ice insulatien). The sacart:icn f r nc'n-sa'ety reia:ac anc Class III si:ing anc cacie tray is :: te a minimum of '. inca fr:m :ne :utsice of :ne ; ice :r insulatten."
TUGC3 Precacure QI-OP-11.3-29.1, Rev. 3, See:icn 3.1.7, s a:as in
- art, " Raceway anc* succor.s snali no: ce icca:ad wi:nin (5) inenes of Class I or C' ass !! : icing welcs r ::m:enen: neics...'
Also, " Cable : ray snali be sacara ac fr:m sioing or :1:ing insulation by a minimum of (1) inen.' II-12 e eu.e e. og.* a to .me hei e 4 -am e.-e w - g em pe w.- e e- ,me -ema* 4 ** aw as ..
' 8, The NRC CAT inscectors icentified Ine fol'cwing installed Class
- 1E caole tray segments which dic cet maintain recuired secaratica
#-:m piping er pipe insula:icn.
Table Tesv No. 713GACE79 GAC 713. css.IS,n2 71305CC78 71305CC79 T'.2,05.C.C72
.....Gesu 9 The NRC CAT ins:ect:rs discussad :ne a :ve fincings wi n the acclicant's lead electrical engineer, and learned :nat :ne a:sif can; is aware of the rcolems. The acclican; is antici:ating 3 :erform rework unere :cssi:le, if nc , c:ner measures will :e evaluatec.
NOTE: NRC Region I'/, 'E anc NRR :ersonnel are meeting with :ne ifcensee a :ne sita curing :ne wee < of A:rii a,1983, : further evaluata :nese and c:ner elec rical/meenanical. separatien c:ncitions. (3) Cacie Trav ' den-ificatien The CFSES FSAR, Section 3.3.1.2, recuires na electrical raceway systems te anysically icenti#ied by a ndre 'al:na-numeric cnaracter ag numcer and celer : ce system, wn4 tn is
- = f centi #y wnetner r no: .ne given raceway :n ai.s safety relatac caeles.
The NRC CAT inscec cr e.taminec :ne selac ac sam:le of Class II ca:le ray anc c nduit :: verify tnat icen f fication tags and color c:cing were cresent, anc ;re:erly accliac, rela-tve a location, ma arials, anc c:nformanca ui n :esign crawings. All har: wart ins:ectac :: nese a::ricutas satisfiec :ne
, identifica icn recuirements.
- 5. Instruments:icn The NRC CAT ins:ectors selected a samole of instrumentation
- m:cnents wnica ecnitor : recess variacies ::merising :ne Reac;;r 3* : action Sys:am (R.CS), C:mocnen C cling Sys:Tm ':051, anc ne ingineerec Safe y Features Actuation Sys am (55F:5).
*nstrumen ccm:crents were reviewec :: :etermire t' installatiens were ac::mciisnec in ac: rcance ui n casign :rawings, a::licacle
- ces anc s:ect'ica-icns. Itams sucn as :ca:icn, cun ing, icen -
fication and protacticn were ::mcarea wi n instaliati:n :rawings 3:r
- ne felicwing ::macren s:
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' Pressure Trans. sitter Flow Transmitters 1 ?T-544 1 FT-2a64 A 1 ?T-2133 1 FT-2066 3 i ?T-535 FT-2463 A 1 ?T-53a 1 FT J21 1 77 4252 1 FT 425 1 eT-21653 1 FT 414 j 7
P essure Swi :n .: Tow Incica:19e Swit:n i 75 4519 . .!5
- 4550 tj-Level incicatina Switen
- 1 LI5 4754 o The NRC CAT ins:ectors ecservec ac:roxima:aiy 700 fee :f ins: ailed
, instrument ucing, sup:cr:3, anc assccia:ac nar: ware. Tu:ing was examinec to verify sucn itams as, Orc er ma:arial, slo:e, scun ing, sacaration, and color ceding. The ins:ec:ces also examined ins:ec-tien 'recorcs assccia:ad witn the instaliation of :ucing anc : sing succor:s.
. Curing the ins:ection of instrumen: tubing, the NRC CAT ins ect:rs actac :nat in several locations, :ucing runs had sustainec mince a damage; specifically, dents, flattening, and disfigured instal-la:1cn. It was de erminec ::a: this damage was eue :: ::nstructicn activities wntch had cc:urred after CC inscection of :ne u:ing.
, Camagec areas were recortec :y the c:n rac :rs CC anc recer ac :n an NCR. -
0:ner Observed activities anc reviewed cecumenta:icn indica ad werx
;erformac in this area satisfiec ne accrep.riata recuirements.
1
- 5. Calibratien The NRC CAT ins ec ces perfor=ed an examination of -he :n-si e 5rewn i Rect (3&R) Calibration Facility. Inclucec in :nis examination was a review of procacures and :ccumentation asscciatec wita :siitra icn activities, an ins:ecticn of the calibraticn facility to insure ccmcitance .vi n -he environr.antal c:nci:1cns s:ecifiec, anc :ne ins:ection of a selec;ec sample of caliera:ac ::cis anc ::mtenents.
The NRC CAT ins:ec Or selectac ne fciicwirg samele of reference standarcs, mea'suring and test ecui: ment, anc certiff i c :ci s. These aere ins:ectad ;c verify :r::erly documentac calibra-icn status,
. :rocer s:Orage concitiens, curren: anc 1::r:vec :alibraticn recer s On file, and tracancility of items associatac wi:n ne sam:les
- hi s:O ry fi
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Reference Standards No. RS-215 -Cigital Multimeter No. RS-C86 ?hotogra:nic Ste: Tacie: Mc. RS-020 Cu: site Mitrema ar .. Mc. RS-094 Go-No-God Gauge !- No. RS-067 Cacace Resis :r Measurin: and Tes: Ecuf: men: No. M&TE 1432 Oynamcme:ar No. M&TE 1132 Pieremeter Ce: n 3auge
, No. M&TE 1553 Megger Tes: Se:
No. M&TE 0155 vernier Cali;ers No. M&TE 1270 Soil & Aggrega a Steve No. M&TE 1961 0xygen Analy:er No. M&TE 2409 Cry .111: Thicxness Gauge No. M&TE 2338 Cur-en: Transformer No. M&TE 1829 3at ery Megger 7es er
'ic. M&TI 1995 UI rascnic Th':% ness lauge Certifiec Tools No. CT-1365 Crimoing Tcol No. CT-C608 Tortue Wrenca No. CT-1606 Cr,imoing Teol
- ams reviewed satisfiec acclicacle recuirements. This su ject is also ciscussed in Section 't!!! of tnis repor l
- 7. 3 rec'edures
*he MRC CAT ins:ec: Ors examined a:crevec TUGCC cc:uman s : /eri fy
- na: instructi:ns , procacures , anc crawings asec : ac: molisn elec-Orical activities affecting cuality ::ntain :ne accrecria:a ins:ec-
-icn and/cr ac:actanca criteria.
. Electrical CC inscections are ceing ;erformed in ace:reance with :ne TUGC0 Electrical Inspection Manual. Curing :ne review cf :nese cccuments precadural inadequacies were icentified in :ne follcwing areas:
The inspec:ces ceserved examoles of electrical ::ncuit. installatiens anose icentificatica numcers did no: ?.at:n ncsa indi:stac :n ne OC inscection recer s in re vault. Discussions wi n Electri:al GC
- erscnnel revealec na: scce crcui s are ins ai'ac 41 n a uni-icentification num er. These ::ncui s are ins:ec ac anc :ccumentac in ac:Orcance ni n ne acclicable CC :recacure. Sucsecuen- ne ins:ection, ne icentifica-1:n nuccer may :e ::angee in ac: rtanca di-n I -E. -1 (E-1) and ~;em 13(c) of 2323-El-1.CO , nnica s a:e :nat, "The :nire enaracter cenoting :lant num:er 0,1, Or 2 is ;erers ac Oy ne : meuter f* m ne firs :acle rautac :nreugn ne ::ncut ."
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7
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- 1. - .
r 6 i~ d e a-A e inese identifica:icn cnanges are inc:r; crated en Field Structural j Engineering (FSE) drawings and sen: to the fielt f r construction
-I use. H: wever, :ne FSE drawings are not centro lac by One Cocument a C:ntrei Center (OCC) nor are :ney usec 3" GC for ins:cc:icn. As 2 2 resui cf this, -:ne CC ins:ection esc:rcs ay not incica:e one : rue U fcentification numoer of :ne installed c:ndui:; further, :ners art
- } no crececural recuirements wnica accress reinscec-ion cf :nese u i ams.
U A list of over ICO concuits affected :y similar cnanges were
. reviewec by the NRC CAT inspecters. These items were fcunc curing i randem ins:ections :erformac y :ne CC gecus. In scce instances,
, l ne list incicatac cnanges ne; only in :ne uni: designa ;r, u: in 1 ' voltage level anc reduncant train casignatiens as well. Ins:ec-icn
" :recacures wnica do no: acdress reins:ection of mcc'fied. Orevi0usly
! accepted com:cnents does no a:; ear adecuate.
o
!. :ss:e::icn Rec:rcs
}
j The NRC CAT ins:ec::rs reviewec rec rcs genera:ec f r ins:ecti:ns N - nat were perfor=ec relative := ca le tray, c:ncuit, electrical
; cacie installation, ca:le termination, electrical ecui men:
. installation, seismic su::cr:s, instrument :: ing, acc ins rument installation. -
- Assessment was performee in this area :: deter:f re wne:ner
- inscection racercs have been Or
- erly prepared, maintainec, anc c:ntained documentec evidence of inscaction ccm:letien anc results.
Electrical and ins:rumen:ation ins:ecti n rec:r:s were s:: rte in ne OA rec:rcs vault, and were identifiable and retrieva:le. Ins:ecticn records were c:m:leted in ac:Orcance wi-h -he a::iicable
- uali:y contr:1 precacures. The rec:rds icentifiac :ne CC
, ins:ector, the type Of c:servation by prececural reference, acce tacility, anc reference : Occumen s pertaining :: identified feficiencies. The NRC CAT inspect:rs questienec the adecuacy of cecumentati n asscciatec with reins:ection of previcusly acce;;ec items. In the case of scme electrical ecuipment receres, ne reint:ec-ion
. signatures for ecui: cent tna: was eticcated per ce gn Change Authori:ation (OCA) were not fcund in :ne ecuieme; rec:rcs :ackage,
- u were on a traveler initiatec by tne civil /structurai gr0ue.
Atcitionally, at ributes verifiec sy :ne criginal ins:ec:1:n re: r: nere nc s ecifically accressac in :Nis sucsecuent ::cumentaticn. The NRC CAT ins:ec Ors were unacle := cetermine, frem :ne rec:rcs availacie, the extent of :ne reins:ection nnica das :erf rmec. Similar c:nci:10ns were : servad in :ne review of etcores associa ac 4 aitn inscection of electrical : ncuit anc ca:1e : ray su:: r:3, many of which have been modifisc :y :m cnent recifica:icn :nanges (CMC 3} - wi:n mul;i;1e revisiens, anc wncsa ins:ection rec:r:s indicate inscection dates wnica do not reflec :ne dates of :ne lates: revision of the CMC (inis matter is ciscussed in cetail in See:icn IX of :nis rt:cr;). II-15
. . . . . . .- n. . . . ,
. . . a. .. .. .. .
A4citionally, there were examples of cccumentation Of c:nditions wnich deviate fr:m recuirements by use of systems c ner than the NCR system. Examoles include the use of Recuest For information Clarification (RFICs) :o cccumen: ceviaticns f*:m recuirements relative to electrical sacara:1cn. The engineering respense :: the RFIC was, in effect, a disposition of the priolem, and in scme instances was used to initiate field rework. The use of the RFIC in
- nis manner is considerec contrary :: CA Or gram recuirements.
Also, :ne NRC CAT inscectors found that cunchlists were usec to Occumen: ceficiencies. Scme of :nese ceficiencies were c:nsicerec o be noncenforming c:ncitions y the NRC CAT ins ectors. Follcwing the review of inspection recorcs. One NRC CAT inspectors related the folicwing concerns to the SAR QC su:ervisien: That seme rec rds co not snew :ne exten; nor cc :ney acequately reflec :ne reins ecticn activi tes for :revicusly acceptec i: ems, at the metacd of Occumenting ceviations from recuirements relative to electrical separatien, and :ne etned of icentifying gg anc dispositioning ciscrecancies tnrough :ne exclusive use of punchlists were no: in ac::rcance wi:n ;A pr cecural recuirements. . A-- e II-17 a php *We p.g
--c .. ,-,.---c -
4 . . . _ _ . - . - - _ . - . _ . .. . _ . . . _ . _ _ _ _ . -. ,_ . ; s !!!. MECHANICAL CONSTRUCTION A. Obiectives The cofective cf tne assessment of mechanical c ns ructicn was ::
- ecermine if installed anc OC acceotea safety-relatec mecnanical items conformec to engineering :esign, regulat:ry requirements and
- licensee c mmitments. Acciticnal cojectives aers :: cetermine wnether crocecures, instructions, and drawings used to acccmolish !
c:nstruction activities were acequate anc whenner cuality-related
. records accurately reflected :ne c moleted work anc ne ::meleted activities.
- 3. Oiscussicn The specific areas of mecnanical c nstruc-icn :nat were evaluated were piping, : ice succor:s/ restraints, mecnanical ecui ment, and
, nea ing, ventilating anc air : ndi-icning (#7AC) ecui: ment. -o a ;cmolisn the Ocjectives sta ec acove, ne felicwing ac-ivities were
- erformed in eacn of -nese areas:
A detailed fielc ins:ection of a samoling of CC accected hardware. A review of crececures and cocumentatien. 31scussicn with res:cnsible CC anc Engineering :ersennel to determine overall '<newlecge of site pr cecures, inspection anc ac:aptance cri aria, and :: identi#y or c' ems w 3 Or:cacures, d design / field engineering /CC interfaces, ins:ect:r cualifica:icn, and QC indecencence.
- 1. Picine -
A samole of pioing runs was selectac :: incluce cifferen systems, buticing loca:icns, c:nfiguraticns, anc si:es. The following lines, c aling approximately 5CO feet, uere selectac for inscection: i System Orawines Sf:e. Ci ameter i Containment Scray 3RP-CT-1-SB-027, Rev 3 2" Auxiliary Feedwater 3RP-AF-1-58-GC5, Rev 11 3" i 10* ERHL-AF-1-53-CC5, Rev 2 Containment 5: cay 3RP-CT-1-58-023, Rev 3 2" Resicual Hea: Removal 3RP-RH-1-RS-C03, Rev 5 3" 3RHL-RH-1-RS-C03, Rev 2 Resicual Hea: Removal 3RP-RH-t-R5-CO*., Rev 13 '2" 3RHL-RH-1-RS-001, Rev 2 Auxiliary ~eecwater 3RP-AF-1-58-025, Rev 11 l' BRHL-AF-1-SB-025, Rev 2 III-1
+. '-- . . . . . . .
1 ... Systam Orawines Size. Diameter. Chemical & Volume 3RP-CS-1-RS-C23, Rev 7 2" Control BRHL-CS-1-RS-023, Rav C Safety Injection 3RP-SI-1-AB-002, Rev 5 A" SRHL-SI-1-AB-002, Rev 0 The above runs were inspectec in the field for prc:er configura-tion, identifica:icn of valves, surface c:nci-icn, valve orientation, bolted flange connections, interferences and support / restrain: iccatien, anc functicn. The # llcwing cccuments proviced tne :asic acceptance critaria ?:r ne inscections:
. QI-QAP-11.1-25, Rev 9, "ASME Pipe Facrica:icn anc Instal-lation Ins:ectiens"
. QI-CAP-11.1-31, Rev 5, *nstalla:icn Inscecti:n of :dechant -
cal Joints *
. Applicable piping 1semetric drawings (3RPs) anc hanger location isemetric drawings (BRHLs)
Pipe Suoports/ Restraints are installed arc inspected to detail drawings. The NRC CAT inspec::rs utilizac -he ERHL iscme ric a verify succor / restraint function and location as a eneck that the piping was installed and succortad/ restrained as analy:ed by the designers. Not all su::crts/res rain:s na: :een ins alled at tne ' time of inspection. The piping c:nfiguraticn, valve identifica:fcn, su:ccr / restrain Iccatien anc function, anc flanged joint ,r.akeuc a:ceared to conferm :s recacural crawing
-recui rements.
~
The Ccmanche Peak Steam Electric Sta-ion (CPSES) :r: gram addressing IE Bulletin 79-14 (79-14), " Seismic Analysis for As-Built Safety-Related Piping Systems", was reviewed.
?rocedures governing these activities are as folicws:
. CP-EI 4.5-1, Rev 8, " General Pecgram for As-Built Piping Verification"
. CF-QP-11.13, Rev 5 "As-Buil: Verifica-icn
. QI-GP-11.13-1, Rev 3
- As 'Bufi: Picing 7erifics-i:n In:;ructicns" The 79-1A ;r: gram is sasically ;ar cf a c:ntinuing as-built /picing analysis iteration prccess. The as-cuil-III-2
. o - E survey package for stress problem 1-017 (a partien of the Safety Injection System) was examined and the as-built survey deficiency punchlist comcuter printcut for creblems 1-003,1-006,1-C07,1-010A ,1-010C,1-0193,1-021 and 1-023 were reviewed. Pr: gram ce: ails were ciscussac with
, ne Texas Utilities Services, *nc. (TUSI) Tecnnical ". Services Sucervisor (rescensible for engineering as:ects)
- anc the TUSI QA Scecialis- Su:erviscr and As-Euilt
, C:crcinator (res;cnsible for field inspections).
As a result of this "As-2uilt" review, the NRC CAT ins ec-
- ors identiffed c:ncerns as t: the exclusive use of punca-lis s to document discrecancies Oe: ween detail crawiqcs and
, as-cuti: nareware (versus a cocumen ac/controllec metnoc of identifying, correcting anc creventing recurrence Of deficiencies). The acove procedures co not involve :ne nonconformance prececure or otherwise involve :ne cuality assurance program in accressing discre; ant concitions on QC ac:actec cicing anc sue:ce s/ restraints. Scecific examcles of this c:ncern are ce:ailac in :ne su::or;/ restrain:
section of cnis re crt. ice re!acec Or: gram recuirements are discussec in Unis report in Secticn X.3.2 "Ccerec:ive Action Systems."
- 2. Pice Succor:s/ Restraints
~
ASNE pipe supcor / restraints are fa:ricatec, installec, anc CC inspected :: detailec drawings crecared by ITT Gef nnel, Nuclear f Pcwer Se'rvices Incustries (NPSI), or TUSI Sipe Succor Sngineer-ing (PSE) and acclicable Cemcenen: Mcdification Car s (CNCs). Small bcre (less than Zi* dia.) Class 2, 315 su:: r:s/ restraints may be shown en typical drawings or en "engineerec" ce:ailec drawings. Sucperts may be final" ins:ected seversi times based en subsecuent changes to criginal design; :ecause of the issuance of revised design drawings, anc crimarily Oy tne issuance of CMCs by PSE.(field engineering). When engineering consicers that the ' final as-cuil " stress analysis has been performed anc :nere is a hign procacility na: furtner changes t: ne succor will not ce recuired, a 7enacr Certified Crawing (VCD) fer large bore supccr:s anc a Cesign Review Orawing (CRD) for small core succcr:s is issuec. These frawings inc:r: crate any "informaticn :nly" :y:e CFCs anc any mccificatiens necessi a:ed by load cnanges. CC Oren ir.s:ec s
- ne sue:or s ice : nose features revisec Oy ne VC /)RO anc 'Or covicus missing : arts.
The folicwing samcie of 21 installec anc OC ac:ectec :i:e sue:cr:s/ restrain 3 ere selectec for ins ecticn c :revice a variety of types, si:es, systems and Icca-ions: I::-3 __._,m=.._y .._,s_.__
a
' i.
4 Succor / Restraint No. Lcca:icn Size Class Tvee
.e RC-1-015-708-C41R Cent. 2" 1 Strut CC-2-AB-023-002-3 Aux. 2" 3 Strut
. S*-1-58-238-C06-2 Safeguards r l'" 2 Box e S*-1-068-706-C42R Cent. 2" 2 Scx
'. RH-1-004-C07-532R Safeguards 12" 2 Strut CS-1-357-001-522R Safeguares 5" 2 Strut SI-2 -031 425-Y322 Yard 12" 2 Box CS-2-063-413-542R Safeguarcs 3" 2 S: rut RC-2-121-401-552R Safeguares 3" 2 Strut CS-2-012 403-Ca2R Cent. 3" 2 Strut CC-1-136-7C6-263R Ziec r. 3" 3 Strut SI-1-037-005-532A Safeguar: 2" 2 Ancnce FW-1-017-702 -C52X Cent. 12" 2 Snubber CC-1-116-037-F43A Fuel 12" 3 Anence CT-1-002-008-532R Safeguares 16" 2 Strut CC-1 -116-006 -F33 P. Fuel .2" 3 Stru:
RH-1-003-002-542R Safeguards 12" 2 Strut CS-1-063-008-5222 Safeguards 3" 2 3cx
- 51-1-031-046-Y325 fard 12" 2 Scring RH-1-001-C01-C415 C:nt. 12" 1 Scring SI-1-092-008-C11X Cont. 6" 1 Snubcer CS-1-001-016-Ca2X Cent. 3" 2 Snubcer CS-1-055-010-5225 Safeguards 4" 2 icring RC-1-075-026-CSIR C nt. 1" -
1 Strut Scme of these sucparts nad VCDs issued, seme cid not. 7he above su:perts were ins;ected agains: their detail drawings and CMCs f:r c:nfiguration, identifica:icn, loca-icn, f as aner/ ancnce bolt installation, clearances, memcer si:e, and damage /orotec-
- icn. *n additien, acprcximately 200 acci-icnal succcr s were Observed in the field for cbvicus deficiencies sucn as icose or missing fasteners, imer0:er clearances cc angularity, camage anc improcer excansion anchor scacing.
Acceptance criteria fer tne field ins;ections are :en ained in the folicwing documents:
. Detail suppor / restraint drawings and a:plicable CMCs
. Typical drawings including small bore General Notes drawing CP A-C01 GI-GAP-11.1-23, Rev.16 "Facricaticn , *ns ai' a-icn *.ns:ec-tiens of ASME C:::enen: Su:cor:s, Class 1, 2, anc 3"
. OI-QAP-11.1-2SA, Rev. O,
- Ins aliation Ins:ecti:n Of ASPE
<3 Class 1, 2 and 3 Snu cers" ,
. OI-0P-11.2-3, Rev. 1*, - "Torcuing and Scacing :f Cancrete
. Anchor Soits"
# m- - e. g
~
l. The follcwing discrepancies were icentified cn -he supports / restraints inspected:
. U-bolt c:nfiguratien not per drawirgs:
CS-1-063-028-532R AF-1-059-001-333R H-RC-1-RS-039-015-2 H-03-1-RB-017-001-2 H-RC-1-RS-C38 -CC A-Z
. Lug to restrain; clearance excessive: j 00-2-019-007 ~33R (11/32" vs 3/32")
. Dimensicn not per crawing:
RC-1-075-025-C51R (2' 4 3/a" vs 2' 101/2")
. Richmond inser an:hcr 001: tarsacs no as recuired:
C5-1-C01-016-Ca2K
. Snubcer load cin missing:
FW-1-017-702-C52K
. Void in concrete near c:ncrete ex: ansi:n an ncr:
SW-1-102-716-Y33R
, L:ose stru: Tecknuts:
SW-1-003-C02 233R SI-1-Caa-025-3320
. Missing / broken cetter pins:
, . la supports
. Class 3 hanger mismarked as Class 2:
H-GH-X-AB-0042-003 Numerous other instances of loose locknuts, U-bolts, and missing or broken cottar ains were caservec en Class 5 succor:s/ restraints :nat were uncer :ne TUGC0 CA/GC Or: gram. 3:ntri-Outing to ne creciems icentifiec sith U-ool- installations are c:nfusing crawings wnica sacw U-boi s witn */15 . inen :learance recuirec, cut also witn dcuole nuts on :ne :uts'ce of :ne retaining late. Clear accectanca criteria is needoc :r d :e :: inscections :eing performec. .
4 . . e . , l The NRC CAT inscec crs witnessed the QC ins;ection of fifteen ' small bore supports and six large bore supports :: the VCD per ' peccadure C?-GAP-12.1. Twelve of the small bare succorts were satisfac:Ory. Three were unsatisfac:Ory cue o crafting /dasign errors, primarily cue to inc rrectly incorporatec CMCs int the VCD. Three of :ne large bcre sup cr:s were unsatisfact:ry with an imcrecerly welced brecket, two uncersi:ec weics, a missing high streng:n pin and an 1: Orc:er clama spacar.
- Ciscussions with :ne Sucervisory Authorized Nuclear Inscec: r (ANI) indicated that there were deficiencies witn the accepted 4
hangers anc difficulties in ce:armining ne s ecific activities
- errormed curing ne facrication/installa:icn :nases. The ANI crfica nas recently teen receiving cemcleted ASME sup;crt/
restrain packages for review and accactance fer ne ASME Ccde Cata Reccr . Cf the initial cackages of CC accastad VCD su:pce:s/ restraints tha were ins ected in the field by :ne ANI, a:pecxi-
. ataly 10% (13 of accroximately *.30) have been returnec to 5&R
%C cue to undersizec cr etnerwise unac:ac a:le weids. The 2N was inscecting :nese installa-icns #ce casic c:nfiguration anc l welcing ce: ails enly. Of accr xima aiy 555 vencer :ar-ified, CC inspectec succor:s forwarced to the ANI en or before Fecruar/
23, 1983, apareximately 100 had been returnec for ceficiency' corrections. The NRC CAT ins;ect:rs reviewed -he ce#iciency :uncalists generated by CC curing :ne VC0 inscection. S'ix puncalist items f r hangers were selected (biasad saccle) to verify that noncen-forming c:nditions were being Orecerly icentified inc action prevent recurrence nas ceing taken. Twc of :nese i;ams, an i=craperly installed U-bcl: (CS-1-C53-Cd6-522X) and ex:ansion anchor scacing nct per drawing (. 4-1-095-CCS-C525), : are :en-
- ) sidered by One NRC CAT inscec: Ors :o be nonconforming conci-
* ' tiens, but were not documented by CC en NCRs. Su:secuent :: :n.e
' NRC CAT inscect:r recuas-ing an. evaluation of these i ems by CC, f
NCRs were issuec. Three Icre items concerninc uncersi:ec, missing, or discrecant delds (CC-X-032 'CO 443R, CC-1-151-CCa-553R,
/ CC-1-131-013-543R) hac been documented en NCRs. Mcwever, review of tne documentation ;ackages anc acclicable revisions to ne succort/ restraint drawings and CMCs indicated tna ncne of nese weids had been changed by drawings or CMC sin = **a " final" CC l ins r** n accectanca of nese sue cr:s.
l q.
.a n . .. .. ..
V - vd . . . , . . . . l , Ncnconf:rming c:nci-ions notac during ne VCD program are accarently no- being Orc erly accumentec cr identifiec to effec action :c arevent recurrence. I I I [p...=0
1 .' 4 Ouring the review of ne 79-14 "as-built" program discussed in j 5ecticn 3.1 acove, :ne NRC CAT inspect:rs selected eign: items from the walkdewn survey to determine if discrepancies on CC accepted sucacris/ restraints were :eing :roceriy dis:ositienec. At leas: :nree items (MS-1-C03-CC3-C725, SI-C29-C19-522R anc SW-1-025-C03-403R) shewec ac:arent ncnc:nforming ::ncitiens for wnicn no changes had teen s:ecified (en crewines er :MCs) since CC acceptance of the succor:/ restraint. J
. inese canditions were not cccumentec in any system :na: aculd rovice icng term correctiva action, but were referred : :ne construction organi:ation via
, : -- ' memo. It is rec:gni:ec nat :.9e as-cuil survey eam may not :
<- using the lates: CMC issued agains tne succor:s they are '
inspecting, which may make tne de armination Of a conc nformin condition difficult. Mcwever, it ac;eces tna: nonconforming c:nditien cated during :nis Or: gram are - ' " M The ins:ection of Class 5, seismic su::cr:s was a :ar: cf :ne TUGC0 CA/CC prcgram as delineated in ar:cecure CI-CP-11.15-1, Rev.14, " Installation Inscections of NN5 Seismic Ca:escry II Succor s for Class V Picing". Paragracn 3.15.3 Of :nis crece-cure termits documentation of discrecancies f;und during ins:ec-tiens ei ner on *nspection Recor:s (:Rs) or Ncnc:nformance Recor s cirected :y the QA/0C Civil Meenanical Sucereisor. t o_ manuai. Altncugn irs reia:ac Oc base metal defects tre incu t -
- e- m. M I
,-the *: rend. analysis," and corrective ac- 9a j - .
- t. ----- _
,, fa
.. 1spos1:1cne .
A review of crocacures gcverning safety-relatec :i:e su::cris/ restraints incica ac :na- nere are Or:cecures ::vering many as;ects of ne cr gram (facricati:n, inscec 1 n, 7CC aal<c wns, l :re-nycro wai'<ccwns , Ore-turnover wal'<cewns ) anc :na: s me are cuite ccmcrenensive anc Or:vice ce:silac inf:rma-icn anc accec-tance criteria. Mcwever, :ne rececures eiarec c ins:ecticn of succor:s are scmewna: :nfusing and unclear as to ne s:ecific attributes recuiring inscection. For examcle, ne written instructions /checklis for 1CD ins:ections recuire a verifica:icn :f su: or " Onfiguraticn" . The intan: Of :nis itam 4as not clear. The inter reta:icns Oy CAi C management, :C su:ervisces, anc C inspec crs f erec curing :iscussiens ai:-
- r. . .k."#
. ...-s .. -- _ -
m .
c ._ .. . . __.__.:_m . . . . . _ _ . . . --_._.._ _ , . . _ _ _ _ _ _ . _ _ _ < ._ _. . . .;_ _ . _ _ ,1 ; 3 1 . i t 4 " NRC CAT memcers varied fr:m a detailed inspection of many .i attributes to a verification that in fact a snutter is installec wnen a snu:ber is indicated as recuired. Walkdcwn prececures also do net :r0 vite sufficient s:ecific ins:ecti:n at ributes
- anc cnecxlist/marxuc c ntrols for identifica-icn anc correc:icn
.j cf the large numoer of icose anc missing cot er : ins anc i fasteners. . A review of :ne dccumentaticn cackages for ne 21 succor:s/ restraints listec previcusly was perf rmed. The packages are cifficult to follcw cue to the large numcer of changes involved with an average of ver five CdCs :er su::cr and as many as 16 on one succor:. Previcus ins ections are no: voicec. :: was observec :na fr:m a total of 55 ins:ecti:n re cr:s, covering accroximately 110 total attributes, :nere was only one incica-tien of an unsatisfac cry IR and only one ,1CR was written. For non-ASME succor s, GC rejects 20 :ercen cf ne su::cr:s sub-citted for final ins:ection in .sedi-icn a re:uiring immediate c:rrection of minor hardware items (icese Oci s, etc.). The near perfec record in the ASME 3r gram incicates nat arcolems are being turned back to c:nstructicn f:r resciu:1:n (:nreuen ~' immediate correction or issuance of CMCs). n
~
c: a # s. Review of 14 of these packages indicated the follcwing discre:- ancies
~
Althcucn the Multiole Weld Cata Carc p.au i na a c. rag aicck, in numer us ins ances ne
- icg incicates CMC revisiens and entry ca es rucn later tnan :ne CC ins:ection signcff ca:es (same entries ever one year after ins:ecticn) . As the ::versheet Ins:ecticn Ae: Ort Oces 90: lis CMC revisions, it is not ;ossible in most cases o icentify wnich CMC (tesign document) was usec :: ;erform tne ins:ecticn.
urther, the "0C Checklis for Snu ber Installation" cces no: s:ecify :ne drawing er CMC revisien. In 10 cf :ne 14 packages, CM,C revisions were issuec after tne ca e :na c:nstruction signed off n the MWOC that ne installa:icn was c: mole:e, and on or snortly before the date of tne CC inspection. Mcs: cf these CdC revisiens were not "information cnly" or clarifica-tien, cut :ertained to imaortant design features such as dimen-siens, material enanges, baseolate c:nfiguratien, e::. In :ne instance, :hree CMC revisions were issuec :e:Seen ::nstruc;icn
- meletion anc GC ins:ection. :n ene case, :ne IR nas signec aff incicating inscect cn ::moletec ne cay :efore ne *WCC w':n ne detailed Onecxlist 1. ems was signec. In :ne instance, ne center-to-cen ar cimension en a snuccer was enangec :n a CMC revisien to meet as-cuil 0:ncitions ', :u hac been 'C ins:ec-ted satisfac crily esica tefore.
The fol10 wing statements summari:e ne assessmen: Of di:e su:ccr / restrain; activities: II*-6
a y .
.. ._._ - s.-... . . . . . . - . _ .
4 q . . i.
. Numerous cases of CC accepted installed harcware not
conforming :: drawings and CMCs were identified by the NRC 1 Ia. CAT inspectors, ANI, and S&R CC during VCD inspections, and by TUSI "as-cuilt" :erscnnel, These c:ncitiens incicate ;ccr 'as:ecti:n worx, unclear / errenecus drafting, anc/cr unautnori:ec, uncontr0liec
. f.
alteratien of ccmcie:ac worx. Numerous instances exist wnere non:cnforming conditicns '
, , nave not been procerly icentified :: :revica ne input ::
the CA Corrective Action Fr gram for cetermining root causes anc preventing recurrence.
- c. Fr:m discussiens wi n si;e ersonnel anc -Me cavicusly
.large number of CMCs, ; , anc :ne actual casign/ analys1s was ;er Ormed a- er ccnstruction and ins:ecticn. This may have resul ec #r:m tne many enanges recuired cue :: reiccatec : icing, inter #ererces, anc :ne CMC Oregram itself.
- a. The accectability of -Me installec nar ware :: meet cesign recuirements based on a series of :ar-iai ins:ecticns (versus a final ccm:lete ins:ection after scri is ccmoleted) is ques:1:nacle basec en :ne f:l': wing Ocints:
. Numcers of " design" changes (CMCs)
. Scmewhat uns:ecific inspecticn :r cecures
. Amcunt of cngoing construction activ1 ies anc :ne accarent lacx of ciscicline egarcing ::nstructicn
;ersennel tamcering with CC ac:ectec nar: ware.
. Orafting and cesign ciscrecancies ec:ac in initial crawings, CMCs anc VCCs.
. The numcer of discrecancies notec cn succcr:s :revi-Cusly accected y CC.
. Inscection dccumentation not incicating ne *:esign" dccument ( rawing anc/cr CMC) revisden :na: was used for ne ins:ection.
- n c:nclusicn, al ncugn extensive ma;;r :ecanica :re lems ,ere not i:entified in :ne : ice su::ce:/ restrain; nar: ware, Ort =::
acti:n is recuirec :: accress ne a ove Or: gram c:ncerns. 3:ecifi 311y, a ten-icn mus: :e f cussed in :ne areas :# :ne ncne:nformance/ corrective action Or gram, comerenensive #inal ins:ections inc inscection cccumentation in creer Or;vice
- nficence in :ne acca;:acili y of installec pi:e su::cris/
restraints.
..,-7
.i.
.. .. ~
. - < . . . .s ><:.. =. - 2 .. ,
; . . . . u- .. .. . . .
.._ ., _. ,' [
s
; 3. Meenanical Ecuiement '
- n The followine samole of installed and CC accected mecnanical
..j equi: ment was selected and inspected for crecer iccation, 1 icentification, fcundation/su:;cr configuration and concitien, J in-place st: rage c ncitions and damage.
4 T3X-GHATGC-01, Wasta Gas Cecay Tank T3X-GHATGD-07, Wasta Gas Cecay Tank d TBX-GHATGD-10, Wasta Gas Cecay Tank
': TPX-TRAHLC-01, Letdcwn Chiller Heat Exchanger CP1-CDAPRM-01, Reac: r Makeuo Water Puma
.i CP2-COAPRM-01, React:r Makeue Water Puma CPX-00APRM-01, Reac cr Makeu: Watar ?um:
CPX-CSATBA-01, Boric Acid Tank CPX-CSAT3A-02, Boric Acid Tank 1 CPX-BRATRH-01, Ecric Acic Tank 2 TCX-CSAHLD-01, Le:dewn Heat Exchanger 7X-TRAHMH-01, Mccerating Hea: Exchanger [ Acceptance criteria for the fiele ins:ectices were taken frcm vencor drawings and technical manuals, and si:e structural /
- foundation drawings.
The Ocerations Travelers detailing the installa:icn and accac-tance cf Reac::r Makeup Water Tank-01, Mccerating Mea: Exchanger-01, 3cric Acid Tank-01, and Latecwn Hea: Exchanger-01 were examined. No preolems were icentified rela ive :: :ne lccation, identifi-
- cation, in-place s:Orage, or camage t: mecnanical ecuipment.
All fcundation nuts were installec and ac:earse :ign:. Mcwever, the makeuo cf ne foundation bol / nut asserolies was inconsis-tent, and in scme cases, the confermance ni:n ver.cor recuire-
- ments was in cuestion. For the tanks and heat exchangers
^
ins ected, scme doits had single nuts, scme nad :cuble nuts, anc scme hac a mixture of single and dcuble c:nfigurations. 'lence r .
- and si a founcation crawings incicatec single nuts. The Ocers-tions Travelers anc site structural bolting cr:cedures do not
- require or accress double nut:ing. The Westingncuse :acnnical manual for the accve heat excnangers ::ecifies that ne nuts on the slotted and of these units :e backec cff sligntly to allcw movement curing :nemal ex:ansion. Most Of :ne :o1:3 en :ne slicing encs of all of One neat excnangers ins:ec:ec were ::ucie nuttec anc accearec : te drawn ccwn -1gn . e slicing enc Of
- ne Latecwn Chiller Hea: Exenanger accearse : nave nac ;rcu between :ne succort bracket and the fcunca:icn :ac. ~he C: era-tiens Travelers cic not adcress :ne tecnnical manual recuire-ment; :ney statec only :na: the units were : :e installed 'n accorcance witn a olicable drawings.
IM-10 4
.. . .. - . - ~ ....u. ,- . . _ . . - . . . . . . . . . ~ . -
. . . g
/ -
4 Featino Ventilatien and Air Conditionina (HVACI
' ,. HVAC systems on Unit 1 and Cent.cn are essantially 100f. c molete and turned over to TUGCO. Many :ortiens are ocerating. Unit 2 ins:sila:icns are accrexima aiy 50%.c:mplete.
In general, HVAC duct anc equipment succcr:s are fabricatec anc installed by the Bahnsen Service Co. (BSC) in accordance wi:h tyoical drawings anc crocacure CFF-TUSI-003. 35C CC inscects exoansion anchor installation and :Me fitup anc welding of suct
. supco rts . After instal'lation, the succcr; as c' uilt configura-tien is sketched by draftsmen. This unsignec, unreviewed sketen is sent offsita :o Cercera a Censulting and Develecment Co.
(CCL) for seismic analysis. If analysis indicates :na: eccifi-cations are required, the sue:cr s are mccifiec, QC inspectac for new expansicn anchors anc weics, and the draftsmen crecare a final and formal ."as-built" drawing. This drawing is checkcc, and accreved and then reviewed by SSC Site Quali y Assuranca. This "as-built dis resucci- ad to CCL for final review and analysis. Because ne ty:ical drawings prcvide ne axial cracing, most cuc: suc:ce:s are mocifiec. Ncwhere in :nis crocass does BSC CC inspect :ne supacr:s for precer loca:ica, proper configuratien or memcer si:e and lengen. No CA/0C verification or aucit of One crafting decar ment's as-cuil: efforts is performed. Sign: duct supcort: and cne fan sup ce were ins:ectec in. One field fer.procer location, c:nfiguratien and c:nformance to drawing, design and precacural recuirements. Five cd Onese nine succorts did not conform Oc :ne c:nfigura:1cn anc :imensd:rs snewn on the as-buil: ske:cnes er drawings. cilcwing is a lis: of the su: ports inspectec anc :ne discrepancies no ec: Succort - Discrecancy -
'A-C3-854-1M aK Ncne A-C3-854-2N-1 AK Mone .
A-C3-854-2N-C5 3 dimens. (ciscrecancies :f 3t", 2 3/3", Ji" fr:m design) A-C3-854 4-2N-C12 1cne A-5Gi-852-1J-1C-01 1 cimen. (it' variance) SGi-852-1J-23 1 emper si:e sma~ier : nan snewn (3x2x3/3 vs 244xi) .
'Cuc: succor s ni:n " final" as-tuil: drawings have :ne 'A" : refix, :ne remaining :nree nad only :ne field sxet:n used for ini-ial seismic analysis a: ne time of :nis ins:ection.
III-11
. ... .. x . u. . .:. .._. _; ... . . . . . _ . . . ,
P l
}
. Sucecr Disc eeancy 1-C3-807-2N-A 1 dimen. (1")
.i
.. 03-607-2N-3 Ncne M
SXRFC-791202 (Fan Su:;or ) 2 dimens. (2 3/3", 2 3/2*) and Hil:1 loca:icn spacing (2 6 missing, 2 in error) d Oucting is ins:ected for prc er installation curing the system
- urnover walkcewn. No eneckiists are usec anc acce:tance cccumentation f:r as many as 38 segments nere ecservec en ene general *ns:ecticn Recor . 3SC perscnnei sta:ec :na: ceficien-
~ cies noted curing .:ne walkccwn were noted en a :unchlist, c:rrected anc reins:ec;ec. Mcwever, -nese ;uncalists are nct a centrollec or a retained document, and 0C reinspec fon and c1csecut are no accressec by site ;r:cecures. SSC persennel s c:ulo not pr:vice an examole of a c:mciatec/closec cu: ;uncn-list.
Turnover package 3601, in the final stages of c:mele:icn, was examined in tne 3SC office. This package c:ntainec 25
, Ins:ecticn Rescr:s, ac:epting 195 cuct segments, wnich were all signed by one QC ins:ect:r en one day. The ins:ec:fcn activities :: verify recer makeuc of 195 joints and ap roximately 200 su:;or: locatiens wculd take several weeks and
,. require c:ntrols over tne precess to assure that all recuired inspecticns are :erf:rmed. 3SC pr:cecures 20 ne: provide c
instructions :: CC ;ersonnel en hcw :: ::cumen:/c:ntr:1 :ne in-pr cess walkd:wn ac:fvities and 3SC ;ersennel c:ul :r: vide
, no evidence :nat the individual inspections incluced in :nis
;ackage nere rec rcac cn log tecks, markec u: :rawings, or
. similar c:ntrol =acnanisms.
The following cuct segments anc in-line ecui: men; were inspected in :ne field for ::nformance :: crawing anc ;r:cecural require-ments: CRKE-1, EMO-7, Segments 6 :nrcugn 12
; Fan CPX-VAFNID-01 an CP1-VAFNIO-07 Fan 071-VAFNID-10 lamcer CPX-VACPMV-05
!sola:1cn Valve CF2-VAD :BC-05 Accroximately 25 cuc segments acjacen- : ne sucecr:: ins:ec-
- ac were als: e.tamined.
4 D e III-12 g e ema e
Five of eignt joints in :ne CRKE-1 EMC-7 Ifne hac cne or more l lecse colts. Numerous anner joints observed curing the sucpert l ins:ection also hac icose er missing fasteners anc missing lockwasners(forexamole: f:ur icose bol s on :ne joint acf a-cent :o su;cor: A-5G1-852-10-10-C;). The flange :al:s for centainment isolation cam er 072-VACPSC-08 anc a similar camcer cn Uni: I did no: precerly fi the holes anc were no: fully , installed. Cn camcer CPX-VACFMV-C5, :ne gaske en ene sice .vas partially missing anc the c:ner sice had no gaske: at all, bolts I were Icose and icckwasners nere missing. The :olting require-ment of SSC procecure CE?-TUSI-CC8 :na recuires acciticnal
- rner bolting en duct accessories was not met en :ne cameer on na cisenarge sice of Fan !?!-VAF11C-07.
The BSC crogram :: ins:ec: anc cecumen: the installaticn of HILT! Concrete expansion ancncrs, hcwever, accears to nave been
- ncr ugn anc effective.
In summary, One numcer and variety of discrecancias retec Oy One NRC CAT inspec :rs between instailec harcware c:nciti:ns anc crawir.g/ Orececure requirements indicates that tne GC ins ection recuire.eents anc centrols have not :een sufficiently cetailec in Oracecures, nor nas the CC inscector performance in the field been acecuate Oc assure
- na- HVAC systads are installec as recuired. There was an accarent f ailure of Me SSC Corpora:e, 3rcwn anc Rect anc TUGCC QA Organi:a-tiens to icentify and correct ne ::servec pr: gram deficiencies curing neir aucit/ surveillance activities.
f III-13
IV. WEI.0ING. NCNDESTRUCTIVE EXAMINATION
', A. Cbfective Cetermine cy direct coservatien and incacencan evaluation of work ; performance, wort in progress, anc ccmcie:ac work, wne:ner f.ield welcing activities associated 41:n cicing, nangers/succerts, steel structures, anc neating, ventila:icn anc air ccndi:icning (HVAC) i systems, are centrolled anc cerformed in accordance wi-h NRC require- ; ments, 5AR ccmmitments, and a;;1tcacle ceces anc scecifica:icns.
In addition, detarmine by direc caservation and review of recorcs na: welders and ncncestruc-tve examination (N02) ;ersennel are acecuately trainec and qualifiec in accorcance wi a estaclisnec cerformanca stancards anc acclicacie ccde recuirements.
.f
- 3. Discussien
- 1. Pice Succor:s/Hancers Twenty-seven pice succor s were selected for inspection of welding. Ali of the hanger welds had been previcusly cuality centrol (QC) inspected anc accap ac by 5ccwn & Rect (3&R). A listing of the hangers felicws:
P1pe uni: 45ME
, Tyce Si e. in. No. System" Class Hancer '!c.
Ecx 1.5 1 5! 2 S:-1-53-235-CC6-2 Stru: 5.0 1 CVC 2" C5-1-357-CGi-522R Box 12.0 2 S* 2 SI-2-03'. 125-Y32R S:ru: 3.0 2 C5 2 CS-2-C63 113-5 22 Stru: 3.0 2 RC 2 RC-2-121 401-552R Anchor 3.0 1 5* 2 5;-1-037-CC5-332A 7 Stru: 15.0 1 C5 2 CT-1-002 -CCS -532R Strut 12.0 1 CC 2 CC-1-116-C06 ~32R Ecx 3.0 1 CVC 1 C5-1-063-CCS-522R
! Scring 12.0 1 SI 2 S I-1-031 -046-Y325 Stru t 12.0 1 RH 2 RH-1 -CC A-007 -532R ^,
Strut 12.0 1 SI 2 S I-1-031 -024-Y32R Strut 8.0 1 CVC 2 C5-1 -C63 -037 -A42R Ecx 3.0 1 CC 3 CC-1-136 7C6-E53R Stru: 2.0 2 CC 2 C C-2 18-C23-CC2 -3 Stru: 3.0 2 RH 2 CS-2-0*.2 403-Ca2R Stru: 12.0 1 CC 3*** CC-1-1*.5-037 -743A Snucter 13.0 1 FW I FW-1-017 -702-C52X Snuceer 3.0 1 CV 2 C5-1-C01-C'.5-CaEK Snuceer 5.0 1 5: 1 3I-1-092-CC5-Ca' A . Spring 12.0 1 AH 1 AH-1-001-001-Ca15
- V-1
* , - - --e
cr .. m f- -. .o ~,. . .. . - . . -. . _. ., - . - _ _ , _ . _ t - . . 6
- s. :,
J
.I .1 d F1pe uni: A5ME 3 Tyce Size in. No. System" Class Hancer No.
Spring 4.0 1 RC 1 AC-1-075-025-C51R 5 Spring 4.0 1 CV 2 C5-1-055-010-5425 Stru: 13.0 1 FW 1 FW-1-017-001-552X Stru: 2.0 1 RC 1 RC-1-015-7C8-C11R Stru: 12.0 1 RH 2 RH-1-003-002-542R j Sox' 1.0 1 CVC' 2 C5-1-53-059-C03-2 d "5I= sare:y injec: Ton; C40 = cnemicai anc volume con:rci; 1 CS = c:ntainmen: spray; RC = react:r :: alan:; RHR = residual
-t heat removal; FW = feecwater j **Undersi:e Weld
***0verlap, Imcrecer Centeur 4
The majority of hangers ins:ected u-ilf:ec '4'ie welcs for joining nanger c:mpenen:s and ce:siis. I Sao Of 27 hangers exnibitec unaccepta:ie wel:s. Manger CS - 357-001-522R exnibitec one uncersi:e filie: welc. The welc was 1/16-in to 1/8-in. under the sce:ified recuirement. Hanger CC-1-115-037-F43A exhibitec everlac and imcr:cer welc con::ur. Thirteen of 27 hangers had been inspected vencer certified drawings (VCD).
,f-ihe welds, for the most cart, exhibited gcad workmansnic. Some of the hangers inspected were painted anc NRC CAT ins:ecticn of 3
- nese hangers was mainly for weld si:e.
- 2. Safetv Related Sue:cr s/Mancers f e 5'ectrical Cencu4:
. anc instrumenta:1cn 4
-1. Electrical Cendui Sucecris/uancers 4 These concuit supports / hangers are fa:ricatec On-si a bc n
~
in :ne shco and in the field. The follcwing concui sup orts / hangers were being fabricated in tne on-site snco. The NRC CAT inspection was performed en the sucject nangers
,f before the hangers were painted.
C03G09150-1 C12GC6431-25 CClG2'035-1 C03G09150-2 C12G10 52-3 C12G2' ' ~ a-2 C03G09150-3 C12G1C652 A C23 HIC 674-5 C12021194 4 012G10455-23
!7-2
-. n-.y y ~ . . __
.., +-. . - - - - . - ,, n,. . . - - . _ - - ,., . -. - _ . . - - , . --w, ~ - - . ~ . . - , . - .
~ . . - - - . .
.=. ..._ _. ,___ .. _ _ ,_ _ , _
a All welds en tne acove suppor:s/ hangers were visually inscected by ne NRC CAT. inscectors for acceptance :o 01-0P-11.10-4 which invcke AWS 01.1, " Structural Welding C0ce." The NRC CAT ins:ectors axamined -he en-site shco facilities, anc reviewed faorica:icn procecures, ins ection records, anc in-process activities relating to material identification, marking, cutting, cisaning, and welding. The NRC CAT inspectors caserved ne arc-stud welcing - process being requalified. Requalifica:icn was ;erformec for 1/4-in. , 3/3-in. , anc 1/2-in. studs. Two stucs for each of the indicated si:es were welced for recualifi-
- a icn. 'delcing was cerformec in ac::rdance wi n WS 31.1.
The weldec stucs passac :ne :end est in acc rcance wi n AWS 01.1, and therefore ene welcing process was jucgec acceptable. The NRC CAT ins;ec Ors examined fiele installec anc CC ac:ectec sue;cr:s/ hangers. The instected hangers =ere constructec for :ne mos :ar cf 5-in. x 5-in. scuare ucing. The succor:s/nangers were relatively c melex as evidenced by tne numerous seconcary sup: orts attacnec to the primary structure. The su:jec: su: orts / hangers inscected (listec belcw) were ;artially facrica ec in
- n-site shcps and c moleted during fiele installation:
C03G13046-7 C22XC6357-1 C*.2X15115-1 C03G18046-8 C22XC5902-1 002C30179-1 C03G13046-9 C12012705-5 CO2030179-2 C03G13045-10 C120127C5-7 5H-IN-C:-M-5b C* 2X06108-2 ' C12012705-l' SH-IN-CMM-5c C13G13999-2 C120150 7-7 5H- N-CMM-5e C33G13046-11 C11GC4797-1 5H *N-C:-M-5f SH-IN-CEM-5g The identified hanger / sue:cr welcs :nat were inscected (botn snoo faoricatec anc fielc installed) were visually ac actable. Review of the snoo activities revealec :na: One work was being c:ncuctec wi-h crecer controls and ne
;uality of ne workmanshic accearec ::nsisten: ai n gecc incustry :ractice,
- b. Instrumentation Marcers/Succcr s The cajority Of instrumentation anc ::ntrol ( *1C) 1:n 45M5 (safety-rela ed) nangers are fa:ricatec in sn::s en-site.
5&R field ::nstructicn installs them :y :ol-ing or aeicing ,
- ne hanger assemoly Oc :uilding structures. So n '*eic arc saco inscecticns were :eing :er#0rtec cy the licensee. )
l I IV-3 4"
. :.a. e > , . . . . , .
. .. . = . _ _ . ._
4 .e-1 s she NRC CAT inspectors observed snc; fabrica-icn of six . hangers at various stages of manufacturing. In-process
; activities reviewed were material traceability and identi-fica:fon marking, a:plicatien :u :ing anc cleaning of materials, welding, filler material control, and hanger
- identification markings.
. Review of documenta:icn for tre she: facricated hangers
; incluced review of material rec:rcs, engineering crawings
; for configuration, inspecticn rec:rds for welding, material u verification, visual inspection of welcs, ins ection re;cr numcer, and verification of surface preparation for painting. Occumentation indicatec ins:ecticns were perf rmec ac::rcing to CC documen: 0 * -QP-11.3-2.
e Ten field installec and QC ac:ected hancers were aisc
. reviewed by One NRC CAT ins:ec:ces. The ins:ected hangers are icentified by ne felicwir.g num:ers:
S-la67 S-2729 S-1806 5-2723 5-2211 5-2724 5-2250 5-2726 S.-2292 5-27 2 The identified field installed hangers were ins ected f:r visual weld ac:estance in ace:rcance wi-n AWS G1.1. anc nc deficiencies were identified.
- 3. Heatine Ventilatien, and Air Condi-icnine (HVAC)
The NRC CAT ins:ec:crs performed field ins:ection of HVAC welcs en hangers, cucting accesscries such as ficw controi cevices, air da cers, and c:ntainment istlation valves. All cf :he i ems inscected by ne NRC CAT inspectors hac been previcusly ac:actec by Bahnscn CA/QC.
- a. HVAC Sucecr:s/Hancers Strue: ural steel nangers su::cr ing :ucting runs were visually ins:ectec to cetermine if welcing met :ne -ecu re-men s of AWS 01.1. Typically qangers are fiele facrica ac fece a-in. x a-in. x 1/2-in. structural steel angle iren.
Mangers ins;ectec anc resui s are as f licws: IV J
- he* m .M+ m- up . .. . .
t No. of No. of Welds Unaccactable No. Hancer No. Ins:ected 'd el e s 1 C3-807-2M- A 21 11 2 C3-807-2N-B 21 10 3 SGI-852-1J-15 .12 11 1 SGI-852-1J-1C-01 12 3 5 SGI-852-1J-ZS 12 10 5 SGI-852-1J-15A 13 5 7 OG-84A-2K-1AR 11 2 3 CG-844-2K-1AT 11 5 9 ' 0G-844-2K-1 AQ '.1 2 10 OG-aa4-2K-1AR 11 3 11 OG-aa4-2X-130 5 3
- 12 DG-844-2K-13C 5 3 13 RS-1A-C1 3 2 la RS-1A-C2 3 1 TOTAL 10 73 .
Thus, a rejecticn ra a Of accr0ximately 15% was exnibitec for welds en structural s eel. hangers. The majcrity Of tncse nelds that were unaccectable were rejected because of uncersi:e weles. The si:e of nelc s:ecifiec on mes :f :ne drawings was 1/2-in. These aelds usually measurec accr xi-mataly 3/8-in. ine res or ne welds aere unaccectacle' because of insuffician leng:n, undercu:, anc im;rcter aele c:ntour. Bannsen QA/QC personnel ccservec anc acxnewiecgec mos: Of
- ne acave welc ceficiencies. It was recortec after :ne 1RC CAT inspection :P.at en ecruary 23, 1983 :ne licensee's auci: of One ins:alled HVAC system confirmec :ne ceficien.-
cies and a 10 CFR 50.55(e) recer: was filed by the licensee.
~
- c. Ccntainment Isolation Damcer The Unit i HVAC c:ntainment isolation damcer cennection flange cetailec on Bannsen sxe::n, !50-001, snews a la in.
circular flange mace fr:m 2-in. ( 3-in. ( i-in, angle ic:n. Ins:ection Of :nis flange in :ne fiele -evealec na- '- is actually faoricatec frem rac cieces of 'lat s :cx joinec Oy fillet welding :: form a rign; angle #1ange. This is c:ntrary :ne sxe:cn. The crawing is also in err r in
- na: 1- fails :c s:ecify a fillet weic joining ne rac
- ieces of material. Furtnermore, One 'ilie: weic 2:en ins:ection was f und :c :e unac:ac:acle in 'a num:er of areas.
~
'V-5
. . . . . . . . -., . h ... .u . .. a . a :a . . =w . .. n . _ . _ ... .. .
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- 1 t.
8 4 Ei : 1 -i' Secause of limited access to the full circumference of the 4 welc, only a 130-cegree segment was inspected. Apprcxi- 'Al mately 40% of the 180-degree segment of the weld was ' visually acceptable in accercance with AWS 01.1. The res: of tne weld exnibited areas of uncersi:e weiding, linear f surface indications, arc strikes, under ut, everlac, and in (! general, a pcce level of workmanshic. d s c. MVAC Welder and crecedure Cualificatiens 1
- 3ahnson procedure GCI-CPSES-CC9, s:ecifies :nat welcers i shall be cualified to ASME Section IX. Gibbs and Hill 4' specifica icn 2323-MS-35 (Bahnsen c:ntract recuirements) s ecifies tnat welding precedures anc welcers may be cualiffed in accordanca with AWS 01.1 cr with ASME Section IX. Actual field facrication is ;erformec :: the recuire-ments of AWS C1.1, Structural Welcing C ce, and AWS C1g.0,
" Welding Zinc-Ccated Steel".
[- The NRC CAT inspec: Ors reviewec welcing ;r:cacure s ecifications and welcer performance ualificaticns. This
-[ . review revealed ne follcwing discrecancies.
J J WPS No. Precess Cc=ments a <3 BSC-12.1 GMAW AWS 01.1, 5.5.2.3 specifies that ne
? cover gas ficw ra a be c255 cf :ne ra a qualified; hcwever, WPS BSC-12.1 f ails d
to specify One ac Ual rate usec curing cualificaticn. . AWS'01.1, 5.5.2.3 s:scifies :1C% fer
. amcerage, :75 f:r v01: age, bu: 3SC-12.1 . fails to s ecify a range; s:ecific values are given for amos anc volts, . but no tolerance is ;ermitted :y WPS V
SSC-12.1. g ASC-15 GMAW The weld crocedure cualification shee BSC-14' for 35C-15 scecifies unicue values for SSC-la.1' amcs vol:s, travel s:eec, and gas ficw. The WPS 3SC-15 s:ecifies 30 2C f 3/hr gas ficw, 50-170 am:s,12-25 vol s, 5-12 in./ min ravel. The ranges are cu: side of :ne ::lerances ;ermi :ec :y AWS 01.1. s SSC-13 3 MAW WPS BSC-13 s:ecifies ranges for amcerage, voltage anc travel. No weid procedure cualifica:icn rec rc aas avaiiacle for WPS 3SC-12 na states wna: values were usec :uring ne ::a:1- ! ficatien test. I tv-4 1 l 1 s . w w e w-w~*- m _ ,c .-e--a- --e- -- - -__-- -- - - - - - -
....y-....+ .. . . . . . . . .
EL .- t- ,, SSC-20 SMAW 95 ames is indicated on Welding Procedure Qualific. tion Record Sheet for this procedure. WPS SSC-20 speci-fies a range of 70-150 amo. This range exceeds the :25" tolerance ;ermitted y AWS C1.1. (71-119).
'
- Same c mments acoly :: tasse crecedures as f:r SSC-15 3annson s:ecification 2323-MS-35 paragracn 2.14b Welding,
, (r.odified by Design Change Authori:ation 9898) s:ecifies that all welding snali be in accercance wi:n AWS 019.0-72.
Paragracn 2.14(a), Of scecifica:icn 2323-MS-35 specifies that welding crecacures anc welcers snali :e ;ualified in acc:rcance wita AWS C1.1 Secticn 5 c_r, ASME Se::icn IX. All of 3ahnscn welding crecedures have been cualified in ac: rdance witn ASME Section *X. ;ualification in ac:Or-dance witn ASME Secti:n IX is ;ermi tec Or: vicing :he recuirements of Section IX meet or exceec all of the recuirements of ah5 C1.'.. as actec accve, we:c ;r:cecures 35C-13, 3SC-14., BSC-14.'., 35C-15, anc 5SC-20 fail to meet all tne recuirements of ah! 01.;, and are not :ualified crececures. The balance of Bannsen's welcing procacures were not reviewed cy ne NRC CAT inscectors anc their status of qualifica:fcn, nas net determinee.
- d. HVAC Weldinc Occumentation A review of Sannson :uaii y centrol rec:rcs anc : cacures reveals a failure to identify wna aele :r:cacure ,as '
emoloyec for a scecific hanger aeid. The welcing ra:erial issue slio, used :o c:n:rol :ne issue cf welcing c cs, nas a space to enter ne HVAC system icentification. Mcwever,
, as determined by review of welcing material issue sif:s, the seace Orovided f r system icentifica:icn Only incicates the plant elevation, cuilcing, and reacter uni: numcer.
Also, traceability of the welding material utili ec and the welders emoloyed on specific welcs cannot be determinec. Sannson field inspection recor:s, althcugn provicing for accectance of welcing, fail to icentify the welcer, the welding precacure, and :ne welcing materials Gsec. a l'l-7
u :.;a l . . . . , . ._ _ .- . _. -: : * - " 'z:,w ;. q;_., . . _ _. _ , ._ , . 3 * . . 3< . s J] e. HVAC CC Inscector's cualifigation 3 HVAC CC inspecter qualifications, both past and cresent, M were reviewed. Training records, excerience, and ecuca-
. ticnal requirements were coccarea wita tne requirements of
.i - ANSI 45.2.5 and Regulatory Guide 1.53. On the basis of i this review, the supporting cccumentaticn scpears Oc mee the minimum recuirements of ANSI 45.2.5; hcwever, CC training accears to have been based on review of QC crece-
- du res . Li::le evicence was available tha reflected -
_ technical training of substance. The QC inspec:crs infermed the NRC CAT inscec:ces :na: up
- o 1 year ago 3SC inspec:crs were no: using fillet gauges, er other similar :cols for measuring weld si:es. Furtner- .
more, curing the NRC review scme inscec:crs exhibited
~; limited kncwlecge of welcing inscection. CC inspector cualification records accear oc mee: the letter of the ANSI re:uiremen tu: it was accarent tha: :na CC inscac:ces lacked proficiency in the inspection of weics.
v.
.; 4 Visual Examinatien of ?icine Welds Seven piping iscmetrics were preselected for visual inscecticn
. of piping welds. Nine accitional picing iscmetrics were selecteq at randem for visual inscectien of welds. All of One
, piping runs had caen inspected and accected by 3&R CC before NRC CAT inspection. The inspected piping runs anc etner system carameters are as follcws: -
?1clnc damcie Gian Pice F:. Mc.
' Unit Isemetric dia. of ASME of
. No. System" No. (in.1 Pice Class Valves 'sel cs -
- t CS CT-1-SB-023 2.0 55 2 2 25 1 CS CT-1-55-027 2.0 50 2 2 29 .
. 1 SI SI-1-SB-004 5.0 35 1&2 2 12 1 1 SI SI-1-AB-002 1.0 130 2 2 12 2 SI SI-2-AB-001 4.0 35 2 C 12 1 AFW AF-1-53-006 10.0 52 2 2 1*.
1 AFW AF-1-33-025 2.0 105 3 2 .. , TOTAL WELOS .13
+
i'I-8 _.. .. - . . . - . . . , . , - . . , . . . . . . . ~ . - . . . - . .
t.., . > - *
- Ranccmiy delectec ?'etng Runs
'Jni: Nc. Systam" Isemetric No. No. cf Weids
! SI S!-1-RS-037 1 1 SI SI-1-SB-042 11 1 SI SI-1-RS-CS3 1 i SI SI-1-RS-052 1 i SI SI-1-RS-033 5 L SI SI-t-RS-043 7 i CS CT-l-RS-017 3 i . CS CT-1-RS-019 3 1 RHR RH-t-53-C03 _i, TOTAL WELDS 39
'C5 = c:ntatemen; spray; si = safe:y in;ec :n; AFa = aux 1, ary feecwatar; RHR = residual her: remcval The welds inscected (identified by 16 picing isemetrics) are ASME III Class 1, 2, and 3 weies. The size of :ne of:ing rangee fecm 2.0. to 24.0 in, in ciamatar; accroximately 14C0 ft of picing was involved in tne samcle. A tctal of 152 pioing welds <
were visually inspectac in ac:Orcance wi n ASME I!I-71 f:r undercut, everlap, lack of fusion, sur' ace :crosi y, acc c:her ancmalies relatac to surfaca ::ndi-icns. Sctn vencer anc fielc fabricated welds were part of :ne samcie. Three of :he 152 welds exhibt:ad encesfr:bia sur'aca c:nciticns: Wald F-1;S (reference SI-1-SB-CCA, safety injection sys am) exnibi:ac insuffician weie in : hat :ne face of :ne nele was telcw the adjacent pipe surface. This . veld was also imor:cerly locatad en -he isometric drawing (ISO) roviced. The ISO for welc F-113 scecified that :ne weid sncuic nave been !cca:ac a 1-given positien fr:m a wall. Actual field lccatien aas en :ne ecocsita side of the wall, apercximately 3-ft fr m :ne icca:icn snewn on :he 150. A ncncenformance recer: (MCR) was :rs:arec :y the licensee for both :ne surface c:ndi icn of the weld anc for misiccation. One weld identified on ISO RH-1-55-003, Rev. 9, jcins a section I of 12.0 in pice to a section of 5.0-in rice. The transi:1cn, recucing elbcw was vencer fabrica:ac. The wele jcining :ne 5.0-in. ci:e sections exnibits a :acer er sic:e Of a::c x'mately 4: 1. The slece :ermittad oy ASME III is 3:1 maximum. This concition was discussac wi:n S&R GC ;ersonnel, and a: : e ene of - the NRC CAT inscaction, :he ma::er das still uncer review. I'l-9
. . 3
- e. . . . . - . -.no. .
. - +; - ---- . , . .
A weld-o-let (vendor installed) identified on ISO CT-1-RB-019, Rev.1 (adjacent :: field weld FW 4) has insufficient weld reinforcemen on the weld joining the weld-o-let to the cipe. The suoject wel: exhibits less tnan full weld reinforcament. l This condition is readily accarent by noting that portions of the welc prepara:icn en tne weld-o-le: side are c:servacle.' A review of tuality con:rol crececures anc engineering anc c:n-
. structicn specifications for vender anc field installec weld-o-lets cr sock-c-lets reveals a lack of criteria for specifying
- ne required si:e of weld reinforcement f;r les's than full re-inforcement aelds. Tyoically, aeld-o-lets can be installed en several si:es er diameters of :i:icg acc wall thi:knesses. On tne basis of the engineering requiremen s (temcerature and pressure) for a given system, the si:e Of the weld reinf:rcement needs :: be specified for incse cases wnere the weld is less'
- han full reinforcement.
- 3. Review of Radicceschs 4
Radiograchs for a ctal of 81 E&R welds, 28 ft of Chicago
, Bridge & Iron (C2&I) containment liner olate welds, nine Scutnwest Welding Co. welds and ten ITT Grinnell weles, involv-ing 1254 film were reviewed for c:mpliance :: aoplicaole
. recuirements. All weld radicgracns nac been :revicusly reviewec and accected my tne licensee or nis au:ncri:ec representa-ive.
. One hundred twenty-five additicnal film we're also reviewec in
- he radiogracnic in:erpretation reca for tne Our:cse Of evalua-ting :nree interpreters' acility :s felicw 3&R pr:cedures and their ability :: procerly inter: ret radicgrapnic film.
Canditiens were disclosed by :he NRC CAT ins:ect:rs in six aeids l
; that recuire attention and c:rrection (See Taole [7-1). These conditions were discussed wi n :ne licensee recresents:1ve.
These c:ndi:1cns are summari:ed belew: j a. 3rewn and Rcot weld radiegracnic film 42a555-C5-1-R5-20 revealec an indication of insufficient fusion (!?) anich was rejected en -he original film by :ne S&R intar:reter bu: was not identified en a suoscuent recair shot. This deficiency was identifiec by ne NRC CAT ins:ector. B r:wn anc Rect cre:ared a ncnc:nformance rescr: (1CR) anc :ne new
. Paciegracn :Ma resulted fr:m :nis NCR c:nfirmec :ne lF.
The new radicgraon also revealec :na: :ne angle of :ne repair snc: nad :een takan 130' fr:m :ne criginal snet,
! Anotner uncers1:e welc aas f:und (no: ;ar of ci:e sam:le) on a 2-in.
ciameter sock-o-let en sicing Iccatec in CC Oumo e em 01. This secx-c-le: is welded :: a 3-in, line snewn en ISO CC-2-AB-015. 1 IV-10
- m. , _ . _ . . . . . _ . . . . . . . . . . .
/
7 3 thus changing the accearance of the IF cn the film and 3 reducing the likenced of its being identiftad.
- b. The NRC CAT ins:ect:r review cf 3&R radicgracnic film d4162-MS-1-R3-081 revealed an area :n :ne film, tha: c:in-cided with tne pice weld :cne, wnich was more cense (thinner matertal) than that part of the film sncwing :ne
. carent material. 3&R precared an NCR and latar c:nfirmed a
- hin wall section by taking an ultrasenic nickness reading. The area was accroximately 0.014-in. under minimum wall thickness. This radiogra:n had been accettac by S&R QC. S&R statac that weld repairs to this area would be mace,
- c. S&R film *29020-FW-RB-020 revealec an area of IF and aligned enlongated porcsity in ex: ass of *.-in. :ha: had been ac:sotad by 312. An NCR was precared and new radiogracns were taken. The new radiogracns revealec :ne same c:ndition anc the welc was rejectad. The licensae recresentative statad this welc sill :e recaired.
- . S&R film v28669-CC-AB-013 reveaiec an indication of !? :na was not marked by tne inter:re ar (:ne same inter;rstar involved in film 29030). This indication was hcwever,
! icantified by another S&R inter;rt:ar wnen a su secuent i
recair snot was evaluated. The NRC CAT inscect:r's review of One origina'l film for the su: ject weld revealec incica-tiens of incomoleta fusion similiar :o ne recair welc radicgraph. 3oth film evaulations were mace withf 9 a snce period of time. This ceficiency in f 41.7 in arertta:icn was ' referre~d to the licensee reertsenta-ive for folicwu: inc corrective action,
- e. Curing the NRC CAT ins:ectors review of radicgracny fils
- (20619-C3-1RS-023, 29013-CS-2 13-091)
- wo weics were cbserved to have densities :na are belcw minimum in :ne area of interes er are cutsice one -300 - 135 f:r senetra-meter recuirements in :ne ASME Cece and 3&R precadure. As i
a felicwup to tne above, it was learnec that 3&R had changed tne three-view RT taennique to a fcur-view taen-nique apcroximately :wo years ago to c:rrtet 1:w-film density prcblems. S&R is new in the recass of enviewing radiogracns of weles sno: tefore :na: time ui n :nrte-view technicua. The NRC CAT inscector was inferred cy ne licenseee :na: any raciegracns *:unc na: :: rc: Teet ::ca recuirements wili :e re-raciegracned.
- f. Ini-tal review of accreximately 10-f of raci:gra:ni: 'M taken Of c:ntainment linear aeic saams revealec linear indica:1cns. Sucsecuen: visual examina:icn :# ne liner
, plata, in :ne area of in arest, revealec na: :ne liner
- lata c:ntained grinding marts adjacent o ne nele. Cn
, the : asis of a cor ela:1cn of One ;rinding marts :n :ne i
- IV-11 i
i I . - . . . . . . . . _
. - - . . .~ __ _ - , . , _ - _ _ . . - _ _ _ . . _ _ _ _ _ - ~ _ _ _ _ _ _ _ . . . - . - - . , , . , . ~ .
7
.. . .c:... ; h; . =.,',,. .
liner with marks en the radicgragnic film, the NRC CAT inspector agrees that tne welcs are accactable. s Mc ceficiencies were identified in the review of Scuthwes: Welcing anc !TT Grinnell sample of radiogracns.
- 6. Review of NOE Prceecures. Practices, and Persennel Oualificatiens
- a. Review of 3rewn and Rect NOE Prececures A total of 19 precadures were Feviewed by the NRC CAT inspec:cr.. It was cbserved that mes; of :ne pr:cecures had been revised within the :receecing five mentns. Wi:n :ne exceo 1cn of QI-CAP-2.1-1 and QI-CAP '.0.2-3, :ne :r:cacures appear to be acequate and in good orcer. C:ecents on :nese precedures follew:
(1) OI-0AP-2.1-1. Ncndestructive Examination Parsennel carcifica:1cn This ;rececure cces nc: :etail s:ectfic recuiremants for work experience anc new :ne NCE Level III maxes
.. his determination for cualification before
. certification as required by SNT-7C-!A, *.375 editien.
(2) Ot-cap-10.2-3, Radicert nic E.ta.mina:'en e 2 This paragraon allcws tne use of Ir 192 en steel as thin as 0.125 withcut ex: lana:1cn. Review cf ree:rcs disclosed that this cr:cecure recuiremen; was cased en a qualification film using two flat pieces of ma:erial in :ne center of tne film. This set of c:cci:icns is not representative of :ne geccetric c:nciti:ns -hat wculd te enc:untered en an actual raciegracn of :i:ing welds. 'Tne qualificatien film revealec a marginal AT sensitivity.
- b. Certification Records Records of $8 sacarate certificatiens involving 22 :ersons were reviewed. There was one certified ul;rascnic ec:rd for a Level II tnat c:ntained no evidence of ex:erience in angle beam, shear, wave examination. A subsecuen: in er-view witn :nis :erson and a signec c:cumen: ::nfd ro na: ne hac no exterience 41:n snear wave ecnnteues. The :erson s:sted tha :nis inf:rmation was cace '<newn c :ne '.avel I!! tef:re ne was cartified. The recuirements f 1NST-70 '.A dere not met in tre :ssa of :nis indivicual. The NRC OA7 ins:ect:r was informec by tne licensee na: :nis :erscn nac no: :erformed any ul:rasonic worx f:e l&R.
I'l-12
]*.. . . . . . - . .. . . . .
- c. Interviews
. Taenty :ersons were interviewed for cetermining :neir uncerstancing of precedures, and tneir ability Oc cerform i the ccerations f:r which they were certified. Eacn ter:cn inter /iewec accearea to have a goed attitude cwarc manage- - ment anc procacures. They all had a current copy of .1 procacures anc were very c;en anc willing to discuss their ; assignments.
- d. Work in Precress (Methcc Cemonstration)
Observations of 21 se:arata noncestructive examina:icns aere performed. Of :ne 21, nine were radiogra:nic, :ne
. ultrasonic, and eleven licuid penetrant. Of :ne nine demcnstrations in radicgrasny, two aere fiele setups, :nree wers for welders' qualification, one for darkroca rece- ; curts, and three for in-crocess film viewir.g.
- . (1) Radiceracnv i In all cases, :ne personnel cemenstratec a ;ccc understanding of hew to use :ne ecui: ment, and new to
! perform :se nondestructive examina:icas.
(2) Ultrasenic
, - One person certified at Level I! was asked :s c:ndue:
a calibration procecure for 3/4-in. plate. The calibration was performed satisfac:crily. (3) Licuic Penetrant i All persennel :erformed in a ta:tsfac:ory manner
, except in :neir uncerstanding for c:ntamina: ten of :ne penetrant acclicatien by crusn. ~his is nc:
censicerec a mador cr:blem. i e. NCE Ecuitment Calibration and Material Verification 4 Over 47 items of ecuipment and matarials were enecked for calibration, certifica:1cn and c:moliance wi:n estaclisnec recuirements . They inclucac: 3 ultrasonic instruments a const;; meters 15 cassetts 3 survey instruments a censity stri:s 3 *rl92 cecay enaris The items ' reviewed satisfiec :ne s;ecifdca:icn anc r:cacura recuirements. I j tv-13
.m,, - - , , , - - . , , - - - - - - , ,n .- . - - -- ~ - ~ - :.,-.-,-,a,. . - - - - , , - . - , - . . - , , , w --,..r
. - . . . . . . . . . . . - . . - . z . . ; .. ..
.L. __ _ __
- 7. Review of Welder Cualificaticn From the sampie piping runs inspected, welder identifica:icn num:ers noted en cicing welds were selectec and tne welder cerformance cualificaticns rec:rcs reviewec fcr ccmcitance ::
WES-031, "Scecifica:ica for :ne Qualifica:icn of Welders and
- Cce rators" .
The welders whose cualificatiens were reviewed are identified below by neir assigned symocis: Art -nu his AWD AUA SII AHX APL AHX AHF AFO ARP AGM 3C0 3GU ABR A review cf 'aES-015 " Schedule of S:ancard es: Wei:e-
, Qualification Matrix and Welde- Ferformance Qualifica:1cn Leg" was c:nducted. WES-015 scecifies :ne varicus weld tests for cualifica:icn and recualifica:icn cf weicers. A review of :ne welder cualificatien matrix, listing :ne cr:ss-matrix cf the welding prececures applicable to each welder, was also c:ncuc ec Oc ensure tha: ne essential variables for weicer cualificatien in accorcance witn ASME Section IX, were not viciatec.
No deficiencies were casertec in ne weicer cualificaticns reviewec.
- 3. Review of 3&R Welcer Cualifica:icn rcerim B&R has instituted a program wneracy :ne weld engineering uni:
utili:es .elcing tecnnicians in :ne field.: Oversee anc assis-welcers in the c:ncuct of :netr werx. Acciticnally, tests #:r qualification, (both :ne types of tests accinisterec :c newly hired welders and to welders being ucgraced) exceed :ne recuire-ments of ASME Sec:1cn IX. Furthermers, 3&R nas imclementec a program to qualify welders by radiogra:nic testing ratner : nan meenanical testing cer ASME See:1cn IX. Section .X :f the ASME Ccce :ernits ei:ner radiegra:nic er recnanical :asting. Raciegra:nic :ssting normally is mere :i'dicui: : ass, :nus recuiring a signer skill level. No :iscrecancies 4ere ::servec curing :ne review cf :nis pr: gram.
- 3. Summare - 491dinc and NCE Activ'-ies Cverall, :ne 3&R :regram for welding ,4eicers cual* dica-i:n,1CE and general training activities accears :s :e effective anc aeli acninistered. This caservation is succor ed :y /isual examina:icas, Ocservec NCE results Inc :ractices, 44: ness *ng
'iele welds in ;r: cess anc the ;eneral levei of 4crxmansnt:
demonstrated. I'/- 14
- ~ ~ - - .. ..... ,.._. .
4'q Findings in the HVAC area with rescec: to undersize elds weld inspection practicas and lack of acecuate docu::enticn indic""a sericus 4A/QC deficiencies for tha: articular wor 5, 0 h d
.O e
[ t d e= pm 4 I7-15 t e L .
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- 1
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L c ...- x 4. .~. x4 ~ is ,v: w 5,Ci
. m e,
m u, om l'i M., .b
~ y, M
.M.y h, ~ b!5m '" h ML"in..m.h Ci mer.'
Itlltaltn
- ,I bi':3
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35 on rJ x r-5 55 no 45 no 5DbC QDUE s r n: o 7. n s: o :-
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- ~
d N555 lii~T - - - ' ~;ik-~~~ - - --- li leTsiiiTicient iiisiBiiiF.iiiliil isiiiiiiiilisil ill'ii. I.iii-j; 24555R C5- 1-Illi- 30 7 1 11 - 1 --- ok --- --- 11 was mise.cil ori s epair .hnt. lit.R wri t t en, suliscisncnt 1 2455511 05- 1 -lt11 .111 2 1 11 - 1 ol: --- --- 7 RI cont iencil f inalitut. It - 4162 H5- 1-illt - U l11 7 4R7 --- nl. --- --- Iinw itcsisi1.y notml. lit It pietearcel liy ItAtt. til , veri f ical area to he 0.i114 in. in iler minimum wall. - ennIltinn hail hecen aseeptcolley Iisensee. I l I W-7-Illi- 070 ' 1 IA --- ok --- --- 10 Inification of Incomplele fusion anil aligncil. linear [29030 purnsity not matheit. l'y II? R-Verif ical ley RI. lH.R wrliten g . ti i l' 1 2ftG99 CC An - fil8 2 48tl --- nk --- --- G Irulication of Incomplele insion not markml on . 711669. .Next repair e.hnt of welet showeil same in-tennplete fusinti anel wae. caist3 h t liy stif fer ent int er-
~
preter. this film (71166't) arul 29030 was misicant - ley same Interpreter. 20Gl9 C5-1R11-023 2 5-1 2.72 --- 3.31 --- G Ilim elensity below minimina in area al interest. 790111 Cs-7-An-091 2 III I.f11 --- --- --- 7 inw elensity. I~ i I l l . l s
!.. l
- I IV-lG j l
-s-><_ y, . ._; ,
?. . . . ..
- 7. CIVIL AND STRUCTURAL CCNSTRUCTICN A 08JECTIVE 4
Cetermine by review of documentation anc by indecencent evaluaticn of l c:ecletec .crx, wne:ner worx, ins ection, are es: ic;ivi-ies etiative
- s ne civil engineering area were ac::molisnec in ac:or:ance witn ;
, ;roject specifica:1cns and prececures.
The s:ecific areas evaluated inclucec c:ncrete slacemen , c:ncreta testing, soil ins:ection, protactive c:a:ing, :dn ainmen- liter, and structural steel insta11a:1cn activi-tes. 9
- 3. O!SCUSSION
!. Conersta 31acement i
- This area cf ins:ection was initiatad by a review cf :ne :recaccres
- elinea ed in ne Oreject c:ncre:a s:ecifica:icn anc :na " Civil Ins:ec-icn Manual As nere was it :le er no wert :eing :er'Or ec in :nis area, :ne evaluaticn of :ne c:ncre e slacaman: :rogram was I limited :c a cocument review :/ ::meletac acrx activi;ies.
j Five c:ncreta clacamen: recorc :acxages ere reviewec. ~he 01aca-ments reviewed were as fc11cws: i I 78.ACEME!IT CATE 201-4305-01 9/13/77 201 1835-013 7/'.3 / 30 . i CO2-E773-052 5/06/30 ; 201 1323-005 ;/21/71 105-9855-C02 9/15/50 i The NRC CAT ins:ect:r attame:ec :s c:r elata all relatac cocumentation rtviewed to tne acove concreta placemen rec r:s. l This related cecumentation incluced 0,1e folicwing ins:ecticn anc ) tes: recceds: Cacwelc :ns:ecticn Recores i Ciewelc Sleeve Ins:ec-icn Rec:r:s i 4 Cacweld Solicar Cualifications . i Cacaelc Solica Testing Recer:s i I t ! V-1 l me w - . D
& mm 9 6 e
-- > - ~ . . - +-c.,--r w r ,-~~.--r e-----. , , , ,,n ._r._,, - ,r,- ,- ,,,...,,,.n,,n ~.__.,,,,,._m..- -ng,_, _.
4
. . .. : . :.. _ . _[
i 4
; Reinforcing Steel *ns;ection Recer:s 1
4 Miscallanecus Steel /Embedment Ins:ecticn te:ce:s Cencrete Inspec: ten Rac:rcs l Curing Inspecticn Re:ce:s . j In-Precess Tes: Aec:rcs 1 . < Acci:1cnally, rec:rds involving 1ss:ciatac gecut anc :rassure gr:u-olacement, were reviewec. Fiye "Ccmcressive Streng n of Corts" inc five 'Cefec;1ve Concreta Placemen " rec:r:s aert tviewed inc ::m-2 :artc :: tne accitcacle .'lenc:nformance Ae:cr: (NCA). [n ;eneral, the rec:rds reviewed reflected tha: ::ncreta placemen tetivities were perf:rmed in ac::rcanca 41:n :ne si a :r:cacural r cuirements and -he it:ensee's ?!AA ::=mitments. Z. C:ncrete 7estino I A review of :ne precedures incluced in :ne Civil Testing Labora:O ry . Manual" was performec. The evalua-ten of ne ::ncreta tes-ing program was limited := a review of c:moletac :asting activity ecc'J-ments. Samples of various civil tasting lacera :ry ins:ection rec:rds wers selectec. These samoles included: Mix Cesign Cata !T c Mixes) C:meressive Strength Tests Selectac Aggregate Tests - Selected 'da:ar Tests Related None:nformance Re:crts Acditienally, cer:1fications for cement, sir entrainment, aggregata, and eccxy gr:u were reviewec. Aeinforcing s sel inc miscaiiacecus s:sel cartifica:1cns aere simi'iarly reviewec. Curing :ne review of tnese varicus as: recorts, 1 nas cisc:vertc tr.3-no mix unif:rmity tas 3 nac teen :er#:rmec :: ne ::mmitments ::n sinec in See:1cn 3.3.'. 5 Of :ne FIAR. 7tr2grt:n i :f :nis secti:n, d
C:nstruction :f C:ncette escuirts na: c:ncreta ::nstruc-1:n,
. including mixing, te in ac::rcance ni n : 4200 ' 25ME AC:-359.
Section 50-4223.2 cf :nis cecument, "Ocera:1cn cf Mixers" s a:as,
The range of mixing :scacities and c:rres:encing nixing :imes ?:r all mixers sna11 te ce:arminec :y :ne :er d:rmanca :f mixer uniformity :ss:s as s:ecidisc in ASTM C?A, 'Seeci'ication O' Aeady-Mixec Cencrt:a'. Tacle C -!200
- cefines :ne *as-ing
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e * .
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. , w. . . . . -
. .. .;..- . ,..=.,..: w _ _ _
t frecuency for these recuired tests. The licensee could provide no l evidence that :nese tests had been ;e ~ cmeo. Thus, :ests' ca verify precer cceration of the c:ncrete mixe. s aere re per# rmed during concrete c:nstruction.
- 3. Soil Inscectien Ten soil inspection tes; re: rcs aere reviewed inc fcund :: have been :erformed in ac:orcance witn site :rececural requirements. ::
was noted nat i numcer of :nese testa recuirsc seversi re:ests, :ut acceptance criteria and test erformance dere noted :o be in accor-dance with Oregrim recuiremen:3. 1, 3*ctective Coatine Three Orctective c:ating acciication ins:ection rec:rds and *.he associated schesion tests aere reviewec. One ir-crecess pectactive c:ating acclica:1cn was ecserved. Curing :his ;cetien of :ne review, NCR C-650 was reviewed relevant
- prc:ective ::a:ing 1:clica:icns. This NCR cccumented l deficiencies relevant to a safety-rela:ed c:ncrete placement. The latest revisien Of :ne NCR recuirec the 1:clication of prc;ective c:atings ever :ne concrete repairs, but celeted ne requirscents !:r l
the ace'icable adhesien test. Licensee recresentatives incicatec this arta wculd be reviewed uncer an ins:ecti:n "back-fit" :r: gram I f:r ;rc: active c:ating acclication. However, since -nis ;regetm nac not caen initiatec at tne time of :nis ins:ection, licensee i recresentatives assured the NRC CAT 1.is:ect:r tha: this atter acul: ' t receive accropriate review.
- i. COntaireent '.irer *ns:2111-1:n I ;recacures for the installa:icn cf :ne ::n;3inment liner tre :efirec l :y :ne "Cht:ago Bridge and Iron (C3&I) Hancicok"; 5:ecificall/, Ecck 2, Secticn II.1 and :ecticn II:b (C:ntrac 74-2127/23U). Three separate traveler :ackages were reviewed :: :nese ;rececures.
Samoles were also selected frem ne !clicwing files: Vendor Material Cer;tficatiens
'tisual Ins:ecticn Recerts
?T Examina icn 9e:ce s MPT Examinatten Re:cr s V7 Examinati:n Ree:r:s
'9acicgracnic :nsca::1cn accer s
'Actua. esc;;gra:ns nere reviewec anc are documentec in :ne '4elcing/NCE section, Secticn IV cf :nis recort.
7-3 i e
1: ..i ..... .. . .: . :.: : . ~.:. ..- .. . . . . . . . . . . . . . n.. a -
. . .. . - - . . .. - :. . _ .__ :.;.y [:
u t
. ihe "C3I Nonconformanca Control List"; and the reference to tne
.); applicable recair enecklists were c:m:ared. "C3I Request for ;
Ac:sotance of 'lonc:nfor tity as a Ceviatiendwas also reviewed, i T: The rec:rds reviewed reflected tha: :ne c:ntainmen: liner ins;alla-0 tien activi:ies were ;erformed in ac::reanca wi:n the site. 4 precacural requirtrnents. ! 5 j 6. 5:ructursi Steel installatien ! 4 Pr:cadures delineatad in :ne "Non-.'.5NE Mechanical Ins:ec-icn Manual"
' rela:ad to structural steel installatten ac:1vities were reviewec.
j Fcur structural and .miscellanecus statl installation ins:ecticn
] rac:rcs were enviewed.
s
> One portion of an installatien was ceserved in :ne field. Acolica- -
tien of ;raceacility, wnere acolicacle, was nc.ac. Occumentation j reviewed inclucea a selected samole of: j, Receiving Ins:ection At:cr:s
;. Material Requisition Requests u -
Applicable Cer.ified Mill Tes: Recorts Applicable Cartifica:as of C:moltanca Faericatien Tesveler Sumaries . f c. Ac:11cablehesignChangesancNone:nforancss
- Occumentatien reviewed and tne activity ceserved riflected :na:
. s: rue: ural rela:ad activistes wert ;erd :r ed :: :ne si a :r:cacural recuirements and :ne Itcansee's ::mitments.
i r a i t i f 1.s
7 _ ._.
'i !, PCCCUREMENT. STORAGE AND MATERIAL T?ACElBILITY A. Cbiective
~
re 0:Jac-ive Of nis :ceticn Of the ins:ec-t:n uas :: examice n-sita crocurement, receip;, s:: rage, maintanance anc - ine tr.e adecuacy of ne lican-2.. . . . . , . . - ....... .. ...ese . . ...es.
- 3. Discussion
*he accroacn used c :erform :nis :ar of :ne ins:ection was :s ::ur
*ne si:a anc ocserve c:nstruction activi-ies in :r:gress. Varicus site
- ersonnel wers intarviewec anc sam !es fr:m :iddartr cisci: lines at varicus stages of worx were selec ac. :eliverec ecui en inc catarial at loca:icns. in s:Orage or installec in the :lant were alsc ins:ec ac.
Ac:licable cccumenta-icn relative :nese ac-ivi-ies was als reviewed. Organi:sti*nal cnar~s arc Or0cacur?s der? *1v'twec in *iscussec n! n si e :ersonnel. Of# ices anc 0:ner si e dacili d es aert ::urec anc examinec, 'nclucing am:crary of# ices, wareneuses, Outsica iay-c wn areas, and the olant areas of Uni s 1 inc 2. A selecticn of 71 samples was mace f:r ;recuremen; anc s:Orage ::nsi-cera:1cn fr:m mecnanical; electrical; civil /structura!; instrumentaticn anc ::ntrol; hea:ing, ventilaticn anc air c:rci-1:nt9; (.-LAC) ; aei cing and miscellaneous catagories. In acci:1cn, 73 samales aere selectac s:ecifically for ma:erial traceability, inciucing !8 welc .'oints, eten involving too materials c ' :e neldec ai:n cre Or mort aele fi' lea atarials.
*be f:llcwing descrite :ne results :# ins:ec-icn in :ne tress l's ac.
- 1. Procurement arc teceivino A total of 71 pur: nase crcer files for safety-rela:ac ecui: ment inc material wert examined. A cetailec inscec;ien nas mace ad 35 samoles. Referenced engineering s:ecifica:icns, :uali y requirements, su:mi tals, ac:r ved-vendor status, revisiens, arc ather aspects were reviewed.
Cuality Centrol (CC) authori:2:icns :: sni: fr:m vencors' :11n:3 arc
- n-sita OC receiving ins:ec:1:n cccurents ::n .ere examinec. Me recai:t f requieta ma: art al car t fi:ati:ns was exam 1 rec.
Rec:r:s reviewec andlec :nt: :r:curtment anc asce4v ng activ' ** es i sert :er or d ec in Sc:*rcance 41:n 3r:Cacur3I recut *1ments. ** *as "otac *.na *ne :*m utart:Sd sys;tm imOlementec 1* *ne s'*a Or0Vi:Sc icecuata C*n*rol ad tender su:mittal s' Ind enjine tring 3C:r0v3I :f ' .
*as: Ptcorts and 03:3.
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- 2. Storage are . Maintenance
- a. Storace Faciitties - General Warehouses anc lay-cewn areas for Class A, 3, C, C anc E s:Orage levels wert examined. The Class A s:: rage space recuiring
- temcertture and humidity centrei was fcund to be within con:rol l limits. Discussiens wi:n warthouse ;erscnnel revealed back-uc . previsiens :: maintain c:ntrol in even: Of a site ;cwer cutage.
1 Wald red st: rage areas in Warehouse A anc evens in .:teld Weld Rec Recm 92 wert examined and founc :: :e maintained'witnin requirad temcerature limits. Storage facilities coserved and ete:rcs reviewed reflec:ec :na:
, the identified storage facifi:1es c:ccly with sita prececural recuirements, hcwever, a lack cf control ever two untacted areas accarently used as scrso yards was cise:verec. These areas and c:ntants tnerein were as follcws:
(1) Numercus piping, valves and cener 1:ams aere icen:ified en the ground in a fenced area with the gata removed on the Eas side Of : e Pipe Shes. No signs identifying :nts area wers cese rved.
* (2) Numer us cables, acters and otne* items were f:und en :ne grcund in an area wt:n no enc 1csure en the North side of :ne Elec rical '.sy-dewn Area. Nc signs icentifying tnis area wert 'ceserved.
The coserved lack of c:ntrol of these :wo areas cces no: meet NAC ce sita :r:cacural requirements for :ne c:n rol of equi; ment anc material .
- 3. Storsce in Wareheuse Buildings .
The s:Orage of equi; ment and catarial in s : rage bins and lay-dewn areas wt:hin wareneuse tuildings nas examined. Activities esserved and recorcs reviewec esfiectac :ss s:: rage activities in warthcuse builcings aert :er ermed d in ac::rcanca niin sita precacural recuirteen 3.
- . Storice in Outside Lav-Oevn actas The s:: rage cd scut; ment and ma:erial in :u:sice lay-cewn artas was examined.
Corresien, cus: and dir aert nc:ac en a num er of safety etistad items due is :ne f act :f :: vers anc/cr :re: active ::a:ings. ~; r examole, 123 ;T7-Geinnel nanger struts anc secr:ximately aCC
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- e NPSI hanger s: cats, with associated bolts and nuts, were noted in outside lay-dcwn areas ncr:n of Wareneusa A. Nuts, bolts and bearing pins were noted to be ver/ rasty. Searings were noteo ::
- e c:rr:ded anc dirty. Scecific samples ins:ec:ec were:
(1) The nu: and :nreadec portica of :ne stru: was c:rr:cac relative to ITT-Grinnel Hanger Strat, Serial No. E3173, IO No. E-77.-040, Cl ass .2. Searing pins anc retainer rings were correcec. Bearings were cry an: c:rr:ced. (2) The bolts, tne nuts ano the bearing sin for NPSI-Hanger Strat, C?-0016A,13 No. SI 2025113512R, Class 2 were i correced. The bearing was cry, D:rrocec, anc c:vered 41:n cust. (3) Corrosien was cc:ac en cuisice Of :ne Cesech Cat Cer;. < Aut:matic Nuclear 5:rainer, CP-C029A, Serial No. 23C80,1977, Zurn Tag No. CP2-SWSRAU-02, Zurn Serial No. 5250. Cerrosion was also nc:ad :n similar ecui::en: s::rse ir. s near:y Iccati:n. These exacoles indicate a failure cf :ne iicansee :: satisfy
- ne FSAR cemitments and to crc:erly #clicw sita crececural requirements 'or pretaction of ma erial anc equi; ment in s::r:ge.
+
- d. In 31 ace St:rs;e of Ecuiement The storage of ecui ment in laca in :ne CPSE3 clant was ins:ec.
ted. Cerrosi:n was notac, anc some accc, metal anc pacer rssa
- aere notac acjacant to safety-rela ac e 1
- cent. Scecift:
- examoles no:ac wert:
(1) The Mot:r Ccerated 'laive, Serial Nc. 43554. Tag No. GGC202SGMO, Safeguarcs 31c;.-2, I eva:1:n 790 f . <as caservec :: De wi*ncut a pr tective ::Ver. De crive mecnanism was notac to te Orf. 4 (2) The Auxiliary Feecwater Nec anc Mot:r, Fuma CF2-AFAFMC-02, . and Mot:r CFZ-AFAFFO-02, were not' c:vered. C:rresien aas notad en ene base and weed, metal anc :acer rasa were actec cn and arounc the base. (3) 3e intake fittar Of the Cemeenent Ccoliny Watar Nmc 'd:::r:
*F A-COAFCO-OlM, Inc C71-COAECO-02M wert 0.cggsc wi:n # rtign
+
- sterials anc ceoris, :nus etstri ing :ne tir !kw :: One meter.
'4) 0:nstr;c:1:n ta:arials wert :rtsan: n :ne Resicual "es:
Removal (RHR) N=c Mc :rs, 73X-RHAf tH-0*.M and 3X :HA;H 42M. 011 was cesersec to :e dri:oing fr:m :ne insice :! :ne thR to:Or casing. l The exam les of 1 recer s::rsge cf installed ecui: en: tent *: rec acove Peerssent :ne licanses's failurt :: satisfy :ne sita
! s::rsge ascuirteents.
i
'/I-3
._ ..; .ya.:p :
e paintenance The activities to c:ntr 1 and perform maintenance by :ne till-wrignts for mecnanical equiccen , anc by :ne elec ricians for electrical equipment were reviewed. A can rali:ec carc record systam is used by eacn greuc. Generic maintanance schedulas are applied wnere applicable, and aceiti:nal ins ruc:icns are cotained from ne accrocriate ciscioline angineer wnen neeced. For eacn piece of equipment requiring raintenanca, a card listing
- ne maintenance work and schedule is used. A (C representative verifies the work ano signs after esca main:anance activi;y has
- een ccmoleted en -he ecui; ment. A numcer of maintenance record car s were reviewec, anc no ceviations were notac.
The informal manner (often by telechene) of icentifying ecuipment recuiring maintanance, particularly s:ecial maintenance :: :: moly witn vencers' instructions, was cues;ionec :y ne NRC CAT inscect:r. The actual main ananca -crk and scnecule are scecified by ciscipline engineers for use Oy :ne crafts, but there was cencern that recuirec maintanance may :e emitted fr:m items in s:Orage at the pian: witncu: a : ore fermal sys am cf c:ntrol. Similar c:ncarns icentified :y :ne licensee relative :: maintanance requirements and seneculing of maintenanca are ciscussed in'the Quality Assuranca sec-icn (Secticn 'l'II) of this repo rt. All maintananca rec:rds reviewec incica:ad main anance was performed s :ne recuiremen:s in ac::rcanca wi n :ne specified senedule. .
*n additien, a lack of adequa a rain:ananca :r:cecures to pr: ac safety-rela:ac ecuicment fr:m c:r-esien anc/cr ::ner camage aftar installatien and before :urncver :: TUGC0 for ::eration nas icentifiac. Curing a ::ur of the clant, ex ansive corrosien was noted on ene tube succor;/s:acer pla:as of a Unit-1 Containmen-5: ray Hea E.thanger nat had the cutar snell removec for an intarnal =ccifica:icn. Ciscussiens revealed ::a :ne licensee was aware of :ne c:rresien c:ncition. There nac :een a : ire delay between :ne ins allation of :ne heat exenanger and One turnover to TUGCO. Thus, the Nitregen ciantet requirsc price ::
ciping installatien hockuo had been deletac a :ne time of i:ing neckuo, and no protecticn against internal c:rresien had been a:clied aftar : icing neckuc. This :r:blem led :: an examina:icn of :ne :wo heat exenangers for Uni: 2 :y ne NAC CAT ins:ec::r. The Unit 2 Containment Scray deat E.t:nangers :ur: nase Cr:er (P.O.) C7-0050, *oseen Cat i Sens, Inc. vencer s : rage anc installa:icn instructicns were enviewec. I was 9c:ac tra a
. Mitrogen :lanxe: cn tne snell sice (safety-rell:ac Class 3) nas s:ecifiec 1: 10 :sig, :: be enecked mentaly, f:r moisture revention. No ins ructions were :revicac f:r :rotacticn after installation acckuo :: Oi:ing. Si a main ananca cares maintainec y One millwrignis :allec f:r cnecxing ne 11:r: gen :lan.<a:
weekly, tu: no instrue:1cns were :revicee !:r :re:ac:1:n after installacien accku:. Waintenanca :ar:s f:r :ne :wo '.'ni- Z nea; exenangers revealec :ne follcwing:
'lI 4
(1) For heat excnanger CP2-CTAHC3-02, site maintenance of Nitrogen blanket was deleted 1/5/80 at installation when opened for picing installation hockua. Thus, this heat exchanger has no been under a nitregen curge or any : ner
;rc: action for ever Zi years.
(2) For heat excnanger CF2-CTAFC3-01, si a maintenance of Nitrogen blanke; was celeted 11/5/31 wcen ccenec for picing ins allation neckuc. Thus, this hea; exenanger has nc een under Ni r gen Dianke: or any etner protec-icn for over 1 year. Licensee personnel are aware of this Or:Olem and are inves-i-gating c:rrective actions :: remcve any corrosien anc :reven: future corrosion as ne heat excnangers are :es ed and :iacac in c eration by adding a rust preventive to :ne liquic involved, and possibly acplying a enemical cleaning : recess. Censultaticns with chemical engineering ;ersonnel of Ccw Chemical Com:an/ anc Westinghcuse (the NSSS vendcr) are in Or gress ;c resolve tnis# pecclem. The licensee statec :nat a solu ten is antic:atec 1 ar
- na for nc ming Hot-Func-icnal Tes: On 'Jni '. has been c:mele ac, and One internai cerrosion c:ncition of ne Unit i nea-excnangers is :et;er defined.
Significant effer was ceveted to ciscussicns anc examina:icns of samples and rec:rds ;er lining c traceacili y of in-clace ma: aria! - back to engineering drawings anc scecificati:ns and Oc :r curement scur as and " eat num ers. Generally,
- was nc:ac :na: :ne li:ensee 9as Oiscac a ci;n cegree :f em:nasis n :ra:aa:ility. A : 2' f 73 samcles were examinec f:r tracsa ilt:y, incluci g:
53 Safety-rela:ac picing aeld joints, eacn involving rac materials anc Oce :: nree weld filler matar a's 1 Class l, safety-related structural weld join 4 LO:s of matari.als f:r safety-rela:ac nangers 1 Lc: of HVAC structural su; or ma arial 2 Lots of HVAC cuct material 3 Multi-lo: sam les of weld filler matarial 1 Lot of safety-related fastening devices 3 Lots of miscellanecus snim s::ck, anc hanger stru: careware Scecific resul s Of ne ins:ecticn regarcing ne racaecili v trea are as ? ll ws:
- s. Safety-Relatec ':ine Wel: Joints A :::a1 of 53 :1cing aela join s aere selectac ?:r raceaci'i y.
C/ :nese,12 aere selectad :y cserration of ::mele ac acrx in
- ne clan:, anc :ne taiance inclucec aelds icenti'ied :y c:ner CAT taam mem:ers in :ne acclicacle sections f nis e:or;.
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.1 Ouring early examina:icn of 27 sample wele joints in tne plant, ?
, -) heat numbers could not be found on the pipe, but recercs were
, founc to be c:mplete. urther examination of :ne joints in the ~
. ;lant resulted in locating the a hea: numcers that were no readily 1cca:ed'. Discussions wt:n licensee perscnnel revealec
^
- nat QA reviewed spool pieces at the fit-up stage and ch cked all
, materials for correctness and traceacility to heat nuccer; prior to welding, and again after welding before QC signatures were i .; acplied Oc :he records.-
a Also, during :ne examination of rec:rds for samole weld joints, record ciscrecancies were notac for 'wo weld joints. A review of
- ne welcs anc records involved revealed :nat cesign changas nac een mace, anc :na':, wnile ne actual deles were c:rrect, the documentatien had not been accurately enangec. Tne wo record
, discrepancies were corrected by a Ccmcenent Mccification Card ! - (C.vC) enange, Serial No. 37228, initiatec 2/1/83, and a Manufacturing :lec:rd Sheet (MRS) c anga, Serial Nc. M:-412, da ad 2/22/83. ,, Samole results of traceaoility examinations of neic jcints are as fo11cws:
. (1) Weld No.: W-21-1 (shco weld - pice :: valve) i Welder: 3YL ,
i System: Servica Watar Orawing: SW1AB14, Spool 7Q3 Class: 3 l- Pip'e: Heat No. N!894 , Valve: Hea No. AJ330 i Weld Recs: Heat Nos. 746100 anc 762550 i (2) Weld No..: W-8-1 (sne; weld - :ipe :: Oi:e) 4 Welcer: AMA System: Chemical Volume Centrol Orawing: .CS1ABCO3, Secol 2Q2 Class: 2 Pice: Heat No. 713876 4 Pipe: Heat No. 28970 l Weld Reds: Heat Ncs.12S2R3C8L anc 462638 o
~ (3) Weld No.: . 4-1A (field wele - pice :s nelc-o-ie:) i l Welder: A87 .
" ntainmen: Scray System:
- Orawing
- CT*.55023, Socci *.02 ,
i Class: 2 ! Pi:e: Mea: No. 280385 Weld-o-let: Hea: No. 320J Weld i Reds: 463638 and 5465C6 l i I s
$*b
. . . ~ . . ...
. ... = . . . . . . .. ._. . . . . . _ .._. .
1 2
! (4) Weld No.: W-1 (shop weld - pipe to fiange)
Welder: AGZ
., System: Safety Injectice.
Orawing: SI15311, Secol 102 Class: 2 Fipe: Heat No. 5WL15M Flange: Heat No 51J2L Weld Red: Heat No. C3220E (5) Weld No.: W4-1A (shco weid - pi:e to pi;e)
< . Welder: AXP, ATX
- l .Systam: Chemical Volu=e Centret Orawing: CS1ABC05, Scool 1202 Class: 2 Pipe: Heat No. 5*.391 Pipe: Heat No. M-2253 Weld Reds: Heat Mcs. 1232R and 163628 1
(5) Weld Mc.: FW-153 (pi:e t: ~' albew) Welcer: 5CJ, ATI System: Safety Injecticn Drawing: SIISBCCA, Secci ICC2 Class: 2 Pice: 1 eat No. 28970 i 45' Elbcw: Hes: No. U LLH4 Weld Reds: Hest Nos. 463552, 463638, 546506
, (7) Weld No.: FW-12A (pipe to pipe) t Welder: AGR System: Safety Injection Orawing: SI1SB004, Secci AG2 Class: 2 Pi:e: Heat No. MC635 Pipe: Heat No. MC623 Wele Reds: Heat Nos. C:nsumacie insert 4292R3CSL
- and red 163638 (8) Weld No.: FW-18A (pipe te valve) r Welder: 3CJr AGR System: Safety Injection Orswing: 5I155004, Socol 502 Class: 2 Pice: Heat No. M0623 Valve: Heat No. (5/N) 35740038 Wela Recs: Heat Ncs. C:nsumacie inser- 53C1772'.5 anc rec Ta61CO (9) Wela No.: i-10 (fiele weld - si:e : 45' elbow)
Welcer: AGN System: Centainment Scray Orswing: CT1SB023, Secol 702 Class: 2 Pice: Heat No. MC624 15' E*bew: Heat No. U4KCH Weld Rec: Heat No. C222CE
' VI.7 9
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. . .. ~ _ . . _ . . . . . . - . . _., .. . _,_ , . -
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~i J (10) Wald No.: FW-7A (field weld - ci;e tc 45' elbcw) l Walder: AFN,ABT J System: Containmen: Spray J Orawing: CT153023, Secci EQ2
'] Class: 2 4'; Pipe: Heat No. MC624 Jt 15' Elbcw: Heat No. U4LLHa a
- Weld Reds: Consumaale insert 2526T3C3 anc
/ red 463730 NOTI: Certifications verifying hea: num ers for :nese samoles
.. were in the central files.
- b. Class i Structural Weld Jcint One weld :n a large structural succer: furnished by Wes:ingneuse
! and fabricated by Teledyne Sr wn Engineering was examined for traceability, with satisfact:ry resul s as f:11:ws:
i Weld: Mcun:1ng Plate :: Shim P* ate (no numcer assigned)
- Welders: 3CA, SKX, CEA, 3XU Item: Whip Restraint for Safety Injection System Orawing: EIV-0527-0508-2, Lecp-3 Class: ASME III, Class 1 .
Plate: Heat No. E130514, Mtl. ASTM-ASES Plate (Shim): Heat No. 1G0200, Mtl. ASTM-SA-36 Weld Recs: Hea: Nos. 431L245, 52592, L22550 and 42175132 *
?! Certs: Chemical and physical certifications for ?la e 1 materials and weld rces were in :ne rec:r:s file.
- c. HVAC Structural Sucport Material
. One lot of angle iren purchased uncer Purchase Orcer No.
i TUSI-1113 for Bahnsen Service C:m:any (HVAC 'c:n:ractor) was examined for traceability, witn the fc11cwing satisfact:ry results: Item: 140 pieces of galvanized angle f ren 4"xx4"xi"x20' Spec.: Mtl. ASTM A-36; Galvani:ing ASTM-A-123 ~3 Heat No.: 425340 Color Ccde a:olied by Sahnson: Red-Grange Car s: Physical, enemical and galvani:ing cer-ifica:icns wers in rec:res file.
- c. HVAC Ouct Material Two lo:s of e:si sheets :ur:nasec uncer Pur: nase Order 'ic.
TUSI-0775 for Bannsen Service Ccm:any (HVAC ::ntrac:cr} were examined for traceability, w1:n :ne folicwing satisfac: cry results: . VI-6
,-.,.._..._.u.. .
1 - JJ l l (1) Itam: 149 pieces of galvani:ec sneet metal 18"x120*x10 GA ' i Spec: ASTM A-525, G-90 coating - Hea: No.: 500H2480 Color C de acclied by Bahnsen: Cautien Yelicw Carts: hysical, chemical and coating :artificati:ns were
. in recercs file.
! (2) Item: 23 cieces of galvani:ed sheet metal 18"x120"x10 2A
," Scec: ASTM A-525, G-90 coa-ing Hea: No.: 515N1509 Color C ce appliec by 3annsen: lare Rec 5
Carts: Physical, enemical anc coating cartifica:icns were
, in records file.
- e. % eld Filler Material Three purenase orcers (70s) for large :uan:ities of wel: filler materials were examined for raceacility as ic11:ws:
('.) 70 20725 :: Sancvi'<, :nc. Examina:icn of rec rds snowec traceacili y :: s:ecifica:icns, QC Receiving Ins:ection Recor:s, anc :artificatiens for chemical analysis and pnysical es:s for varicus iets. (2) P0 C?F-1C49-5 :: Murex Welding ?-:cucts, Agen: f:r Air:: ' Welding Procucts.
. Examination of recere snewed traceacility s;ecifica:i:ns, OC Receiving Ins:ec-icn Recer.:s , anc material' certi M eations for :nemical analysis anc pnysical as s for varicus ic:s.
(3) 70 14920 :: Arc:s Cer;cratien. 5xaminaticn of receres snewec :racaa:ili y :: s:ecifica-icns, CC Receiving Ins:ection Reports, and cartifica-icns #ce
- cnemical analysis and pnysicai tas:s.
! f. Safety-Related Fastening 'Cevice (Stuc)
A samole frem Me field was randemly selectad. Informaticn Oroviced was that the item was a stainless steel stud with notations 0F50, 550T anc YA. The sameie satisfact:ry ins:ec ::n resul s aere as f E:ws: VI-3;
/
* /
- - - .e . ,
= . , ,+,
, . s' t* "
[ .r. . at 3
- s. + m g
- . + . -. . :. ~. . .
.x . . a . . m . . .. . . . . ... . _ .; . .. . , .- _ .-- . ..
,y .- -
'i 3
1 1 j Systam: Heating and Ventilating System (VA)
-j Orawing: VA-X-AB-004C, Rev. 6
- 7. 0.: 30228, Item 20 ir I am: 7/S"x5" Stainless Steel Stud, f, Scec.: ASME iA453, Grade 660 h Vencor: Texas Sol: Co. (en ap roved vencer list)
; CC Receiving Ins; action Recort: RIR 13347 Car:1 fica:icns: Chemical and physicai asts in file Heat (Tract) No.i UF60 C:de Class: 2 fi Material Recuisition: MR-154401 Acolica:icn: To attaen flange of ciping fr:m Hycr ger. :urga Exnaust Fil ar ?ackage : flar.ga Of Hea-in; anc Ventilating Systam.
i. i MCTE: T' after 660 designa es the vencer (Texas Scl: Co.)
- g. Miscallanecus 4
(1) Two purenase creers for safety elatec snim st:ck were d examinec, wi n the folicwing sa:1sfac::ry resui s: (a) P.O.: C7c-1618-5, stainless Stael Shim 5:cck Receiving Inscection Reccr : RIR-19319 Certifications: Received with sni; ment.
, (c) ?.0.: CPF-1725-5, Stainless 5: eel Shim St:ck Receiving Ins:ection Recer:s: RIR-1889g, RIR-139CO
- and RIR-19633 Cartifications: Received witn sni: ment (2)~ Harcware anc materials for a samole Hanger Stru: were j
examined. Results are as follcws:
. NPSI sucolied nanger.
CWG. No.: CT-2-001 406-532R, Rev. 1 Drawing Item No.: 4-Rigid Sway Stru:
~
Harcware and Material Icentifica:icn Centrol Numbers t (traceable to heat nuccers): Pipe - NP 298 Eye Red - NH 531 Nu: - NR204 Eye Rod NH 531 Nu: - NR 236 3racke: - NF 1297 Hex Nut - NS 280 Pin - NH 346 Activities ecserved and rec:rcs reviewec reflectac na: ne racancility ac fvi-ies were :erformec in ac::rcance wi n si:a Orccacural recuirements. Em:nasis en :ne necessity fcr ;ccc tracafoility was accarent -brougacut ne si a crgani:ations. 9 Y!-10 1
- ,~~
! 's A. Obfective The Ocjective Of this cr icn of the ins:ecticn aas :: de ermine if cuality centrol inscect:rs func: ten freely in ;erforming tneir :asxs, wi ncut intimicaticn by craft ;ersennel cr sucervisien; and if inspec:icn ;erscnnel are cualified, trained and have :ne organi:ational freecem :: perform :neir tasks.
- 5. Ciscussien
'.. Procram Recuirements The Quality Centrol Program is definec anc inclementec Oy :ne C:r-
;crata Quality Assurance Manual, and ty mere detailed Cuality Assurance crocedures and instructions wnich are encorsed by management directive . .uanagement anc su:erviser res:ensibilities have teen described in these pr:cacures.
ine -inal .are y ,naiysis .,ecer: .. en, : ntains implementing sta aments for 10 CFR 50 Ac;encix 5 anc provice c:mmi: ment stataments to regulatory guides inclucing NRC Reguia ry Suide 1.53 (ANSI N45.2.5), wnich define Quality Program recuirements. Quality Assurance precedures were develeced :: im lement :nese c:mmi tments. For examole, CAP-2.1 cefines :ne recuiremen s for One . raining and certifica:icn of mechanical ins ec-ion perscnnel.
- 5. Seocram Im lementatien n:lementation of this :cr;ien of the :r: gram aas cc:erminec # :m discussions wi-h Quality Centrol cerscnnel anc :neir su:ertisces,
#r:m a review cf ins:ec:Or training, fr:m a review of certifica:icn Or0cecures and ne inscec:Or verifica:icn #iles , and fr:m a review of ne sequence of esc rding anc permanen-ly filing ne resul s cf ins:ections.
Interviews were held with 53 inscec crs ranccmiy selected fr:m Texas Utilities Generating Ceccany (TUGCC) and contractor crgani:aticns cerforming inscections on site (Brown i Rect, fannsen Services, anc Grinnell). These discussion included :ne ins:ect:r's area of assignment,c' ackgrounc, training anc educa:icn, per ection of new
- ncrcugnly ins:ectors were rainec and :recarec : :er# r, ins:ecticns, interfaces , anc.
involvement ela-ive. : ::ns.truction craf: inciucing :ne cresence or any Orm :..in im1ca-1:n. C:mments c:ncerning ins:ec::r cualifica-icns in ne area Of ncncestrue:tve examinati:n (Secti:n I'/} anc s:eci'ic ::mments concerning electrical (Secticn *:) anc recnan' cal (Secticn :::' inscection procacures are presentec in :ne a :licaole secti:ns :f
-nis recort. ;
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M a. Ins ector Cualificatien/Certificatien
- I ij (1) Inscectors were required :o attend training sessiens, perform
, indepencent reading of stancares relative c their ins:ecticn area, anc were testad witn regard :: this training material.
;d Ins:ectors were required to ;articica e in On-the-job
'4 training (0J7) whien was verified :y a senice inspec: r 3 designated in writing to verify nis training.
A y Apprcpriata f:rms and certificates were n file, in
- s accorcance with ANSI N45.2.5 re:uirements, attesting to the
- inscect:r's background, ex:erience, ecuca:icn and training i
. and cartifying an ins:ec;ien a acility in ; articular
~
y 4 construction activities,
.s.
1 (2) Curing the review of the training recares and fr:m intertiews with su;ervisers, it was f und na scme OC ins:ec ces were i certified with less ex:erience than recuired. For example:
, ?.
s .? a. Three fraivicuals were certifiec Level II (L *I) as
'2 mechanical ins:ect:rs having au:ncri y :: witness puma Or
. c mpenent cisassemoly anc reassemoly ni n cualifying experience only in welding anc noncestructive examination.
, b. One individual was cartified L-II as mechanical ins:ector y having autnority to witness ;um: cr ccm enent disassam:ly
. and reassemoly using ecucation as a fac Or in :ne 4, qualification precass when the ecucati:n was fr:m a ncn-technical, unrela ed college cegree.
- c. One individual was certified Level I (L-I) electrical inspector after eniy 3 neeks of electrical ins:ecticn 3.:
- ex:erience,
- d. Two individuals were cartifiec L-I anchor colt ins:ectors
. - with less than 1 montn ins:ecticn exterienca.
ANSI N45.2.5 recuires a minimum of 5 mentns ex:erienca fer e L-I when a cancicata has a hign school di lema er ecuivalen: ecuca:icn, and 3 years ex:erience for L-II when a candicata j has a high sencol diploma er ecuivalen ecucation.
- b. Recercinc Ins:ection Results -
(1) The Ins:ection Reccc: (IR) form nas ne documen: ;rimarily
. used :: recere ims:ec:icn resui:s. A review was mace Of tre recording, reviewing, anc filing of ins:ecticn reccr:3 41 n
- ne folicwing resul:s:
- a. The FSAR and the TUGC0 rescense :: NRR Generic le::ar -
31-01 stata na: TUGC0 is in c:mciiance wi:n Regula :ry Guice 1.58, wnich encerses ANSI N45.2.5, "Cualifica:icns of Ins:ection, Examination anc Testing Persennel for Nuclear ?cwer Fiants." ANSI N45.2.5, Secticn 2 anc Ta:le '. s
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. -. ^.- . .
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4 provices the levels of cacability fer project functions and cefines ine= as:
. L-1 capable of recording inspection, examination, anc testing data and imclemen-ing ins ectica, examination, anc testing procacure.
. L-il cacable of :erforming as a L-1 :lus; Planning inspections ....
-Evaluating :ne validity anc accet:a:ility of ins:ection .... resul s 2,e ceting inscection, examina:icn . . . . resui ts Supervising. ecuivalent . . . . :ersennel
, Qualifying icwer level perscnnel The requisite qualification for nese cacability levels is crovided witnin Section 3 of :ne stancarc.
- b. The .'IRC CAT inscectors f:unc curing in arview anc cocumen review that:
^
. With a few exceptions, as statec above, inscec:crs were ex:erienced. )
. In the areas reviewed GC inspect:rs c mole ed and signed irs. L-11 certifiec inscec:ces in ne electrical and instrument areas of inspection reviewec the irs, but cic not document their review. In scme inscecticn discipli-nes, sucn as ,echanical (ncn '5ME) acc ::ncui succer :
ex:erienced leac inspec:crs (cesignated L-1) reviewec ne IR before it was sent to file. In other inspection disci: lines, such as Civil CC, ners were no routine reviews ;er#creec Oy an ex:eriencec leac inscector. In each of :dese cases, :ne rescr:s were no: signed by the reviewer. Occument reviews revealec that ins:ection ins: ructions anc inspection rescr:s were ce: ailed and inclusive. Licensee recresentatives indicated that ne reason for these detailed instructions and reccc s was tha: they c uic be ccmcieted :y ins;ec:crs wita no acciticnal reviews y c ner inscac: Ors or sucertisces. The intent of ANSI M45.2.5 is :na: a L-l' Ce :ne ins:ec::r
- Of recorc. The praC ices in place at "ne si a ci: 9c:
ensure :na: this recuire. tent was sa-is'fisc. i
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- c. Ins:ector Intimidatien Ouring discussions with ins ec crs it was reveaiec tha: in ene section of ne ins:ec-icn organi:a-icn threats anc intimidaticns hac been mace. This information was transmi tec :c NAC Regicn I'I since an investiga:icn in this area was currently in ;r:gress.
.an inscec:cr from a different ins:ecticn area re:cr:ec previcus . threa s, wnich resultec in :ne craf: :ersen making ne nrea s being removed fr m -he croject.
Asice frc -he engcing investiga-icn, 1 uas reveaiec that agressive action was taken ~y management ;revent instec :r in-imica:icn. Ins:ec:ces frc cne s e-d~ r ecer:ec :nat
. '~1 1s ins.m. :ctic.n was issuee
..,s :-.
ay memcrancum anc 1s c4.scussec ; :ecticn u. anc :ect.cn 4
-his rescr .
D 4 4 l v e 6 epa . a.
V!:!. OUALITY ASSURANCE A. 05 factive The cojective of this review was :: de: ermine :ne acetuacy of :ne licensee's Quality Assuranca (QA) Program. The :r: gram.was reviewed to cetermine if it was aporecriately estaclisned in instructions anc manuals; and if the constructica anc cuality assuranca effer: was monitored thrcugh audits and other management acticns. In addition, a sampling review of s ecific staes aken :y One licensee regarding :ne eversignt of contracters, c:ntrol of measuring anc est ecuiement,
, cccument contrei, anc control of QA recercs was mace cetermine if s:ecific parts of the program were implemen ad.
- 3. Discussion
'.. Pr: cram Recuiremen:s The OA program is cefinec by a management encersac nierarcny cf general directives and imolementec ey crocacures at the cor: ora e anc site levels :: con:rci c:nstruction activi :es, inese crececures were implemented to satisfy the licensee's Final Safety Analysis Recor: (FSAR) ccmmitments.
- 2. Procram Imolementation Implementaticn of this ;ortion of :ne program was cetarminec casec
- n reviewing :ne organi:sticnal structure, incut f-em :ter NRC CAT inspec rs, :ne c:nstruction audit program, sam: ling :rawing revi-siens in :ne constructic' work areas, and reviewing ne centrol of measuring and :as ecui:jent.
- a. Or:anization The QA organi:stion includes the si;a c nstructica cuali y control organization which is incecencent fr:m :ne site construc-tion management. The quality assurance organi:aticn reccris the Vice Presicent for Nuclear Operations, wncse res:cnsicilities include the construction and oceration of the Ccmancne Peak Steam Electric Staticn (C?S 5). The autnerity and cuties of ne positions involved were :escribec in ne FSAR anc cor:crata manuals. The auci cegani:stion was icca ac -a- ne r:cra a Headcuar:ars in 0 alias. -
~
cesitiens were vacan; and hac :een f:r an ex ancec :ericc. l 1 1f 7. u - r ,
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, b. 2ucits The licensee's audit program was reviewed. A: least 18 audits Out of 50, :erformec betweer.1973 and 1983, were selec:ac and reviewed wi n emenasis on :ne f:licwing majcr areas: g
; auci planning and 2enecui.ng; auci: instructicns anc caecx sn .ts; auct rescris; audit resul s anc follcwuo; and the overali effectiveness of :ne auci: program.
(1) (a) The licensee cualifica:icn anc certifica:icn cr: gram for
.. audi: Ors anc lead aucit:rs was estaclished in CA N Procecure CG1-QA-2.1 "Cualificatien of Audi: Persennel".
(c) The certifica:icn rec:rcs f:r en teac audi rs were reviewed. The lead auci: Ors me: One TUGC0 anc ANS* NA5.2.23 requirements. The review revealed, newever,
. that auditors not meeting the ex:erience recuirements for the lead auci:cr ;csition had been assignec as " Acting . Lead Audit:r", but the limits of an acting lead auditor's autacrity and the quicance r viced was not cefined.
l (2) Audit Planninc and Senedulinc (1) Occument reviews anc inter /iews revealec -hat auci: ;ians were deveicced anc a sys:em f caeck snee:s were usec as guides :: the auditors :: ensure :nat scecific :cints were reviewed. Ocen auci: fincings were als: reviewec curing tne aucit. It was revealed na: ne : neck snee:s were ceveloped cy -he audi Or sssignec :ne audi: or by the audit grcue su:ervisar but were net accr0vec by :ne CA Services Manager. Interviews revealed tna audit scnecules were devel :ec using previcus audit findings, schecules , ex:erience, and discussions with constru
- ion si e sucerviscrs cancerning
- c:nstruction preciems.
n.2 , nere was nc un :rececure : cescr;:e ne re:nce : be used :: develco auci senecules or ana: mar.agemen: 1 Orovais ne schecule snculc receive. (3) Auci Recorts au ci: re cris ;revidec a descri::icn of ne auci; sc :e; icentifica-icn Of aucit:rs; :e"sens ::ntactac; a summary f resul s anc a cescrittien Of, any ceficiencies :r fincings.
'lIII-2
. g.
...m
- t 4 -
A , (a) Audit Ceficiency :ecorting and Felicw-u: Audit deficiencies were clearly wri :en anc recuirec timely res:cnse by the management of :ne auci ad Organi:ation. The c.cm letec ceficiencies were reviewec by ne auci: eam icacer cr acequacy. veficiencies were reviewec in sucsecuent (5) :recram Ef#ectiveness Althcugh the audi: br: gram was in lace, here were several weaknesses in the Oregram :na cecreased its effectiveness. (a) Audit Effort The auci: crganization is lcca:ac a- ne TUGC0 Of# ices in Dallas. All auci s are :erformac frcm na location. There are eign; auci Ors in -he audit section. Althcuga auci teams are scmetimes su::lementec by :erscnnel from other sections of the QA crgani:atien, ne signt mem er audit secticn is assigned : :erform audits of su:Oliers, succentract:rs at tne construction site, c:nstruction activities and startu:. Of ne eien; aucit:rs in -he audit crgani:ation, four hac tecnnician backgr:unc and four nac a general nontechnical :acxgr:unc. Nene of r.e audi crs assignec :: ne grou: nad engineering tacxgrounc or ex:erience. In.tertiews reveaiec na a :roxima aiy 1200 man :ays were s:en: recarine for, c:ncucting anc re;crting audits a ne si a in 1982. A review cf -he 32 audits performed in 1982 reveaiec :na: accu: 330 ran cays were scent en site :erferming nese auci s. hd a: ears
- be a smali ;ercentage of the :::s1 auci; effer:
c:nsicering :ne level of effer: cngoing a :ne site. Interviews revealed tnat five acci-icnal auditor
;csitions hac been au ncri:ec for more -han cne year but ne ;csitions were still vacant.
(b) Audi- F-ecuency
- nterviews anc ::cumen: eviews reveaiec na :
. Twelve auci s of ::nstructi:n activi-ies were cer# rmec in 1981. Of nese, six were of engineering anc acminis-rative areas sucn an auci:3 of :E Euile: ins, anc Ort-curement anc six were of c:nstruction fielc activi: es.
. Thirty we audits were :erformec in 1982. Cnly nine :f
- ne aucits were of construction fiel: activi-ies, ne
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-) 0:ner 23 audits were perf:rmec of engineering activi-ies
- $ anc other succor areas. Of the aucits of construction
#ield activities: cne aucit was cerformed of rechanical ljl piping activities, one of restrain; anc snuccer instal-Q laticns, two of electrical work, one of civil work, cne N of instrument anc con:rcis, and :nree :f pr : active j c:atings application. . -s ., ine recuency of audits of c:nstruction activi;ies has been very icw anc may nave centributad : ne :r clems in :ne k; :ecnnical disciplines identifisc in Sections II, III, anc IV of this report.
p a (c) Audi: Effectiveness
. Areas of the c nstructicn ac-ivi y were aucitec; newever,
.. One audits die not icentify maj:e ::ns ructicn Or: gram
; preolems, for examcle:
. Sannsen Servicas was aucitec yearly since 1980. The y '
last audit was in April 1982. Aitncugn facrica:icn anc installatien activities anc ;ersonnei qualifica:icns were in the scc;e of the 1982 aucit, sucn NRC CAT
- icentified items as undersi:ac welcs, cu: of :leranca
' dimensional characteristics, and an inadecua a struc-ural welding inscecticn cr: gram were nc: icentified anc resolved.
. The electrical area of cons rue:icn was auci ac cnly f ur times since 1980. The audits die no identify :ne ca:1e
; separation issue as discussac in ne Electrical
- C:nstruction Section (Section II) cf :nis recert.
Ineffective c:rrective action has teen taken as a resul: Of 1 aucit. findings; for exameie:
. E:uicment maintenance was auditac in August 1979 (auci numcer (TCP-5)). An aucit finding.icentif'ed ::a: vencer instructions were not being inc:r: crated into engcing main ananca instructions. ine eu;y, 1931 OA auci: Of star;u activities (7UG-5) anc ne June 1982 Ocali y Surveillanca Summary, GSR-52-022, icentifiec acci:icnal pr biems 41:n ensuring :na: manufacturers recuirements anc cualificatien recer: recuirements nad been inc:r-
;cratac into ne maintanance ;r: gram. An Oc::cer 1952 aucit (IUG-14) icentified One OnCarn ::a: ecui: cent cualification recerts were nc: reviewee curing ne cr:cass of estaclisning maintanance recuirements. The VII! l
. . . . .. --- - - . .
- eL - . -
g problem was identified in 1979, anc i nas no :een resolved as evicenced by ne 1982 aucit. Thus, :ne effectiveness of :ne ::rrective action sys am for auci ts is not effective.. The NRC CAT ins:ector's c nclusien is :na: neaxnesses exist in the estaciishec audi ;r: gram. Thesa weaknesses incluce One scneculing anc frecuency c# auci s, :ne inck cf effective construction *:r: gram monit:rir.;, anc in lack of effective resciuticn :f scme audit finding:.
- . CA 3recram Interfaces The NRC CAT ins:ector reviewec :ne QA crgani:stiens' cverview cf Occuments that rescribe actions :a.<an by engineering, :enstruc-
-icn, and quality assurance persennel, a Ol anrec ar systemati
- r: gram is in placa wi n :ne as:ec :eing Or:cecure review.
In erviews anc dccument reviews revealed nat: (1) The QA/CC manager Or a senice re resentative reviews ali inspection er cacures. (2) ~he CA/0C manager Or a senice re resenta tve reviews :n-structicn Ocntrol pr:cacures :: ensure :ne :r::er construc-tien-inspecticn interf aca exists. (-,; _-ngineering c:n:roi crececures are nc: rev'ewec :y na 2r..Cc
/ .
manager or a re resentative Of -he OA Organi:t-icn. The lack cf cr0:er intarfice witn engineering may nave contri-buted :c :ne issues ciscussac in Secticns !!, ::", anc :X cf :nis reco rt. These issues reia a :: :ne final ergineering esign anc
- ne final inscection rescr:s not :eing reflectac in ne hareware.
- d. Construction Monitorine A crogrim cf constructicn meni cring was establisnec. ine program consisted of: monitoring :ne ASME : nstruction anc installation activities ;erformed by 3rewn anc Rect (3&R), anc surveillanca cf ::ncreta ancnce bci: ins:aila:icns.
The monitoring Of ASME ::nstruction activi-ies ::nsis ac Of a systa.matic review by two incivicuals of 3&Rs ::molianca accrevec instructions, Or cecures anc/cr crawings na: im:l e. men
- ne recuirements Of ne 3&R 2SME GA manual . The :cni Or dqq nas scneculed in acvanca anc :revicac a review, al ncu:n less formai nan aucit, of eng:ing ,SMc vert activities. _ine survet ' .ance :-.
ancnce bol- installation was :er'Ormec in acccr:ance wi n ~L'GCC
;recacure C?-GP-l' .2 Rev. 2 anica incica:as na: !C
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} installa:icns shcuic be monit: rec eacn shif- f:r eacn cisci line.
A review cf the meni:Oring record since Sec:amcer 1932 revealec
,! : hat nc .nearly :na many ins allaticns were Ionit: rec. In
- Sectamcer 1982, 35 :: 40 installations were monitorec. : rem
, Oct:cer 1982 througn January 1983, 30 40 installations were monitored each Icntn. This is an average of meni:Oring cne .
1; - installaticn each, day instaac cf :en eacn cay.
*n terviews revealec tna:, al heugn :ne ins allatien activi y nad cecreased, inscec rs were assignec : ins:ect ccmciatec wert anc
- ha- -he surveillance requirements of ne cr:cacure were r.c:
.' being met. The ancher belt monitoring pregram is not :eing performed as recuirec by tne implementing crocacure.
Occument control incluces the : n rol of ali :ccuments associa ec wi:n :ne cesign anc c:nstruc:f on cr: gram. This includes drawings, precadures, s:ecifica:icns anc manuals. Crawing c:ntrol was selected for review as an incica:icn of the imolementaticn of cccument control. Centrols had been established : formally icg drawing anc revi-
, sien num:ers into a :ard system and su:secuently in the c:m:u-ter. Occument c:ntrol greuc :erscnnel recreduce -he crawings and revisi ns fcr distributicn. Distribu:icn is mace to a centrai cicku; area. Each major construction organi:aticn has assignec personnel :: distributa the drawings fr:m :ne :icxuo Ocint c crawing con:r:1 ;cints and :: werx areas arcugneu: ne si a. A master drawing revisien list is issuec every :nree men:ns anc eacn cegani:ation is recuirec :: auci: ne1r areas to ce arm 1ne
- na: ait drawings are of the ::rrect revisien.
l ' l The c:ntr:1 cf crawings was reviewed in re centrei anc :istri-l bution centers for :ne 1pe snco, welcing engineering anc ne electric snco. In aedition to :nese offica areas, drawing c:n:r:1 was reviewed in construction wert areas of :ne auxiliary builcing, :ne ca:le s; reading recm anc na c:ntr:1 reca. These reviews revealec :na : I (1) The crawings in ne c:n:rol and cistributi:n :an ars were :f the lates: revision. (2) On . ecruary 3 anc 15,1983, in construc:1cn werx areas, cu-of 70 crawings checked by :ne NRC CAT ins et::r, '.5 . vere nc: l cf ne latest revisien. * (3) Cn Fe:ruary 22, '983 c:nstructicn areas were urec :y :ne NRC CAT ins:ec r. I was f:unc : a: '::r:vemen s ac :een race in rawing c:ntrols. One drawing : an Out-of-ca:a l i
,/ r.11-0
9
. . . _, . _ _ . . . _ _ .. . _ _ . _ . .. . s.__. .
. o revision was found curing ne tour, hcwever :nere were s-ill drawings in :ne areas :na: had caraged :r missing title biccks.
Intarviews witn TUGCC cersonnel revealed na: crawing control cad
- een an ongoing ;r:clem -*= -a-Hrec ' scuent a- ention. 3ased on this samole, c .
- f. Centrol of Measurinc and Tes: Ecui: ment The system of c:ntrol of measuring anc es ecui: men was reviewed. The review inclucec ecui: ten cali:ra-i:n, issue, anc recall.
Ouring a cur of :ne caif bra:ic, facility :ne felicwing areas were reviewec; :ne use Of certifisc ecui; men: during ne cali-bration ;recass aad the relationsni Of the certifiec ecui: ment
- national stancarcs; :ne s:: rage anc se:aration of calibrated ecuipment from equi: ent cut of :alibra-icn or cefective; anc ne general conci:ico of the facili y.
I was fcund.:na: :ne equi: ment used curing calibration was traceable :: national standarcs anc : at ecui: men was crecerly s:Ored. Ctner c:mmen s concerning ne #aciif ty anc s me s eci'ic instruments checkec are contained in the Electrical C nstruction Secticn (Sec-icn II) of :nis re: ort. The metnces of ::n:r:1 anc issue of elec rdcal cacle crimoing
- cols in the s:Orage area, anc condition Of :cis in :ne electri-cal c:nstrue icn ::ci issue station was reviewec. The s:Orage area anc the control of termination equi: ment was satisfac Ory.
The ecui: ment recall was cc=:u eri:ed wnicn alicwec tre :: notification of craf;smen and instec: Ors anen ne : col :ali'cra-tien due date was accreachec. The control of measuring and esting ecuipment ascearec satis-f ac:c ry. s see e eee*f g .g 9 Sd 6. @ 6 e
= 1 .. ,...u..,...._.. . - . _ _ . . .. . . . . . - _ . . ' l IX. CESIGN CPANGE CCflTROLS AND CCPoECTIVE ACTIO!! SYu PS A. n.BJrei.yr r.
he :ur:csa of :nis assessment was : review Or: gram .im:lemen atio.n with emanasis on actuai sarety-relatec harcware :. ns a. .e . c :n :ne, 1ei d, as well as recercs inclucing design cnange c:n rcls and any identified ncnc:nforming c:ncitions related :: installec narcware. Samcles were selected in the tecnnical disci lines := cnecx :r: gram imolementation, as well as to ensure design c:ntr:1, cesign interface, and cesign verifica:icn Orccecures satisfy NRC recuirements anc licensee c mmitments. 2dci icnally, a sam le of recorcs were reviewec 0 determine new ncnconforming conci-icns were identi#isc, cis:csi-
-1:nec, anc :ne extent :: wnica corrective ac:icns aere axen.
- 3. D:5CUS5:CN
- 3. Organizaticn The organi:ation of the generai si a engineering, ::nstruction, and procurement eff:r:s were definec in crececure C?-E?-3.0, Rev.
- 5. By unis precedure, tne Engineering anc Construction Manager is rescensible for the C:manene Peak Steam Elec ric 5:sticn
(-.ec.-a-,; s.esign , engineering , anc precurement. -.inese ac :v -,es . are normally celegatec :: Gib:s & Hill, Westingneuse, anc ctner centrac Ors / vendors. Mcwever, Texas Utilities Services , 'nc. (705 ) has been designa ec by Texas '.':tli-ies 3erers-ing C:::any (7"GCO), the licensee, :: retain :ver:11 res:ensi:i'ity 'Or :nese activities and :: ;erfern cesign func icas as recessary. The TUS: Project Engineering Manager is res:cnsitie 'er ne general cirection of engineering activities. hese cu-ies anc rescensibili-ies uere im:lamentec nrcugn ne -: ancne :eak Oreject Engineering (CFCE) staff and Organi:2 :cnal structure managec Oy ne Comanene Peak Projec: (C?o) Disci:line Engineers. The CPP 01scioline Engineers were res;cnsible *:: acminister an cederiv design change program which c:mplements construction a'nc assures design acequacy".
- b. Discicline Encineerina Reviews
('.) Civil Encineerine Most of :ne cesign enanges 'n nis area aere :recessac via Casign Change lu:ncri:a:icn (CCA) . :rev' us :esign cnanges processac ;rict : :ne cur en: Or:cecura! scuirements were
- recessed via ne 24 sign Change / esign :ev'a:1:n lu nor :1-tien (CC/ COA). A ::si cf :hirty Of :ctn of :nese y:es Of
~
design cnanges ere reviewec. en, 'ni:ia:ec as a esul :f a ncnc:nforming ::ndi-icn, anc were sviewec anc ::::arec : ne assccia ec nonconformance recer: (1CR). Ten were racec b s\
- e
. .. . .. . ,. - . ... . . . . _ ..u.:. - -
- . . - . . - . . . . . . . . - . . .. - .. w. .w.
4 , c by the NRC CAT ins:ector :: O'e lates; drawing and/or specification. The final ten were initiated based en a licensee field ins:ecticn effort. These last ten were
- =:arec by ne MRC CAT ins:ect:r :: :ne crawing and/cr s:ecification in effec 3: ne -ime of ne ins ection.
Cesign enanges reviewed in this area were f:unc :: :e processec in ac: rdance wi n Or:cedural recuirements.
~he NRC CAT inspec Or notec :na: some original designs in tnis area were Oracassec via C:meonen: Mccification Card (CFC). The C.C, M which is a design change accument, was
;recessec in ac::rcance with desien enance centrei recacure secuire.ments for ne esign or Nuclear ecwer Plants recuires
-hat field changes be justifiec and su jected t: design c:ntrols c:=ensura:e ai n :ne crigina' cesign. The res:cn-sible licensee engineering re resen ative incicatec -hese 1 C:ics wcu'c be :rea:ec as a fiele ::ange, and necce, wcule receive Gibbs i Hill review :: satisfy ANSI .V5.2.1..
A selected sam le of Engineering Evaluation of Scacing
'lariance Reports (EES'/), usec in ice succor: tase plate installations , was reviewed. The ex:ansion anchor bel-
. i-e- 11=-4-n Oregram was reviewec :: :.e recuirements cf NRC
/ 7,- av. 2,,"Ancnor Boi- 5ase Plate Flexibility
/0 '
(2) Field Str;ctur=1 Encineerine Mos cf :ne engineering eff:r: in tne TdSI Field S:rJc: ural Ingineering 3r0u: a :ne time of :nis ins:ecticn invcirec cesign cf c:nduit su;;cr:s. The original design Of :nese succor s nas ;erfor ec :y Gi:bs & Hiii : Revisicn S :? ne
. 2323-5-910 drawing "cackage". Frem :nis point, :ne :esign effor; aas dalegatec :: tne TJSI organi:ation a: :ne si a.
Installation was c orcinatec between engineering anc ::n-struction, wien installa-icn :erfor ec :: ne 2323-5-310
":ackages". After installa:icn, ne su::ce. :esign was :nen sen : Gibes 1 -iill "Or -eview : satisfy ANSI NtS.2.'.'.
recui rements. TJS* ;ers:nnei :re:arec no 2CAs : ne 'S-310 trawings, :nly CMCs nere cesign or :esign enanges aere necessary.
. ca-icn of ::ncuit succor:s were :e:e .'nec y ne use f criteria estaclis .ec fr:m ".:ca- On of Su::cr;" (LS: Orawings anc :ne recuirements Of :ne site Electrical Scecifica-icn ES 'CO .
Ca:leracewaysuEccr:3 were ins allec :: li::s i Mii' :rd -
;inal :esign Orawings. 'JS* :ersenrei 'ssuec ICAs :: ne
, su::cr crawingr f:r generic cesign :nanges, sucn as ne
!X-2
f addition of a hanger to the raceway c:nfiguration. They issued CMCs for cesign cnanges affecting an individual sucport. The NRC CAT insced or reviewed two structural calculaticos for cable tray and c:ncui: su::cr:s w1:n :ne res;cnsible licensce's representative. The design incu , verification, and cut:ut (infomation cr viced by :ne Field Structural Engineering Cesign Review Leg) were reviewed. Stress levels, as defined in FSAR Section 3.3.3.3.3.1 anc .:5AR Section 3.8.4.3.3.1 were properly incor: orated into :ne supace: designs reviewed. The NRC CAT ins ector samclec and reviewed sixty CMCs anc Instatiations to tne design document nac been cer o mec cr were in-crocess, :u :ne cesign cccumen; nac nc: teen ' final' reviewed. A review of tne 3ibos 1 Mili " MC Master :ndex" (structural) incica ec :nere were en na cr:er ;f f ur-to-five tncusand of sucn cnanges :nt: nad :een generatec cut had not yet been
- final" reviewed by Gibos i Hill .
It was detemined :nat procer verification of such cnanges
- night ultimately be ac::molished. Mcwever, .he volume of CMCs anc CCAs remaining to :e reviewed by :ne criginal nnicn w1:i :e previcec c:nsicering ne a:Orcacntng :ectem:er,r
- 1983 Fuel Leac Cate. ,
1 Curing -his same review of CMCs and CCAs, :nere was evicerce l of a prcolem relevant to wna: revisien Of ne,cesign cecument a ccmcenent er an activity nas been ins:ectac. OF -C ?M-6 .10 , Rev. 6, " Inspection Item Removal Nctica :cm' accresses subsecuent inscecticn following reccval cf an item, but not for acdition of an item to a cemconent. Fur:ne more, con-structicn and/or engineering ra:rer nan cuality c:n:rci l determine wnen an item is c be reinspectec or wnen scme yce [ of inspection is requirec,
- b. 1 - %
i {f . cr examoie, su cr:s ;r 2,enty :::ia ray anc i f I
, \ cencut- installa:icns were examinec. 0# :nese tventy, Ovelve / l d
were not
- final" ins:ec ec :: :ne lates: issuec cesign dccument, even :ncugn receres in :ne A vault incica ed M[g#p5' C' final" ins:ectionnac:eenper'crmec. _ater CMCs ::vering.
- 7[1 7
cesign changes existec for all evelve of nese installations In accition, :ne licensee's CC ins:ections =ere :er# mec in j six instances to CMCs witn earlier revisions : nan :ne latest l revision issuec anc in ef#ec- / 9 s i)
, ) +nf./
,.3.
7
. . . .. .. . . . ., .a. ...a. . -
T i cerformec [Similar conci icns were disc vered anc discussed in :ne diectrical and Instrumentation Construction Secticn cf the re;cr Secticn II)].
'NSI Na5.2.11, Section E, s a:es :na: pr:cacures snail be 3
- rovicec wnica " assure na: :ne im;ac: Of the cnange is carefully considered, ...anc information concerning :ne cnange is transmi :ec c all affected :ersens and crgani-rations." As a resui cf ne ;r:cacures anc recorcs reviewec relative :: this area, the NRC CAT ins:ector considers existing :recacures have nc assurec Ma: ne informati:n c:ncarning the change is ransmit:ac :ne accrocriate Organi:ation; i.e. , ne Quali y C:n:rci organiza-icn res:ensible for tne inspection cf :ne cacie tray anc c:ncui succor s. This was suestan-iatac Oy recerts in :ne 0A vault.
(3) Electricai Encineerinc
- a. ?recram Review Original design in ne electrical area was performec y Gibcs and Hill. Cesign enanges were recessac via :ne CCA in :nis area.
Tne NRC CAT inspect:r reviewed a::r:ximately -hir:y CCAs in the electrical ciscipline. As nas ne case in teld Structural Engineering, reviews :: be ::mcieted by ne original designer in accercance witn ANS* N45.2.11 are usually cnly ;erfcrmed after :enstructi:n er f as;alla-i n nas been performed. CCAs reviewed were ;r:cessac in ac::rcance witn CP-E? J.7, Rev. 7. Curing nis same review, tne NRC CAT inscec:ce aisc notac a Or:ciem similar
- ::a: in :ne Field Structural Engineering relevan: ::
wha revision of the design cccument a ccm:cnen: er activ-f:y nas been inspectac. As a resul- - *= 9:e c'~ 3-e-ac-f ne s
- b. Cesien Chance 'lerification Gr:ue Review in :ne diectricai Area d
C?-CP-15.1, Rev. 3, *Cesign Change Veri #icati:n :revicec cri aria...(for) review of cesign anc ins:ecti:n cc:uments
- assure incor;crati n cf :esign :nangas int: ne :nysi-cal :lant.
This Or cecure recuirec na des.ign cccuments :e.C..-, . eviewec by One iuuvC v,esign .,ange .leri-,.ca : n 2rcu: t. .e. :: i ascertain weetner Or not incer;cration of :ne ia:es; cesign dccument nad akan ;1 ace, anc if fur:ner ins:ecticn was recuirec ne CCVG 'snali :er" rm" it. Al so , :er :rececure , ne CCVG uas -ecui' ec : selec a re resentative sam:le of cesign cnanges :er :rawing s
.L l. ==
* * * -O^ O' . *O .e-eq pew gage ,y e . m.We e6 .e m,
1 _s (thirty percent or three cesign changes, minimum, wb.icn-ever is greatar) for verification. A review by One NRC CAT inscector of :ne *CCA 'lerificaticn Log" indicated tna: cf some four-nuncrec eleven casign
- anges verified, fifty-three nac nc: teen inc:r;cra:ed in o the actual cesi- "ar" nts usec for GC inscecticn.
- - - c,-- -
Mcwever, the NRC CAT ins:ect:r bac :Sner addi-icnal
, contaggi_with :ne licensee's crogram a: :ne time of :nis assessment, i nese were as toilcws:
. With the accreaching Fuel Leac Date, anc ne num ers of H CCAs to be verified, :Pe ince:ugnness O' -he review by CCVG may be jeccardi:ec Oue : :ressures fr:m :ne ::n-struction c:mpletien schecule. i ,
i
. Precedures for the DCVG in some cases recuirec ;nysical verifica:icn by inicection of cesign inc:r:cratien. I:
das unclear as to wna was to cerform :ne recuirec ins:ec tien, a CCVG incivicual or a CC ins:ec::r. Acci icnally licensee recresentatives coulc not assure ne NRC CA-insgector nat :ne incivicuai wnc acuic ;erform :ne inscection func icn wculd satisfy tne acclica:le cualifi ca:icn requirements as cefined in ANS: Na5.2.5,"qualifi catier. Of Nuclear 2caer Olan .ns:ecti:n, Examina:icn, a. Testing Perscnnel" in ac::rcance witn ne licensee's .:IA. c mmitments.
. Unsatisfac: cry conditiens discoverec :uring :ne verifica-
-ion process were nc ::cumentec on an ':CR -~ ra-ner :n a "Recuest for Infor aticn Of Clarification;' Orm since concicions were consicered to :e cniy go:an ;ai, ,
ncnconformances. As tne ccm enent or activity may not have received :ne " final" ins:ection, :ne Itcensee statec:
"it was not considered to be a ncnc:nf:rmance."
. Occumented deficiencies relative to installa:icn were not documented en :ne Master ?unenlist for ' turnover", as ne licensee representatives incicatec cnly NCR items aere referencac On :ne master :uncalis . hese :e#'cienc'es were accliec a se:arate ":eficiency' its:, anc nan were tracked by ne CCVG.
. It a::earec from :nis review -ha- ne CC'/3 :ic not inc:uce in tneir review electrical systems :urnec over" for :es or electrical systems wnica have :een energi:sc. he NRC CAT ins:ect:r c:uld :etermine no basis wny nese sys:ams snculc te exclucec in ne samole for cesign verifica:icn.
.,O b e%
- e- me . , e,=
N* *W 84 6h a wee k
. . - - - . . . - -._ e . . - . _ ...
-z...._. . . . .
. - - . .... .. -_- .. _ . . .u.b. ., h 1, , .
/
The 'respcnsible licensee re::resen atives, during :ne discus-g sion of :nese inadecuacies, c:rnnited :: revise :ne :r: gram by the adciticn of a tracking system unica would Orreia*.e
\
- ne affe :ec ins ection Occuren:1:icn :: ne latest rela ac
_.iney furtner rei erated -heir in an; cesign cnange document. j@' i
- ensure tna; all safety-reia ec ::::cnen s and/or activi-ties have teen er will be ins:ec ac :: ne ia es; design J document to satisfy .'IRC recuirements.
At :Me -ime of this ins:ection, as nc work has been essen-tially fully cc:aleted frc :ne cesign anc ::nstr;ction stancpoint for :nese areas, :ne .'iRC CAT ins:ec:Or canne: deternine fr:m a sa= ling review wne:ner :r no: ne activity or installation was ;er# r ec :: :ne a::r vec cesign anc na: an adecuata ins:ection was :erf:r ed :: :nis at:r:vec desien. Ecwever, fr:m a sam ling review of d:cumenta-icn fr:m :ne " program in-clace, inclucing One final" ~C '*v ins:ecticns
=nc --a 12 =e- -aed-- ~- :s is in :ne Electrical Constr;cticn section of nis re er:
(Section II.3.3)}. Whether or no: the licensee's :re:osed revisiens to the cr: gram are acecuata :: ac::: lisn :ne desired objective in the allotec tima s:an canno: :e determinec. (4) Instrumentaticn Encineerine -
*nstr; entation installa:icn activities inc cccumen.a-icn reviewec by :ne NRC CAT inspec :r indica ec taa :nese instalia-icns nac recaivec an a: Orc:ria a ins:ecti:n :: ne .
lates cesign d:cument. Su::cr:s for instrumen ation ere usually cesignec in-place. .: rem a review f #ifteen cesign
- nange cocuments, One num:ers of cesign enanges as ::::arec
- One instalia iens cc=cletac aere c:nsiceracif less : nan
- nose in c:ner discipline areas reviewec.
Mcwever, adecuate engineering reviews :: the design cnanges were usually mace aftar installa:icn, similar :ne Or cass described relative to c:ner installatien activities ciscussac
- nus f ar in :ne recort.
(5) 3icine Su::cr- Encineerine-Two c:n:rac: :i:e cesign grou:s ( ~. -Geinnet anc 175::, an:
- ne si a 31:e Su::cr; Engineering (?SE) 3rcu , were res:en-sicle for :ne cesign of large :cre ci:e su::cr:3. These nree greucs :recare a :esign f:r eacn i:e su :crt. This -:es ;n is nen incor: ora:ac in:: :ne '3rewn anc Rco: Hanger Craw 1ngs (SRMLs), wnich are Me :rawings usec for instaliation
- ur:cses. f ne Oi:e sue:cr instalia:fcn :erecnnei ce ar-Cine na a su::cr. : anne :e instailec as :ssignec, ?SE field engineers are nc-i'iec anc taxe enanges as necessary ::
l i precuce a :esign -ha: can :e installec.
.;a., -c.
i e N-**63 e ei. e. io g .g p,
When the pice and scme of its supports have been installed, l the quality Con:rci program star s its "as-buil " inspectica, ! cccumenting the as-built dimensions of the pice and installed pice su:ccr:s. The drawings for the Oi:e anc ci:e supccr:s are revisec :o reflect the as-built c:nfigura:icns, anc are stamped "as-built verif f ec.' When a significant pcr:icn of
- ne succorts en :ne length of :i;e have been 'as-built
. verifiec,' a ;acxage is assem led and foruar ec for a reli-minary stress analysis.
, A stress analysis is :erformed :: cetermine 1ew stresses in the pi:e and new loacs en pi:e supper:3. :# :ne design recuirements are accrocria ely satisfiec, ne rawings are Onen stam:ed "venace certifisc" (7CC).
.cr small bcre piping designs, the pipe su::cr: cesign activity commences wnen small bcre piping is installed and is designed en lccation. PSE then issues :ne small bcre su::cr drawing. !nstallaticn nan ;receecs, wi-b necessary cnanges performed as recuirec. After inc r:craticn :f :nese cnanges, the crawing is "as-buil:" reviewed and a Cesign Review Crawing (ORO) is then issuec.
Class V hangers are "vencer certifiec" :nij nen a stress preciam becomes evicen: in :ne performance of :ne stress analysis. Otherwise, Class Y hangers receive a review ecuivalent :: ne :riginal design and inc:r:cra-icn of tne Ia:est change :c :ne :esign. Cesign cnanges race : any Of these designs are ;recessac via CMCs. Acpr0ximately seventy 'ive CMCs nere reviewec :: program requirements cy ne NRC CAT ins;ec:Or. The review of :nese CMCs anc inspection cccumen;ation in :ne ASME area by -Me NRC CAT .
.g, - - -- ~
< s s
/,
;'< /'a bl _ _ . . ---...... .nese Oractices 20 n0 Or:vice incentives to the crafts to preterly construct in strict ace:rdance with the design cccument.
(a) NRC Sulletin 79 '.a Encineerdne Walke:wn This CPSE3 :r:gesm =as c:::arec :: :ne recuiremen s accressac in the NRC Bulletin 79-14, ' Seismic Analysis f:r As-3cil: Safety-Relatec :i:ing Systems.' C?SE3 engineer'ng :r: gram recuirements were aisc :efinec in CP-E: 2.5-1, Rev. 3,
*3eneral ?regram f:r As-Guilt 71:ing Verifica-i:n." Ciscus-
, siens wita the rescensible if tensee engineering :erscnnel indicatec that a cetailec 'as-cuil " ins:ection is :einc
- erformec anc is :eing analy:ec in accercance 4i n ne ~
engineering recuirement: Of :ne :ulietin. IX-7 g gy .- 4 gw W+ eWp 6 . .- *
. . - . - . . . . ~ - . . . - . . . . . . . . . . . . . . . . . . . . . 2.. . > . .
i The respcnsible engineering represen 2:ive for the '79-la" program was intarviewed, and One actual =cchanics of the walkdcwn were discussed in de: ail. Sasically, the cr: gram recuirements regarcing Iarge bcre hangers were ciscussed. Also, ten selectac calcula:ica pacxages uere reviewec rela-tive to -his ype of ci;e su::ce . Acpr:ximately fif y-five ;e..r:a_n: of :h. e 79-la" veri. fica:icn prcgram was c:mele:ac. .c r .. .e. : ..n, u 2. .,. ve ry sma n i, num:er of
- packages" have ccmoletec :ne ecuirec reviews and are considered " final." Tnus, i: nas nc :ossible for -he NRC CAT ins:ec: r :: cetarmine 'r:m a samcling review ar.e:ner :r not tne activity or instalia:icn was ;erformec :: ne a:-
provec ' final
- cesien anc :na: an acecuate ins ecti:n was performed :: this aparavec
- final" cesign.
(b) VCD procram 'ialkcewn tv CC Inscec::rs
- rem this review , i was cifficui: f:r :ne 1RC CAT ins:ec::e
- determine ne -hce:ugnness Inc acecuacy cf this wai'<d wn 1 and/or ins ection. In:er:reta:icns of ne Or:cecural
" requirements by licensee :ersonnel ranged fr:m a detailac , instection of many attribu:es : 1 "l: cation-cnly ' walk::wn.
Addi-icnally, the NRC ' CAT ins:ector ::uic .ct de:armine wnen a ccmponent was " final" ins ectac. The NRC CAT inspector c:nsider; :na: these ins:ac: dens :c ne: necessarily provide adecuata assurance : at :ne elements /
- meccents have been instaliac :: recuiremants. As in:i:1:ac in Me Mecnanical C nstruction Sec-ion (Sec-ion ::!.5.2'; Of this re crt, numerous exam les of nar: ware acca::ac by VCC dalk cwn were found no: :: c:=ferm :: :esign :y :ne NRC CA-inspectors, the ASME Authori:ad Nuclear :nscector ( AN ), anc QC during subsequen ins ections.
-- a c:ntr : nc :resen: :n ne ciner
..sw.r... .. .eae.. ..cansee re resentatives again stated
- heir intent satisfy :ne necassary NRC recuirements, Inc also ex:ressec confidence in :neir VCC or: gram.
I (5) Picinc anc vecnanical Encineerire 7 *eview ne cesign :nanga c:n r :r: cess in nis area, l ac:r0ximately fi' y CCAs/CMCs were reviewec. Seve-ai trave-l ler anc ins allation :acxages ware c:m arec :: ne :r: gram recuiremen:s f:r ci:ing instalia-icns. As re#erencac ir :nts recc'rt, few ceficiencies in :ne review of safety sys am
; icing anc ASME C:ce -acicgra:ny'f:r nis : icing were ':'n: e by the NRC CAT ins:ect:rs wi:nin :ne narcware areas. ew
. cesign enanges were f:unc relevant 0 nis area.
Several raveier anc ins;ailaticn :acxages were eviewec ::
- he Or: gram recuirements f:r ecnanical i . alia-icn.
.4e.( * ,
.-w
, ..r . . . . . . . . .. a . . -
.& ~
; ;c'i- s Similar caservations relevan: to :ne processing of cesign
, enanges and cocumentation of nonconformances were made.
1 These were discussed in prevdcus paragr:cns of this secticn
.' cf the report as to electrical ecuitment, electrical suo-a ;cr:s, anc pice hanger /sucpert installa:icns. Discrecancies relevan: to ins:ection pr:cecures are ciscussac in :ne Mechanical Constructicn Section (Section III) of :nis report.
- c. Summarv C:mments C5ncerninc ne Cesien Chance 3 recess The design change ; recess at C?SES is c molex, anc a :imes, cumcerseme. The NRC CAT ins:ec: r's review Of cesign :nange pro-cesses in tne varicus cisciclines revealed a design cnange
- r: gram with c:n:rels inc:r:cra ac uncer a " design-c:nstruc -
cesign reviewd pnilosceny. ints :nitoscany resulted in a large numcer of dest;n enanges and a re:etitive inscection : recess.
,. (NOTE: There are accroximately 70,000 CMCs and 15,C00 CCAs that have been issued. This numcer dces not incluce .evisions).
Altnough this cesign : Mange cr: cess may :e cifficult, there is nc:ning in NRC recuiremen s a :isc:urage er cronici: :ne use Of sucn a system. In general, cesign enange c0n:rois a: CFSE5 - satisfied the applicant's FSAR c:mmi men s anc :ne ANSI standarc recuirements. Mcwever, witn this :ype of sys;am in-clace, ac ual verification of hanger, succort, electrical, anc mecnanical equipment installaticns o the accro:riate cesign recuirements cannot :e performec until *wert activities" nave been ::: letad. Few, if any, installatiens c:uld be verifiec as few nave teen cesignated as c:moleted under the licensee's contex: of "c:moletion." Thus, the final acecuacy cf :nese :en r Is c:uld not :e determinea ly tne NRC CAT ins:ec cr.
- 2. Car-scrive ac tion Systems T'e C?SE3 FSAR recuires a ncnc:nformance rec 0r :e "u-ili:ec f r :ne icentification, cocumentation, cis:csitioning, anc verifica:icn of ceficiencies in ::aracteristics, d:cumentation, er Or cecures wnica rencer :ne cuality of an i am unac:actacle or inceterminated . The FSAR also recuires cr cacures for trencing of nonconfermance re:cr:s to identify trends acverse :: cuality anc for :ne initiatien of corrective acticn requests for significan; anc rece:itive ncnconfer-mances.
CF-E?-15.1, " Precessing Ncnconformance Recerts and ne a::lica:!e aeferences 4ere imolementec at ne si a 0 feet :nese recuirements
- satisfy ne licensee's FSAR ::mmitments.
A review of accr xima ely :ne-hurcrec selectad ncnc nf:rmance recor : (NCRs) indicatec :na: icentified cenconformanics ::nci-icns (documentec en NCRs) were cis csitionec anc ;recessac in ac:Or ance witn procacural recuirements. . IX-g
E,,,
.-,......,.a..-
c - . . ... t As discussac in Secticns II, III, and IX of :nis repor., overall findings indicate numercus instances in the electrical and recnanical areas where ncnc:nf:mances were identified. hwever, varicus metnces (e.g., ;unchlists, ins:ecticn re cris, vercal, anc c .ner inicrmal methcds) were usec o adcress and rescive :nese n0ncenicmances, previcing no collective evidence cf a::r: grit e c:rrective action and/or Justifica icn. a.cciticnally, :ne .1RC CAT inscec:ces ciscoverec :na: -he .Meenanical/ Civil OA/CC Su:eritsce f rectec his su ervisors o c0cu en: ncnconfor?.ing conditiens on an unsatisfact:ry Ins:ec:icn Recer. (IR) only, c:ntrarf to ne licensee's F3AR c =rait=ents anc OA :r:grara recuire.eents. t w. e A e f
- s y.
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- X-10 7... .
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- 3w.
f Rf GT) AP ENDIX O
~
U. S. NUCLEAR REGULATORY C0!@il5SION REGION IV NRC Inspection Report: dC/?t? 50-446/83-15 Docket: 50-445 Ca tegory: A2 50-446 Licensee: Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas , Texas, 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 Inspection At: Comanche Peak, Units 1 and 2, Glen Rose, Texas Inspection Conducted: March through July 1983 Inspectors: bhj Sf 9//7/83 R. G. Taylor, Senior Resident Inspector Da te / Construction (SRIC) Approved: 2Y) .,h# D. M. Hunnicutt, Chief 3//S/B3 Date ' Reactor Project Section A Insoection Sumary Insoection Conducted March throuch July 1983 (Recort 50-445/83-24 and 83-446/83-15) Areas Insoected: Special inspections, announced and unannounced, related to allegations made to various NRC persons including the Atomic Safety and Licensing Board in their procedings regarding the operating license for Comanche Peak Station. The inspections involved 449 inspector-hours by one NRC inspector. Results: The inspection confirmed the need to issue four violations initially identified by the Construction Appraisal Team (CAT) (NRC Inspection Report 50-445/83-18; 50-446/83-12). These involved the areas of HVAC, Equipment Installation, Document Control, and Storage of Equipment. F0IA- 5-59 wee #EE
2
. Details
- 1. . Persons Contacted Principal Licensee Employees
*R. G. Tolson, Site QA Supervisor
*C. T. Brandt, Non-ASME QC Supervisor
*J. R. Merritt, Engineering, Construction and Startup tianager
*J. B. . George, Project General Iianger'
*D. N. Chapman, QA Manager
*B. R. Clements, Vice-President, Nuclear Brown & Root (B&R)
*G. R. Purdy,. Project QA Manager
*D. Frankum, Construction Project l'.anager The SRIC also interviewed many other licensee, B&R, and subcontractor personnel during the course of the inspection.
- Denotes those persons who attended one or more management interviews with the SRIC.
- 2. Licensee Action on Previous Inspection Findinas (Closed) Unresolved Item (50-445/82-22-02), " Analysis of Weld Discrepancies."
This unresolved item concerned a substantial number of identified defects in a large whip restraint essentially surrounding the mainsteam and feed water lines located several feet outside of the ASt1E code boundry point.
. The device was engineered by the licensee's A/E and manufactured by NPS Industries. Due to the overall size of the structure, it has been nick-named " George Washington Bridge" by the site labor and quality forces. The licensee had reported the finding of the defects as a potential 50.55(e) item to the SRIC on September 30, 1982, which was subsequently stated not reportable in a letter dated December 27, 1982. An NRC inspector followed up on the matter during a visit to the offices of the A/E, as documented in NRC Inspection Report 50-445/83-12. This review pertained to all of the defects involved with the exception of two cracked welds that had not been analyzed at the time of the inspection. The engineer has recently analyzed these two defects and has detennined that had they not been detected, the structure could have fulfilled it's function. The SRIC has reviewed the location of the cracks and their length in relation to the size of the welds and the functional application of -the structure. Since the structure has no continuous service application and is essentially subject to a one-time loading, the cracks would not have the potential for further propagation.
Further, the cracks are at points in the structure that would receive rela-tively low stresses in the one-time impact based on their small size in relation to the members being welded. It appears that the cracks formed due to the stresses developed during the tightening of high strength bolting in
3 the imediate vicinity of the welds during the site assembly of the structure. Taken in conjunction with the earlier documented review of the engineers calculations and the SRIC's review of these cracks, the SRIC has concluded that the engineer's overall analysis was adequate and that deficiency (s) were not reportable under 50.55(e). Both the licensee's initial report (CP-S2-12) and the above identified unresolved item are considered closed. It should be noted for the record that this closure only applies to the reportability aspects under 50.55(e) and not to the correction of the defects. The defects, including the cracks, have been documented on a nonconformance report. The final disposition and closure of the NCR will be evaluated during future routine inspections.
- 3. Review of Licensee Self-Evaluation (Usino INP0 Criteria)
The SRIC has reviewed a report of the licensee's self- evaluation performed during October 1982 which was based on criteria that has been developed for the purpose by INP0. The evaluation was perfomed in behalf of the licen-see by personnel in the employment of Sargent & Lundy, an architect-engineer fim with substantial nuclear power involvement. /, copy of the report was furnished to the NRC, and subsequently, to the Atomic Safety and Licensing Board in the matter of Comanche Peak Station operating license by letter dated May 2,1983. The purpose of the review by the SRIC was to detemine if any of the 47 findings in the report were of a type and of sufficient significance to have been reported to the NRC as required by 10 CFR 50.55(e). The SRIC reviewed each of the 47 findings and the supporting documentation in the report pertaining to each finding. This review revealed that none Of the 47 items were based upon identified deficiencies in structures, systems, or components nor were there any significant defici,encies in design, engineering, or testing that would constitute conditions reportable under 10 CFR 50.55(e).
- 4. Car Wash In Containment During the limited appearance statement portion of the Atomic Safety and Licensing Board hearing on May 16, 1983, a person stated at transcript page 6152 that he understood that the containment looked something like a car wash. The person stated that it was his understanding that the situa-tion developed at about the same time that there was a meeting at the 0/FW Airport between the NRC and any interested parties to discuss NRC decen-tralization. That meeting took place on April 5,1983. For the purposes of evaluating this allegation, the SRIC expanded the period of interest to include the 3 weeks prior to the meeting. During.this entire period, the Unit 1 reactor system was undergoing what is referred to as " Hot Func-tional Testing". This particular test is an accurate simulation of the operation of the reactor system and its appurtenances but without a reactor core being in place. The heat and pressure in the system is generated by the reactor coolant pumps in conjunction with the chemical and volume con-trol system charging pumps. The test could readily be construed to be a pressure test but in fact is an operational test at pressure. This parti-cular test extended overall for about 90 days beginning late in February
4 and continuing until late t'ay. The SRIC monitored the test but was by no means continously in the containment. The SRIC interviewed personnel in the licensee's startup test group, QC inspectors who had reason to be in the building and others to obtain a picture of the events that occurred in the Unit 1 Cont (tinment Building during the period of interest. The SRIO also reviewed the licensee's control room logs for any indication of oper-ational problems indicative of a major leak in any of the fluid filled systems under test. The picture obtained was that there were several small leaks, generally at the gaskets between valve bodies and their bonnets. In l addition, there was a considerable amount of condensation dripping from the l reactor coolant pump motor. cooling coils. This was caused by the cold water l in the coils condensing the humidity from the atmosphere within the building l and was not indicative of a leak in the reactor coolant system. The SRIO found from the control room logs that on March 29, a steam leak occurred during one phase of _ the test when a drain valve was partially open. Perhaps this valve should have remained closed. The room in which the valve was located was apparently filled with steam vapor which would have condensed out on the cooler walls as water. On March 30, the reactor vessel head vent valves were partially opened, which in turn would give some amount of steam blowoff into the reactor refueling cavity area and would rise up into
. the building until cooled and condensed out as water. .None of these events .
' are typical of any major leak indicative of piping or piping component (such as a valve) failure. The type of small events described above are,
;- within the experience of the SRIC, typical of what would be expected during l such a test and is one of the reasons for performing the test.
- 5. Desion of the HVAC System Suocorts By letters, both dated March 11, 1983, Citizens Association for Sound Energy (CASE) notified the NRC's Offices of Inspection and Enf n ement and the Executive I.egal Director of a concern that the HVAC system for Comanche Peak had not been properly supported, nor had it been properly considered in regard to seismic load conditions or its treatment as potential mis-s iles. CASE specifically states that from their review of the FSAR, it appears that the licensee has not analyzed the HVAC supports for a seismic load condition. Specific reference is made to Sheet 21 of Table 17A.
In addition, the personal observations of Messrs. Walsh and Doyle are relied upon to point out that there are no lateral supports on the HVAC systems within the containment. CASE also states that all HVAC components and supports inside containment should be treated as missiles under Cri-terion 4 of the General Design Criteria for Nuclear Power Plants, 10 CFR 50, Appendix A. i-Sheet 21 of Table 17A of the FSAR lists the containment ventilation sys-tems as being Seismic Category II. Apparently, it has been assumed by CASE that this category excludes seismic loading in the design. This assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seismic Category II as being those portions of systems or components whose
5 continued function is not required but whose failure could. reduce the func-tioning of any Seismic Category I system or component required to satisfy the requirements of C.I. A through C.1.Q of Regulatory Guide 1.29 to an unacceptable safety level or could result in incapacitating injury to occupants of the control room. These systems are designated Non-Nuclear Safety (NNS) Seismic Categor a safe shutdown earthquakeSSE) (y II and are cause will not designed suchand constructed so that a failure. CASE also states that if the HVAC systems within the containment failed during a SSE, this would allow the temperature within the containment to rise quickly to unacceptable levels ~which could over time cause compon-ents and monitoring equipment to fail and which could also mean that it might be impossible for wcrkers to enter the containment due to the heat. Containment heat removal is required by Criterion 38 of the General Design Criteria for Nuclear Power Plants. The system to remove heat from the reactor containment at Comanche Peak does not rely on the HVAC system but rather is composed of two separate containment spray recirculation trains each with 100 percent capacity. Each train contains two separate pumps, one heat exhanger, and seven spray headers, and each system is fed from its individual electrical Class IE bus. The containment heat removal system is designed to ensure that the failure of any single active compon-ent, assuning the availability of either onsite or offsite power exclusively, ; does not prevent the system from accomplishing its planned safety function. I CASE's concern with being able to enter the containment following certain design basis accidents is unfounded in that it is not a requirement. In order to assess the adequacy of the design of HVAC supports, an inspec-tion was conducted at the home office of " Corporate Consulting & Develop-ment Company, LTD. ." the support design consul tant. It was determined that all permanent HVAC supports are analyzed for seismic loading. Two methods are utilized: Zero Peak Accleration (ZPA), or 1.5 Times the Peak Accelera-tion When the Fundamental Frequency Falls Below 20 Hertz. Of the latter method of design, only about 6 out of 4000 supports have been designed that way. A typical HVAC duct run is supported axially at every third support This may explain why Messrs. Walsh and Doyle may have felt that there were no lateral supports on the HVAC systems. The NRC inspector reviewed the design of a typical HVAC duct run at elevation 852'-6" in the Auxiliary Building. Supports were designed utilizing two computer programs entitled FEASA-2D and FEASA-3D. The acronym stands for frame eigenvalue and stress analysis. The -2D version is used on the transverse supports and the -3D version is used on the axial supports. The inclusion of equivalent weights from both up and downstream transverse supports and accesaries such as vol-uma dampers and vane turns in the design of the axial supports was verified. This inspection verified the adequacy of the siesmic design techniques being utilized for the design of HVAC supports at Comanche Peak. The concerns expressed by CASE have been found to be without merit. Persons contacted during the course of the inspection at Corporate Consulting
[ 6
& Development Company, LTD. were:
J. Roland Yow, Presid9nt & Chief Executive Officer Gary Hughes. Vice-President for Operhtions David 1.indlay, Principal Engineer Stephen Lehrman, Seismic D90artment Manager Daryl Hughes , Project' Engineer ,
- 6. Heating, Ventilation, and Air Conditionino System (HVAC)
During the CAT inspection (NRC Inspection p.eport 50-45/83-18; 50-446/83-12), the CAT inspectors noted that a significant portion of the welds on the ducting support structures were deficient in relation to the applicable weloing code requirements. The dominate deficient cohdition noted was that the welds were significantly undersized. Based upon this information the SRIC toured various areas of the facility with special emphasis on the ducting in the Unit 2 Containment Building since that was one of the more recent areas of installation by the HVAC ' contractor. In accordance with the design drawings, the bulk of the welds should have been fillet welds with binch leg size. The SRIC noted by visual comparison to the hinch thick base metal that very few of the welds were of proper size. The CAT inspectors also found cases where the bolting and gaskets between ducting sections were loose and/or missing. The CAT inspectors also found that some cupport members were not within the dimensional ~ tolerances on the design drawings. It was noted that the contractor's inspection records did not reveal these various facts , indicating ineffectual QC by the contractor. Further, a review of the licensee's audit program indicated that the . licensee was unaware of these several problems in the fabrication, installation, ar.d inspection of the HVAC systems. Based upon the CAT inspectors' findings and his own observations, the SRIC recommended that a notice of violation be issued to the licensee pertaining collectively
-to these matters (Notice of Violation issued on May 31, 1983.
Reference 50-445/83-18 and 50-446/83-12, item 4).
- 7. Installation of Major Items of Eouipment The CAT inspectors noted during their inspections of certain major items of equipdent that there were several variables in how the equipment was fastened to the building equipment pads; In some instances, tanks for example, CAT inspectors found that there were two nuts (double nuts) on the embedded bolts securing the equipment, other bolts had one nut, (single nut)
CAT personne Ea so note tnat certain heat exchangers had slotted holes in one of the mounting bases to allow for thermal expansion during operation. The holddown nuts appeared to.be installed too tightly and may have prevented freedom of movement. The SRIC obtained- the design and installation drawings for two of the referenced heat exchangers identified in the CAT report. Both were found to be horizontal Utube heat exchangers whose function is nonsafety, but whose pressure boundary in the tubes is safety-related since the process fluid could be radioactive. The SRIC found that the construction drawings for the mounting pedestals had a flat steel plate on one
7 pedestal that would be suitable for the type of mounting detail on these heat exchangers. The SRIC then reviewed the installation travelers for each heat exchanger and found that these documents did not note or address the slotted details, the plate, or the fact the bolt should be left loose. The SRIC would note that the vendor manual which provides the details does not provide information on how loose or tight the nuts should be nor how these nuts are to be locked at that looseness or some torque value. The SRIC with the assistance of site QC and craft labor had one of six nuts loosened on heat exchanger TCX-CSAHLD-01. On all six of the studs involved, each had only one nut (single nut). The one nut that was loosened had been very tight, as evidenced by the amount of force required to break the nut loose. On another heat exchanger of comparable design, it was found tha > and when the top nut was loosened, the second nut was approximate y one flat (about 1/6 of a turn) from being fully tight. This degree of looseness should allow sufficient freedom of movement. During the document review, the SRIC found that the engineer had specified that all rotating and vibrating equipment should be double nutted and that other equipment could be secured with only one nut. No document could be located that established the identity of vibrating equipment
@l overa a vio 1on o
. n1s was cons 1dered riterion V of Appendix B to 10 CFR 50 (Notice of Violation was issued on May 31, 1983. Reference : Notice of Violation 50-445/83-18 and 50-446/83-12, item 1).
- 8. Maintenance of Eouipment In Outdoor Storace Areas The CAT found that a considerable amount of equipment such as pipe support struts , clamps , and like items, normally stored outdoors ,
was not being properly maintained in accordance with procedure MCP-10,
" Storage and Storage Maintenance of Mechanical and Electrical Equipment", as evidenced by rusting bolts and adjustment screws on s truts . In addition, the strut bearings were dirty from dust and l the bearing load pins, in some instances, were rusted. By a tour of the storage areas , the SRIC confirmed the CAT inspectors find-ings. The SRIC would also note that the INP0 Self-Evaluation l
Report at page 111 describes essentially the sage finding. This l situation was determined to be a violation of Criterion XIII of l Appendix B to 10 CFR 50 (Notice of Violation issued on May 31, 1983.
Reference:
Notice of Violation 50-445/83-18 and 50-446/83-12, item 2). The SRIC would note for the record that there is little evidence that any items which indicated substantial deterioration from such storage conditions have in fact been installed in the nuclear power block. It would appear that the various items involved have been cleaned and restored prior to installation such that they can perform the required function.
- 9. Obsolete and/or Illeoible Drawings In The Field The CAT inspectors n one particular area Tha SRIC discussed
8 the finding with supervisory personnel of the licensee's central document control center who indicated that they had located the drawings identified by the CAT inspectors along with many more that were obsolete in other areas. It was stated that distribution system for engineering drawings had become faulted by the simple volume and by the need for so many points of distribution and audit verification thereof. Since problems are obviously still present, it was detemined that the licensee had violated Criterion VI of Appendix B to 10 CFR 50 (Notice of Violation was issued on May 31, 1983.
Reference:
Notice of Violation 50-445/83-18 and 50-446/83-12, item 3) and that substantial steps would be required to correct the problems.
- 10. Alleaations Relative To Improperly Supported Items In The Control Room The president of CASE in a letter dated March 11, 1983, addressed to Mr. Richard C. DeYoung, Director of the NRC Office of Inspection and Enforce-ment, indicated that CASE had received infomation from an unidentified sourci to the effect that:
- a. There is field run conduit above the control room supported only by wire.
- b. There is drywall (or sheet rock) that is supported by wire.
- c. There may be lights that are supported by wire.
The SRIC has examined the suspended ceiling and the area above the sus-pended ceiling in the control room area and has examined the pertinent engineering drawings depicting both in relation to these allegations with the following findings:
- a. There is a considerable amount of both safety-related and nonsafety related conduit in the area above the suspended ceiling. The safety-related conduit is supported by Seismic Category I supports typical of those used in other areas of the facility. The nonsafety-related conduits are generally supported by simpler and less substantial sup-ports that are typical of those that the SRIC has observed in large open factories and are not designed to seismic standards. In each case examined, the non-seismic support was structurally paralleled with a small stainless steel cable that would assume the full weight of the conduit were the nomal support to fail in a seismic event.
- b. The drywall materials were found to be part of the suspended ceiling above the central part of the control room and to form a part of the sloping wall area below the control room observation room. These dry-wall materials have been securely fastened to a metal frame work (metal batten) which in turn is supported by conventional and non-seismic straps and wires to the concrete primary building. The frame work is also attached to a system of stainless steel cables which in turn also attach to the primary structure such that if normal sup-ports fail during a seismic event, the weight of the framing and drywall will be assumed by the cabling thus preventing the materials from falling.
9
- c. The lighting fixtures in the control room are supported from an intermediate substructure of "unistrut" by light-weight conduit.
The substructure is likewise supported by the same type of conduit from the primary structure ceiling. The conduit used appears to be the typical of that supporting the light fixtures in most offices with suspended ceilings. Paralled with each conduit are two small stainless steel cables which would assume the load if the conduit or its attachment were to fail. In the case of the actual light fixtures, the cable is attached to the light fixture at the edge of the reflector assembly. The SRIC would note for the record that above described design features appear to fully satisfy the intent of the licensee's commitment to comply with NRC Regulatory Guide 1.29, " Seismic Design Classification." The licensee has used terminology in the classification system that is at variance with that of the regulatory guide but is explained and defined in Section 3.2 of the FSAR. In essence, the licensee has defined all safety-related items that must remain fully functional during and after a seismic event as Seismic Category I. Items not having a safety function but whose failure could damage components which have a safety function or cause injury to the occupants of the control room during an event are referred tc as Seismic Category II. In the case of the items involved in this allegation, all are Seismic Category II since their falling could
. cause injury to the control operators. The cabling system described can be expected to prevent such a fall even though the normal supports could possibly fail. The stainless steel cable used in this design feature, which at a short distance away looks much like bright galvanized common steel wire, is of relatively high strength. As an example, the test strength of an 1/8-inch cable is in excess of 1760 pounds. With four cables attached to a light fixture, two at each end, the total support capability of the cables is over 7000 pounds. It is apparent that the designers have elected to use conventional suspended ceiling and light fixture support techniques in order to use conventional and available materials and then provide a high strength backup support system in a seismic event.
No violations or deviations were ident'ified during this special inspection effort.
- 11. Placement and Curino of Concrete Durino Freezina Weather
~
During the limited public appearance portion of the Atomic Safety and Licensing Board (Board) hearing conducted on May 15, 1983, there were two references to the placing of concrete in freezing weather at the Comanche Peak Station which in turn lead to a question from the Board to the NRC transcript while the Board question is at 6109. Also at 6109, an uni-dentified voice responded to the Board that the matter had been reported in IE inspection reports. Research of the NRC inspection reoorts revealed that there had been such a discussion in NRC Inspection Report 50-445/77-01 which was categorized as an unresolved item pending the licensee's review and action on their finding of the problem. The unresolved item was further discussed in NRC Inspection Report 50-445/77-04 with the closure of the item by an improvement in the QA procedures.
10 The SRIC has reviewed the matter, particularily with a view toward deter-mining whether the practices involved actually caused damage to the concrete involved. The primary focus of NRC Inspection Report 50-445/77-01 (Details II, paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had been prepared approximately 2 weeks earlier than the inspection period covered by the inspection report. The first of the licensee's reports (C-134-77) was directed specifically to findings by a licensee inspector that the surface temperature of Concrete Placement 101-2808-001 some 6 hours after the placement was completed were well below freezing in some locations. The.other licensee report (C-135-77) was directed toward records and was not considered in this review. The 'SRIC obtained the necessary records to review the matter and found that placement 101-2808-001 had taken place on December 30, 1976, being completed at approximately 6:00 p.m.
. 1.ater, the same evening at approximately midnight, the licensee inspector found that some surface areas were chilled to as low as 200F. The records reflect, however, that there was disagreement between the S&R inspection personnel assigned to monitoring the curing of the placement and the licensee's inspector as to what the surface temperatures actually were.
The B&R personnel contended that the licensee inspector was actally mea-suring the air temperature rather than the temperature of the concrete. No resolution of that disagreement was reflected in the records. The SRIC interviewed the licensee inspector df record during the course of this review to gain a clearer understanding of the events which took place. The licensee inspector stated during the interview that he was confident that his measurements were accurate and also stated that there was no phy-sical evidence that the concrete was frozen even though the surface temperatures were well below freezing. The records also reflect that in order to resolve the issue, swiss hamer tests were run on the suspect areas after the concrete had fully cured. These tests indicated that the suspect areas had attained strengths comparable to known properly cured areas, indicating that the concrete had not been damaged even though the possibility exists that it had been frozen for a period of time. The records reflect that good concrete curing temperatures, i.e., above 40oF l were established and maintained shortly after the licensee's inspector's observation. For the record, the SRIC would note that Placement 101-2801-001 took place in the Unit 1 Reactor Building. The placement became the open area floor
- at the lowest full floor in the building. This floor area, while suppor-ting some equipment, serves primarily as a walk area. As such, it is fully topped with an architural' concrete making the structural concrete no longer r accessable.
NRC Inspection Report' 50-445/77-01 also discussed comparable events to that documented on Surveillance Report C-135-77. One of these events was docu-mented by Surveillance Report C-068-76 on January 7, 1976, and on B&R deficiency / disposition reports (now titled nonconfonnance reports). These documents indicate that on January 7,1976, the surface temperature of Placement 105-2773-001, the foundation basemat for the Unit 1 Safeguards Building, were found frozen as evidenced by frozen wet burlap over certain areas that were not covered by insulating blankets. The records also l L
=
. l 11 reveal that the reported finding took place almost 7 days after the place-ment of the concrete. Although the placement should not have been allowed to freeze in the time frame involved in accordance with the project speci-fication, the placement was accepted "use-as-is" on the premise that the curing temperatures during the 7 days were conducive to a good cure and that after 7 days there would be little free water in the concrete to freeze even though the burlap was froze. This conclusion is considered valid by the SRIC based on his review of publications of the American Concrete Institute and the Bureau of Reclamation. Further, in responding to a separate finding that the field cure test cylinders made for the placement tested lower than allowed by the project specificatioris, swiss hammer tests were perfonned. The swiss haniner tests indicated the concrete placement had full specified s trength . Relative to the low reported strengths of the field cure cylin-ders, the SRIC would note that in his experience field cure cylinders will frequently test low under cold weather conditions. The reason is that the cylinders' small mass generates little heat of hydration, thus making them either more vulnerable to freezing and/or curing much slower than normal due to their depressed temperature. The final events covered by NRC Inspection Report 50-445/77-01 included DDR-C-460 which in turn discussed low temperatures during the curing per- ' iod of three separate placements that were made during the late December time period of 1976. In each case, the records reflect that the placeme'nts were accepted "use-as-is" since the least amount of cure time was 9 days, again with good conditions until the cold weather occurred. The NRC inspector involved in NRC Inspection ' Report 50-445/77-04 which closed the unresolved issue has stated that he had visually inspected each of the placements discussed in NRC Inspection Report 50-445/77-01 for evidence of damaged concrete and found none. NRC Inspection Report 50-445/77-04 did not reflect those inspections since the NRC inspector was aware that the concern was for prevention of repetition rather than any specific concern about the quality of the placements involved. The SRIC would note for the record that there are no regulatory or industry prohibitions on placing concrete in cold weather conditions. The American Concrete Institute and the Bureau of Reclamation both indicate that if the fresh concrete is above 400 F at the time of placement, the chemical process of hydration will generate sufficient heat to prevent the concrete from freezing provided that precautions are taken to prevent heat loss. In mass concrete applications, the greatest danger to the concrete is on the exposed surface areas, particularily at corners and other edges of the placement. It would be exceedingly rare for the mass of the concrete to freeze and sustain damage. These publications also indicate that even if frozen, the concrete will normally cure to full design strengths if temperatures con-ducive to the hydration process are restored. 12. During April 1983, NRC personnel received allegations to the effect that
1
=
l 12 A second allegation from the same person indicated that the QA group charged with responsibility for verifying that design changes have been incorporated into the plant and that the inspection records for the installations accurately reflected that incorporation was being required with the use of a computer generated status document to make the verification of records. The allegation was that the computer list-ing was faulty and therefore, the verification effort was equally faulted. The SRIC has examined each of these al' legations as to the factualness of the allegation and as to whether the allegation has or will have an effect on the safety of the facility when operating. In regard to the M e saf t ta the as-built inspection was not developed to assure that the " Vendor Cer-tified Drawing" was an accurate representation of the support in all aspects. The as-built program was established to assure only that the support loca-tion on the supported pipe and the direction of support is accurate for the purposes of performing the final pipe stress analysis. The responsibil- i p ity for assuring that the support members and other characteristics of the
, individual" support reflect the design drawing requirements reside in other
[ QA groupsthese associated with To also perform functions in the the fabrication and installation as-built verification efforts. inspection would be a redundant inspection that would not contribute significantly to the safety
< by..lv,!. function of any given support.
i
- f .y hek:/p Regarding the the SRIC found that ut! b k#f only at the spec 1 legation was made. Whe ega- )
y[dj ic .. .. i ' tion, the alleger provided the NRC personnel with a reference to a QC 3 m.
'y//M I ,
inspection report which he said would fully display his concern. This // report, identified as IR DCV-00421, w,as found to contain notation that the verification was based on a computer tabulation and that the report was being completed at the direction of the inspector's supervisor. The original report was dated April 4,1983. The permanent file copy was found to have been marked " voided" by the originating inspector as of May 20, 1983, with a notation that the report had been superceded by IR DCV-00423. This latter inspection report was examined by the SRIC and found to document essentially the same inspection effort by the same inspector but without any notation of having been based upon a computer tabulation and' without notation of apparent protest of directions given by supervision. The SRIC interviewed the QC inspector who prepared and signed all of the reports noted above in order to ascertain what had and is transpiring in the QC design verification program effort. The inspector stated that the attempt to use the computer based data in the performance of the assigned task was in error from the beginning because of errors by persons genera-ting the computer data. The interviewee stated that only the one verifica-tion effort had been done using the computer based data 6 i
3 13. 1 procedural deviation was the one instance stated in the allegation. Dis-cussions between the group supervisor at the time the allegation was L received and the SRIC indicated that he had attempted to use the computer tabulation to expedite the task on a trial basis by management direction and that he had caused the original inspection report to be filed as it was , - to give management a picture of the faults in the computerized data. It thus appears that the design verification effort has been performed in accordance with procedures except for the one-time pertubation that was subsequent correctly reaccomplished in accordance with approved proce-dures. < ^ No violation to NRC requirements wef e revealed during this special inspection effort.
- 13. Improperly Certified Liquid Penetrant Examination Materials The CASE informed the Atomic Safety and Licensing Board by a letter dated
; May 18,1983,-of a potential problem with the liquid penetrant materials in i
use at the Comanche Peak Station. The letter stated that CASE had been made aware of the potential problem during a phone conversation with Charles A.
- Atchison, who in turn learned of the " problem" from a Dallas area represen-tative of the Magna-Flux Corporation, the orginal manufacturer of the material.
- The letter states that the problem surfaced only 7 to 10 days earlier. Based on the date of the letter, it would seem that the problem arose between approximately May 8 to May 11, 1983.
The situation bears close resemblance to the situation outlined beginning 4 with NRC Inspection Report 50-445/82-18;50-446/82-09 based upon an inspection conducted during the period of September 7-10, 1982. The NRC inspector noted J that some certified test result documents had been altered by " pen and ink" changes not immediately explainable. The matter was considered unresolved at that time. During a second inspection of the matter, conducted during i November 1982 and documented in NRC Inspection Report 50-446/82-11, the i inspector found that previous corrective actions were riot adequate and fur-ther that the " pen and ink" changes sometimes didn't match the type of material being certified. A Notice of Violation was issued as part of the . inspection report on the matter. The licensee responded to the Notice of
' Violation by a letter dated December 21, 1982, wherein he stated that a
! supplier had altered the certificates but that the original manufacturer had been able to furnish valid certificates and further, that all future } purchases would be direct from the manufacturer rather from a " middle-man" supplier. The licensee also stated that specific receiving inspection pro- ! cedures had been implemented to prevent repetition. NRC Inspection Report 50-445/83-10;50-446/83-05 documented verification that the licensee's actions j were acceptable and the matter was closed. I
- It appears that the situation outlined in the CASE letter parallels the 4 NRC findings in all details except for the dates which probably arose as a result of misunderstood or incomplete communications between the i
i I 4
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g . . . . f 14 Magna-Flux representative and Mr. Atchison and/or with CASE. CASE also posed two questions on the matter as follows:
~
- a. Has an NCR been written on this problem?
. Answer: The above discussed inspection reports document a total of five NCR's that were issued.
- b. Has either TUGC0 or Texas Utilities or B&R notified the NRC of this problem?
[ Answer: The roles of reportability were effectively reversed in that the NRC identified the problem and notified the licensee. A need for further NRC action on this matter has not been identified and ttie matter is considered closed. 14 Penetration Seals This special inspection was undertaken to ascertain the validity and sig-nificance'of allegations received initially by an .NRC Headquarters Duty Officer on or about. March 22, 1983, which were confinned and added to during a telephone interview with the alleger on March 23,~1983, by the SRIC and a NRC inspector assigned to NRC Region I. The allegations, as understood by the SRIC, were:-
- a. The overlap seal for flexible boots should be 3 inches whereas 2 inches is being used by BISCO.
- b. There maybe a problem with the strength of the fabric used in the
, flexible boots since the material supplier and BISCO are involved in a lawsuit.
- c. The aggregate used in a radiation seal may separate giving rise to improper personnel protection.
Since BISCO was and is on the Comanche Peak site installing seals, Region IV was selected for the purpose of this special inspection although the com-pany has involvement at several other nuclear power sites throughout the United States. -The SRIC obtained from the BISCO site manager all of the production and quality procedures applicable to the work at CPSES as well . as some that are not. The alleger specifically mentioned that the NRC ! should review Procedures QC-507, SP-504, SP-505, SP-505-1, and SP-505-2 in regard to the flexible boot overlap problem. Each of the above procedures was in the books offered to the SRIC for review. A brief discussion fol-lows as to the contents of these procedures:
- a. QCP-507: This procedure covers the final inspection of installed s , . ,-- n' . 4 ,, ,, - - , - ,,,,.r-,,, n , n ,.- .-- - - - ..~, , - - -----, , , -,----
15 l I flexible boots. The amount of overlap is not mentioned in the procedure, although the procedure does require that the , seam be examined for evidence of poor sealing such as " fish-mouthing" which is taken to mean that the exposed edge of the overlap is puckered and not adhering to the base fabric.
- b. SP-504: This procedure provides instructions and a calculation sheet to initially cut the fabric into a shape that would subse-quently allow the fomation of a truncated cone. The formula on the calculation sheet requires that 1-inch be added at each edge of the fan shaped fabric which is evidently to pro-vide the overlap. The base formula prior to adding the 1-inch provides a dimension just equal to the circumference 1 -
of the pipe and/or sleeve to which the boot will be attached. Thus, the 1-inch at each edge will provide for 2-inches of overlap, assuming that the pipe and sleeve are concentric. If pipe and sleeve are not concentric, the resulting cone will be skewed and the seam overlap will be something other than 2-inches. 1
- c. SP-505: This is a generic procedure for the installation of flex-ible boots. "It was noted that the procedure requires that the adhesive for the overlap seam be spread over a 3-inch depth from the fabric edge prior to fitting up the fabric where it is to.be installed. Although no.t so stated, it
-appears that the 3-inch width of adhesive is to provide sufficient area of adhesive in the event the above men- '
tiened cone skewing occurs.
- d. SP-505-1 and SP-505-2: These are additions to SP-505 having appli-cation when the boots are used as a simple pressure seal only and for when the boot is used as part of a fire pro-i tection seal, respectively.
The SRIC interviewed the BISCO site manager as to whether the procedures had ever required a 3-inch overlap. The site manager indicated that 3-inch seam had been used up to sometime in 1979 and that his homeoffice engin-eering had then changed the seal seam detail. The SRIC reviewed the results of a pressure differential test performed by BISCO-in September 1979 which indicated that the fabric boot would withstand a differential pressure
~ of 44 psig without sustaining damage. The project specification '(2323-t15-38F) requires that the pressure seal maintain its integrity only up to 2 psig.
While the BISCO test data does not specifically state what the overlap seam width was on the test boot, it would strongly appear that the strength mar- . gin is so high that even a reduction of 1/3 in the area of the overlap would have the effect of changing the safety factor from 22:1 to approximately 14:1. i It is the SRIC's conclusion that while the allegation relative to the reduction in seam from 3 to 2 inches is correct, the reduction would have no significant effect on the performance of the boot in service at CPSES and that, therefore, the allegation has.no technical merit.
,1 .
16 i t Regarding the matter of the possibility of some undefined problem with the boot fabric, the BISCO site manager stated that his company has been engaged in a law suit with the supplier of the fabric but only in regard to the per-formance of the fabric in one application which is understood to involve the tearing of the fabric after being punctured. It is understood that the 4 puncturing has occurred when a gel type radiation seal hardens under.radia-tion. Since the specific design involved is not scheduled for use at CPSES, the allegation has no technical merit. Regarding the matter of possible sep,aration of the radiation seal aggregate material from the carrier material, the SRIC can only conclude that the al-legation is potentially correct but without apparent merit. The BISCO test reports indicate that the seals involved met the engineers specification. The separation of the aggregate (powdered lead) from the carrier (a silicone material) would appear to be process sensitive in that if they are not well mixed, pockets of lead might form with resulting pockets of silicone without sufficient lead. Since the specification and the BISCO procedures require careful control and monitoring of the mixing process, the SRIC can only con-clude that these measures are effective in production operations as they were in preparation of the test samples.
- 15. Electrical Cable Splicing The SRIC became aware that the Comanche Peak project electrical engineer had authorized the splicing of safety-related and auxiliary electrical cables within several control panels during the inspection period. Since the licen*ee has committed in FSAR Section 8.1 to comply with IEEE 420,
" Trial-Use Guide for Class lE Control Switchboards for Nuclear Power Gener-ating Stations," which forbids splicing of wiring in such panels, the SRIC judged that the licensee was deviating from these commitments. The licen-see engineer indicated that he interpreted the IEEE standard to prohibit such splicing only between the cabinet terminal boards and the cabinet devices and did not prohibit such splicing in the field run cables attach-ing to the terminal boards. The engineer stated that action had been initiated with the NRC Office of Nuclear Reactor Regulation to clarify the issue in the FSAR. The SRIC confirmed that such action had been initiated
! by a telephone conversation with the NRR Licensing Program Manager for Comanche Peak. Pending action by NRR, this matter will be considered as an unresolved matter.
- 16. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of non-compliance, or deviations.
One such item, disclosed during the inspection, is discussed in paragraph 15 above. This item is identified as " Splicing of Electrical Cables in Cabinets." (8324-01)
.1 1 17
- 17. I'.anacement Interviews The SRIC met with one or more of the persons identified in paragraph 1 of this report at frequent intervals during the inspection period to ~
discuss the licensee's position and proposed, actions on a significant number of issues which occurred during the period.
b--- . . miV ou frWSl, Q$st OCT '# 3 1983 In Reply _r To: . . r _ Occkets: 45/83-2 _ _ 50-446/83-14 i l Texas Utilities Generating Company ATTN: R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower - Dallas, Texas 75201 Gentlemen: This refers to the inspection conducted by Mr. L. E. Martin of this office ' during the period June 27 through September 16, 1983, of activities authorized by NRC Construction Permit CPPR-126 and CPPR-127 for Comanche Peak, Units 1 and 2, and to the discussion of our findings with R. G. Tolson, and other members of your staff at the conclusion of the inspection. This inspection was the Region IV followup to the Construction Appraisal Team's (CAT) inspection as documented in NRC Inspection Report 50-445/83-18; 50-446/83-12. The scope of this inspection was to perform a more in-depth review and evaluation of the " Potential Enforcement Findings" (unresolved issues) which were identified in Appendix B of the CAT report and referred to Region IV for resolution. Within these areas, the inspectioh consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector. These findings are documented in the enclosed inspection report. Within the scope of the inspection, no new violations or deviations were identified, in this report. Four violations were identified as a result of the CAT inspection and documented in NRC letter to TUGC0 of'May 31, 1983, and the attached Notice of Violation. One unresolved item is identified in the Electrical and Instrumentation Construction Area of the enclosed inspection report. The Region IV inspector reviewed the and procedures for piping and supports, electrica , and room wa kdowns. These programs and procedures contain the appropriate elements to ensure that the actual as installed components and systems will be in accordance with the latest design requirements. The actual implementation of t w will be followed through subsequent Region IV inspections on a room or area basis as required by the NRC
/[
W wW Inspection Manual, MC 2512.
.ce , 6 ES LMartin dsm RPS-A h DHunnicutt RPBlee, GMadsen DRRP&EP JGagl do bjI&E JTaylor 9/4!lP83 9/p/83 g9/#/83 9/.;M 3 gg /s/)/83 FILE C0? '
Texas Utilities Generating -2 OCT 3 79g3 Company In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, within 10 days of the date of this letter, and submit written application to withhold information contained therein within 30 days of the date of this letter. Such application must be consistent with the requirements of 2.790(b)(1). Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely, "cos.n. : i n ca eji W. A. cacssi.u.v' G. L. Madsen, Chief Reactor Project Branch 1
Enclosure:
Appendix - NRC Inspection Report 50-445/83-28 50-446/83-14 cc w/ enclosure: Texas Utilities Generating Company ATTN: H. C. Schmidt, Project Manager 2001 Bryan Tower Dallas, Texas 75201 Texas Utilities Generating Company ATTN: 8. R. Clements, Vice President, Nuclear 2001 Bryan Tower, Suite 1735 Dallas, Texas 75201 bec to DM8 (IE01) bec distrib. by RIV: RPB1 0. Kelley, SRI-Ops RPB2 R. Taylor, SRI-Cons TPB Section Chief (RPS-A) J. Collins, RA J. Gagliardo, DRRP&EP C. Wisner, PA0 L. Martin M. Rothschild, ELD J. Taylor, I&E MIS SYSTEM J. Scinto, ELO RIV File TEXAS STATE DEPT. OF HEALTH Juanita Ellis David Preister _a
APPENDIX
. U. S. NUCLEAR REGULATORY COMMISSION REGION IV I NRC Inspection Report: 50-445/83-28 50-446/83-14 Dockets: 50-445; 50-446 Construction Permits: CPPR-126 l CPPR-127 Licensee: Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station, (CPSES) Units 1 and 2 Inspection At: Comanche Peak, Units 1 and 2, Glen Rose, Texas Inspection Conducted: June 27 through July 15, 1983, July 27-28, August 4, 1983, and September 15-16, 1983 Inspector: / fjV/A/fr L.~E. Martin, R Etor' Inspector ' patef l
Engineering Se ion Approved: k h h 3 83
- 0. M. Hunnicutt, Chief gate /
Reactor Project Section A Inspection Summary Inspection Conductec June 27 throuch July 15, July 27-28. Acqust 4 1983 and Septencer 15-16. 1983 (Report 50-445/83-28: 50-446/83-14) Areas InsDected: This inspection was the Region IV follow up on Construction Appraisal Team Inspection Report 50-445/83-18 and 50-446/83-12. The inspection involved 144 inspector-hours onsite and 22 inspector-hours in-office for a total of 166 inspector-hours by one NRC inspector. Results: Within the scope of this inspection, no new violations or deviations were identified, one unresolved item was identified in the Electrical and Instrumentation Construction area involving separation of NIS conduit from fluorescent light fixtures, see page 7 of this report. b r -
2 Details
- 1. Persons Contacted __
Principal Licensee Employees
*G. 8. Crane, Construction Manager
*R. G. Tolson, Site QA Supervisor .
A. Vega,'Suptrvisor QA Services C. T. Brandt, Mechanical / Civil QA/QC Supervisor
- 8. C. Scott, Eiectrical QA/QC Supervisor R. E. Camp, Lead Start Up Engineer
.M. R. Blevins., Maintenance Supervisor J. C. Finneran, Pipe support. Engineering Supervisor
*L. M. Poppelwell, Project Electrical Engineer I. Voglesang, Assistant Project Electrical Engineer F. L. Power, Area Supervisor Electrical Engineering D. Westbrook, As-Built Group Supervisor H. Williams, QC Superintendent, Non ASME Brown & Root (8&R)
*0. C. Frankum, Project Man.ager G. R. Purdy, Site QA Manager
- 8. O. Cromeans, QA Records Management Supervisor H. Hutchison, Project Control Manager W. Mahan, QE Electrical J. 8. Leutwyler, Electrical QC Supervisor
- 8. McNe111e, QC Inspector Others J. Foland, Site Representative, Westinghouse Electric Corp. (Westinghouse)
M. Coats, Lead ANI, Hartford Steam Boiler
- Denotes those persons who attend the exit interview.
The Region IV inspector also interviewed other licensee, B&R, and subcontractor personnel during the course of the inspection.
- 2. Recion IV Followup to Construction Appraisal Team's (CAT) Potential Findinos This inspection was the NRC Region IV follow up on potential enforcement .
findings identified during the CAT inspection and documented in NRC Inspection Report 50-445/83-18 and 50-446/83-12 for CPSES, Units 1 and 2. The scope of this inspection was to review, evaluate, and provide appro-priate action or closure of the 16 items addressed in Appendix 8 of the above mentioned inspection report and the specific items addressed in the
~
. 3 Details Section of the report. Four of the 16 items had been identified as violations prior to the commencement of this inspection. These four violations are referenced in this report, but were transmitted to the licensee under separate letter dated May 31, 1983, and will be documented and closed out in a normal manner through the routine inspection program.
The specific items in this report are addressed in the same order as they appear in Appendix B of the above report. Electrical and Instrumentation Construction
- 1. CAT Potential Findino " Contrary to 10 CFR 50, Appendix B, Criterion X, and FSAR Section 17.1.10, certain inspection activities were not executed to verify installation conformance with procedures including cable spacing in trays, cable bend radii, cable fill, cable supports, and tray installation hardware (Section II.B.1.a L, d, and Section II.B.4.b(1))."
The Region IV inspector followed up on the specific items identified in the appropriate sections of the CAT report as follows: Section II.B.1.a " Cable Spacing in Tray" - Eight tray sections were identified that contained improperly spaced medium voltage power cables. The requirement of Project Specification 2323-ES-100 for the spacing of power cables, is one quarter of the diameter of the largest cable, based on the FSAR conmitment to the Insulated Power Cable Engineers Association (IPCEA) Publications P-46-426 (1962) and P-54-440 (1972). These publications are recognized and utilized in the power industry to properly size power cables, depending on raceway configuration, copper temperature, ambient temperature, cable type and configura-tion. These publications provide tables and calculation techniques for determining the amperage capacity for a given cable. Changing from maintained spacing to random lay in cable tray requires derating of the cable amperage capacity. Three of the eight tray sections (T12GABF27, T120SB006, and T12GABF14) had been addressed by engineering in design change authorizations (DCA's 15,515, 15,060, Revision 2, and 15,171, Revision 1, respec-tively) prior to the CAT inspection. These three tray sections were l appropriately designated by engineering as random lay tray sections. The remaining tray sections (T120ABB30, T12GABP71, T11GEA323. T120ABB10, and T11GSAB06), as a result of the CAT identification, were appro-priately addressed in a nonconforma r t NCR) and reworked to provide the maintained spacing
4
- 4 Section II.B.1.b, " Cable Bend Radius" - Five cables were identified that were installed with less than allowable bend radius.
After identification by the CAT, the licensee performed two inde-pendent inspections of each cable utilizing senior inspectiop person-nel and templates, and could not confirm that any of the five cables were less than minimum bend radii. The Region IV inspector examined the cables and the associated NCR's and found the cables to be acceptable at the time of this inspection. Visual inspection of the cable indicated that the cable.had not been damaged. Because of the location of these particular cables, the probability of identifying any problems with the cables during thgg h ired by QI-QP-11.3-40 wou e igh. This inspection had not been performed, as the battery rooms were not complete. The following cables and NCR's were specifically checked: E0102532 1#* @ Battery Charger BC1ED1-2 2/* NCR E-83-00623 EG102592 @ Battery Charger BC1ED2-2 NCR E-83-00624 EG102595 E @ T12GCBF82 NCR E-83-00626 E0102534 @ T120CB036 S# /C12030560'5 NCR E-83-00625 EG100032 5 " @ T11GEAB37/C11G05112 NCR E-83-00519
*See Attachment Section II.B.1.C, " Cable Tray Fill" - Two cable tray sections were identified as having cables that extended over the side rails.
The CAT had previously determined that Tray Section T12GSBG22 was not an overfilled tray section. However, the Region IV inspector did find that Tray Section T13GSCE25 which was directly below T12GSBG22 had three cables that extended above the side rails due to excessive i slack in two of the cables and insufficient slack in one cable as it exited the tray to a conduit above. The Region IV inspector also determined that Tray Section T13GACD14 l had a group or bundle of cables that extended about 1 inch above the i side rail for several feet. Neither of these tray sections are overfilled or would require tray covers, however, the cables needed to e dressed into the tray. This area has not had the ue to high level of ongoing construction activity in e area. Section II.B.1.d, " Cable Supports" - Several runs of 4/0 7/ (see Attachment) 6.9KV SHLD cable were installed in vertical riser Tray Sections T11GSAB01 and T11GSAB45 for over 100 feet without the Kellem Grips required by Project Specification 2323-ES-100. l 1
5 The Region IV inspector concurs that, in fact, Kellem Grip type supports were not installed on the cables in these tray sections. However, the cables were tied to the trays every 18 inches. The constructability aspect of these type tasks led the canstructor to the conclusion, with the consent of engineering, (as documented on RFIC-EE 10,968) to install these grips after all cables had been installed in a given tray section. This was to insure maximum utilization of a given number of grips and limit congestion. The aspects of the inspections for this task were adequately addressed in the QI-QP-11.3-40, " Class 1E Electrical Post Construction Verifica-tion," and QI-QP-11.3-50, " Cable Grip Support Installation Inspec-tion," procedures. Section II.B.4.b(1), " Cable Tray Attachments" - Five tray sections were identified as not being properly bolted together or properly attached to the associated supports. The Region IV inspector found that these problems have been addressed individually by the licensee as follows: NCR-E-83-00316 was written on February 1, 1983, to track the missing hardware on T13GRCLO8 and T13GRCLO9. This hardware had evidently been removed sometime after the initial inspections on September 23, 1982, and July 16, 1982, respectively. I
. Inspected Item Removal Notice 122229 was initiated to install the missing hardware. The missing bolts between T13GCF019 and T13GCF020 were replaced and inspected on January 27, 1983. The missing splice plate was installed and inspected on February 10, 1983. The missing hardware on T13GACE94 was replaced utilizing NCR E83-0028 and Inspected Item Removal Notice 133329 and was inspected on February 10, 1923, (IR. E55872).
These items were in areas where th All of th items were actory on the This item is considered closed.
- 2. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion XVI, and FSAR Section 17.1.16, the established inspection program did not provide adequate controls to assure that deviations from electrical and electrical / mechanical separation criteria as defined in the FSAR were promptly identified and corrected (Sec-tions II.8.1.f, II.B.4.a and II.B.4.b(2))."
Section II.B.1.f, " Electrical Separation," - Identifies two concerns:
- 1. There appear to be instances of separation problems w l and/or corrected.
i
6 The licensee instigated the Electrical 'n the early part of 1982. - 7 gLeo.Oe -
-(((( 2. Two barriers were installed whose configurations provided e inadequate protection between redundant wiring as well as barriers installed in the control boards of a type other than Service Air Company Stainless Steel Flexible Conduit Type SS63 as required by Project Specification 2323-ES-100, Section 4.2.2.3.
At the time of the CAT inspection, the control boards had been continuously undergoing rework. At the time of the Region IV inspec-tion, site engineering was doing a detailed examination of the control boards to provide the design of any additional barriers that may be required, in preparation for a final QC inspection, prior to installing the fire stop material in the floor openings underneath the control boards. The QI-QP-11.3 electrical separation post construction verification inspection had not taken place at the time of the CAT inspection. The CAT report identified barriers in the control boards other than Service Air Company Stainless Steel Flexible Conduit Type $563, and quoted from Project Specification 2323-ES-100, Section 4.2.2.3 that
" Stainless Steel Flexible Conduit Type SS63 shall be used as a barrier." This section of the specification had a DCA 8830 dated October 22, 1980, that changed the above quote to read " Type FC33 flexible conduit and approved fittings can be used as a barrier."
The " Control Board, Nuclear Safety-Related" Project Specification 2323 MS-605 aise authorizes the use of rigid conduit, 1/8 inch steel plate, st uctural tubing, glastic, stainless steel flex, and steel wireway as barrier materials. All berriers and barrier materials are engineered and approved by the contral board manufacturer. This evaluation and engineering process, by design had not taken place at the time of the CAT inspection. Section II.B.4.a. , " Electrical Conduit" - Four Class 1E Nuclear Instrumeatation System (NIS) conduits were identified which did not have the 2 feet separation from fluorescent light fixtures as required by FSAR, Section 8.3.1.4 and QI-QP-11.3-29.1, Section 3.1.2. The Region IV inspector reviewed the history of the commitments in QI-QP-11.3-29.1, to have NIS conduits maintain a minimum of two feet separation from fluorescent lighting fixtures. Revision 8 of QI-QP-11.3-29.1, dated January 20, 1983 was the first revision to include this commitment. This revision was prompted by DCA 15,733, dated January 17, 1983. Prior to this revision, the QI-QP had only ( addressed separation from noise sources asscciated with various
. 7 levels of power cables and power cable raceway. The original QC inspection did not have any spatial requirements between fluorescent lighting fixtures and NIS conduits. The original inspection effort took place in January 1981.
The Region IV inspector determined that conduits C16810045, C16Y10039, and C16Y10041 and Junction Box JB1A-915Y 8/ (see Attachment) were in fact less than 2 feet from fluorescent lighting fixtures in the ceiling of the control room. Discussions with the~onsite Westing-house representatives revealed that the 2 feet dimension is a recommendation, and not based 'on engineering data. The Region IV inspector also reviewed preliminary test data for the source and intermediate range nuclear instrumentation channels taken during hot
. functional testing (HFT) in May 1983. These circuits were monitored continuously during HFT and this data is now being reviewed-by Westinghouse. The preliminary indication was that the system has a very low noise level, which indicates that the close proximity of the fluorescent light fixtures will not cause any problems with the nuclear instrumentation system. Both Westinghouse and the licensee will continue to monitor the noise levels through the testing phase.
If the preliminary evaluation, of very low noise on the sys. tem is correct, the licensee will submit an FSAR change. If not, the light fixtures will be relocated to at least the 2 feet requirement. The Region IV inspector found that the licensee t,ad identified and documented these same separation problems with NIS conduit and others, before or in parallel with the CAT inspection. This area of the control room was turned over to QC for preliminary verification ; walkdown on February 2, 1983. QC identified trese problems on Electrical Punchlist 10001 on February 2, 1983, and sent it to engineering for review. Engineering responded to QC on meno CPPA-27.195. To ensure further tracking, these items were identified on NCR E-83-01013 dated April 8, 1983. This is an unresolved item pending completion of the noise evaluation of the NI system and, if appropriate, the revision of the FSAR. (8328-01) l Section II.8.4.b(2), " Cable Tray Separation" - Five areas of tray were identified which did not maintain the required separation between redundant trains. The CAT inspection identified the following tray sections which did not maintain the required 3 feet horizontal and 5 feet vertical separation as stated in the FSAR: Train A Train B T130ACG51 T13GACZ92 E#* T130ACG54 T13GACZ71 T130ACG63 T13GCF008
8 Train A Train B T120ABB23 T23GACD85 T120ABA42 T23GACD85
*See Attachment The Region IV inspector examined the areas identified above and the associated raceway drawings. G&H Raceway Drawings 2323-El-701-01, Revision 7, dated March 17, 1983, and 2323-El-712-01, Revision 24, dated May 11, 1983, clearly indicated that tray covers were required in four of the five areas. The Region IV inspector also determined that the revisions of these two drawings that were active at the time of the CAT inspection also indicated that tray covers were required.
These tray covers were not in place at the time of the CAT inspection and were not in place at the time of this inspection. The FSAR, Section 8.3.1.4, requires the 3 feet /5 feet separation and also states in a subsequent paragraph "Where plant arrangements preclude maintaining the minimum separation distance (as stated in the beginning of this section), the redundant circuits are run in solidly enclosed raceways or other barriers provided between redundant circuits." The observation of these raceways would indicate a separation problem, because tray covers are not normally installed on trays until all cables a,re installed in the associated tray and the area is ready for turnover. The separation problem between T130ACG54 and T1'3GACZ71 was documented on the inprocess separations punchlist for Room 207 on July 12, 1983. Tray Section T13GACZ71 had undergone a design change that relocated it horizontally to within 2 feet of Tray Section T130ACG54 This relocation would then require the addition of tray covers to T13GACZ71. This problem has been documented and is being appropriately tracked to insure that proper corrective action is taken. In Section II.B.7, the CAT inspectors addressed concerns with the use of request for information clarificiation (RFIC) and punchlist to document deviations and deficiencies in lieu of the NCR system. The Region IV inspector reviewed more than 500 electrical RFIC's covering the period of November 1981 through December 1982 to determine if Inspection results or engineering information were being documented on the RFIC's. In all cases reviewed by this NRC inspector, construc-tien or QC was asking for engineering clarification to determine if a cor.dition was nonconforming. In all cases, if engineering determined that the condition was nonconforming, an NCR, CMC, or DCA was initiated to correct the nonconforming condition, or the craft was advised in a three part memorandum to correct the installation to meet the design documents or specification.
. 9 In November 1982, TUGC0 QA recognized the problem with the control of RFIC's and initiated a draft change to QI-QP-11.3-29 to provide better control of the documentation of separation problems on RFIC's and punchlists. QI-QP-11.3-29, Revision 3, issued on January 29, 1983, required owner approval on separation problem documentation.
This was implemented by requiring engineering to review RFIC's and punchlists initiated by QC and document this review to the QA super-visor via a CPPA office memorandum. This action made the evaluation of RFIC's and punchlists permanent plant documents. The Region IV inspector reviewed two of these memoranda (CPPA-26,996 and CPPA-26,947). j QI-QP-11.3-29, Revision 11, utilizes electrical separation deficiency reports, IR's, and NCR's to document inspection results. The Region IV inspector reviewed numerous RFIC's, NCR's, DCA's, IR's, and trend reports and determitied that the licensee has been continually identifying and addressing separation problems. The licensee's inspection program for electrical and electrical / mechanical separa-tion provides adequate controls and criteria to assure that deviations are promptly identified and properly tracked to close out. TUGC0 Procedure QI-QP-11.3-29, " Electrical Separation," Revision 0, dated January 15, 1980, and all subsequent revisions to date, identified a poten,tial problem in accomplishing the final separation requirements with a single in process inspection during installation. The procedure requires the final inspection af ter notification from construction / engineering that processing is complete and that separation has been accomplished. TUGC0 Procedure QI-QP-11.3-40, " Class 1E Electrical Post Verification," Revision 0, dated June 4, 1982, and all subsequent revisions co date, require by checklist attached to the inspection report, that the following items, in part, be verified:
. Verify cable tray
. Verify conduit
. Verify cable
- 1. Free of damage and debris
- 2. Bend radii not violated
- 3. Trained and secured
- 4. Spacing and separation maintained
10 l TUGC0 Procedure QI-QP-11.3-45, " Release and Inspection of Electrical Areas I for Application of Thermo Lag Fire Protection Coating," Revision 0, dated
! August 18, 1982, and all subsequent revisions to date, require by inspec-
! tion that the following items, in part, be verified:
. Cables shall be free from damage and debris.
l . Cable bend radius is not violated. : i
. Cables are trained and properly secured . . .
. Power cable spacing has been maintained . . .
. Cables below side rail. l
. Separation in accordance with QI-QP-11.3-29.
n pending implementation and review of the final inspection
; , a was under development by.the licensee at that time. This was
- () The final inspections addressed in the above potential findings had not i q .[
g}f>g e been executed. By procedural requirements the inspection was not to be Y/ executed until the room or area was complete. This was a controlled, i
'pt[s /
, hlanned,andeffectivemeansofaddressingaproblemthatexistsinareas
./.J # gpf high construction activities. These problems are generally not of the
,./ <
- y. type that get " cast in concrete" and require expensive rework or possible
) degradation of a structure or component.
r
}OneunresolveditemconcerningNISconduitsisaddressedintheRegionIV
' ,/ g j g ,f y' followup to Section II.B.4.a. I'Y l 3. CAT Potential Findino " Contrary to 10 CFR 50, Appendix B, Criterion V,
\\. ! FSAR Section 17.1.5, and IEEE Standard 450, procedures to implement 4 inspection activities relative to certain aspects of battery mainten-3 ( ' ance have not been developed or implemented (Section II.B.3.e.)" 10/
l / (see Attachment) l Section II.B.3.e., " Station Batteries" - Identifies that the procedure
! used to perform surveillance of the vital station batteries (ELM-701, Revision 0) did not implement the requirements of IEEE Standard 450-1975 and Regulatory Guide 1.129 for yearly inspection of cell to cell and i.
_ _ _ _ ._ -. _ ._..___._ _ . _ _ _ _ _ _ . _ _ _ _ - - ~
, 11 terminal connection detail resistance, and there was no documented evidence that these attributes (resistance checks) had been inspected or verified.
The CAT, inspection report identified the commitment in FSAR Section 8.3.2.1 that "All maintenance and testing procedures and criteria for replacement are in accordance with IEEE 450-1975, and Regulatory Guide 1.129," This commitment to IEEE Standard 450-1975 and Regulatory Guide 1.129 was incorporated into the FSAR in Amend-ment 27 which was effective October 2, 1981. The following is a chronological tabulation of installation status of the Class 1E vital 125 volt DC batteries after the effective commit-ment date of October 2, 1981, for IEEE 450-1975 and Regulatory Guide 1.129, Revision 1, 1978: Events Batteries / Dates 1ED1 1ED2 1ED3 1ED4 Batteries Removed for 11/25/81 2/2/82 NA NA UPS* Modifications Batteries Installed and 1/27/82 3/29/82 NA NA Returned to TUCCO Batteries Removed for Rack 1/28/83 1/28/83 NA NA Modifications Batteries Installed and 2/14/83 2/14/83 2/14/83 2/14/83 Returned to TUGC0
*L'PS - Uninterruptable Power Source
. From the above data, it can be seen that Battery 1ED1 was the only battery that had been installed for a period of sufficient length to meet the time period (1 year and 1 day) for the yearly cell to cell resistance check. Recognizing that the batteries were going to be removed for rack modification would justify the postponement of the cell to cell resistance checks on Battery 1E01. The CAT inspector identified that Electrical Maintenance Proce-dure ELM-715 was in the process of being issued to cover the require-ments for the cell to cell resistance checks. The procedure was in draft .n November 1982 and was issued as Procedure EMP-706, " Battery Inter-Cell Connector Resistance Test," Revision 0, on May 6,1983. The Region IV inspector reviewed this procedure and determined that it was an acceptable method of implementing the cell to cell detail resistance requirements of IEEE 450-1975. l l
12 At the present time all four of these batteries are in place and operational. However, many of the inter-cell connectors are temporary jumpers. These jumpers will be removed and the permanent hardware will be installed prior to final turnover of the batteries to TUGCO. This item is considered closed. Mechanical Construction
- 1. CAT potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion V, FSAR Section 17.1.5, and QI-QAP-11.1-28, certain QC accepted ASME p1pe supports / restraints are not installed in accordance with the design document to which the pipe support / restraints were inspected (Section III.B.2)."
Section III.8.2 - Twenty-one pipe supports were identified that had specific prcelems, plus numerous unspecified instances of broken or missing cotter pins, loose locknuts, and U-bolt problems. 494B Documentation Requirements Prior to System N-5 Certification." Five specific ASME Class 2 and 3 supports were identified as having U-bolt problems. Four of these pertained to the nut position with respect to plate. This problem predominately concerned small bore pipe, where due to space limitations between the pipe and the U-bolt, a nut cannot be utilitized on the side next to the pipe. The licensee has issued DCA's that clarify the design criteria for large and small bore pipe. The inspection procedures now reflect this information. The Region IV inspector determined that Supports CS-1-063-028-532R, AF-1-059-001-S33R, H-RC-1-RB-039-015-2, and H-RC-1-RB-038-004-2 all had appropriate clearance even though the nut arrangement did not agree with the drawing. . Support H-CS-l*RB-017-001-2 was found not to have the 1/16 inen gap specified by the design. This was identified in December 1982 by the as-Duilt walkdown. Design review of this situation indicated that the 1/16 inch gap was not required, however, NCR M9647S was initiated to have the U-bolt installed per the drawing. , Support OD-2-019-007-F33R was identified as having incorrect clearance between lugs on the pipe and the support (11/32" vs. 3/32"). The pipe was found to have shifted a 1/4", probably due to testing, causing one of the four lugs to no longer be properly positioned with respect to the tube steel of the support. The remaining three lugs would have adequately restrained the pipe. NCR M-5322 was initiated to modify the support to agree with the new pipe position, to provide the design clearance on all four lugs. Support RC-1-075-026-C61R, NCR M-53215 was initiated concerning the dimensional deviation. CMC 66083 changed the drawing to agree with the installed support.
f 13 s identified as having ar.chor bolt 11/ I (see ac ment) threads that were not as required. ' CMC 58038, l Revision 8, called for the threads on the anchor bolt to be upset. The normal method of thread upset is to peen the threads in two or more places with a center punch. Thread upset is to prevent the nut from backing off the bolt. At the time of this inspection, the threads had been upset with a center punch. identified as missing the snubber load pin. This support was initially inspected on January 8, 1983. Evidently, the pin had been removed. NCR M-51375 was initiated to replace the pin. The pin was in place at the time of this inspec-tion. identified as having a void in the concrete near the concrete expansion anchor. A small 2" long x 1" wide x 1" deep void was adjacent to the base plate, approximately 2 inches from a Hilti bolt. The void did not cause a violation of the minimum embeddment of the anchor bolt. The void.had been repaired at the time of this inspection. identified as having a loose locknut. This total strut was replaced by NCR M-74215 on June 9. 1983. 5 identified as having a loose locknut. The locknut was tight at the time of this inspection. The CP-QAP-12.1 design verification was completed on May 17, 1983. (see Attachment) was identified as being mismarked as a Class 2 support. The hanger was correctly marked but the base plate had been marked as -003-2 and tnen remarked as -003-3. The support was design verification inspected to CP-QAP-12.1 on May 11, 1983. Supports -- - ' (see Attachment) were identified as punchlist items considere y CAT to be significant enough to question why NCR's were not initiated. The licensee immediately initiated NCR's M-53255 and M-53265. Both items turned i out to be drafting errors. The drawing for FW-1-098-008-C625 had
" field to verify" notation on the Hilti bolt dimension. The subsequent revision indicated 10 inch instead of the 10-3/4 inch. The new VCD shows the 10-3/4 inch dimension correctly. When CMC 60353 for support CS-1-063-046-532K was incorporated into the drawing the draftsman incorrectly showed the arrangement of the nuts on the U-bolt. The drawing has been corrected.
Three additional Support h >C nd [ M re identified as punen ist items that had been documented on NCR's M-50365, M-48635, and M-48615, respectively. The concern was that the NCR's did not indicate that the original QC inspections were improperly performed or that unauthorized / uncontrolled
14 work had been performed on these supports and the NCR's did not provide a space for evaluation or signoff for action to prevent recurrence. There is no requirement to have a recurrence control evaluation or signoff on ASME or other NCR's. The undersized welds were skewed type weld on tube steel. CAT inspectors did not identify any under-sized welds in the approximately 224 supports they inspected. These particular NCR's were written during the time frame that B&R was ' evaluating their program to address M at the authorized nuclear inspector activity tracking, and documentation of the initial packages submitted to the ANI. The Region IV inspector discussed, with the lead ANI, the concerns that CAT inspection report addressed in Section III.B.2. The lead ANI indicated that he had been concerned with the acceptance / reject rate of supports and packages submitted to his organization by B&R. His concern started in late October 1982. He had addressed these problems through his normal communication mechanism M reports) to B&R. When asked if these concerns had been relayed to B&R as generic problems the lead ANI responded that a special report had not been prepared. The lead ANI indicated that and ., He had not been able to evaluate the results of this effort at that time of the CAT inspec-tion. The lead ANI said that over 2,000 support packages have been accepted and submitted to the vault and that the. final N-5 walkdown inspection of the Fuel Building supports was going smooth. The lead ANI said that the weld data cards provide more appropriate details of operations and scope of work performed and the inspection records include drawing numbers and revision levels utilized for the inspec-tion of the supports. Supports MS.1.0A." W- , S M , 14/ and SW-X-026-003-J05R 15/ (see Attacnment) were identified as havinQ apparent noncon-forming conditions for which no changes had been specified since the QC acceptance of the support, s-Built Verification 5610 indicated on CMC 41442, Revision 11, issued February 3, 1983, that possible interference existed with FW-1-017-712-C72K and MS-1-003-010-C72K. This was being evaluated by engineering at the time of the CAT inspection. This support is scheduled to be relocated to clear the interference problems. Nce .
- the original as-built inspection identified a (1/4" vs. 1/16") with this box type support.
subsequent as-built inspection, 3058, indicated that the initial A measurement was incorrect and that the clearance was the required 1/16 inch. This was confirmed by the Region IV inspector. This item did not require a CMC or drawing change.
I i 15 l ! l this support was subsequently downgraded to { L' lass 5 on March 9, 1983. The Region IV inspector and the 79-14 1 As-Built Group could not identify any potential problem with this support. Review of the above CAT identified problems, including discussions with the ANI, the scope of the perceived problems, and the fact that none of these supports had been final inspected per CP-QAP-12.1 indicates that the program appears to be functioning as intended. This item is considerea'clcsed.
-2. CAT potential Finding " Contrary to 10 CFR 50, Appendix B Criteria X \
and XVII, and fSAR Section 17.1.10 and 17.1.17, j M4 . ? (Se'ctions III'.l[TaWB 3). '" ) Section III.B.2 - The potential finding above states in part that, "an inspection program has not been established to verify and document installation conformance to drawing require'1ents in regard to pipe sucports/ restraints . . . ." The question is not one of failure to establish an inspection program, but a question of whether the program that has been, established is adequate. Rather than try to answer each question, or staterent of concern, the Region IV inspector has reviewed the program to dettermine its adequacy. This review included a partial review of tne program as it was at the time of the CAT inspection and the program as it stands at the time of this inspection. The Region IV Inspector reviewed the following procedures and instruc-tiens' relating to the final inspection of piping systems and supports for ccmpliance.with the FSAR, 10 CFR Part 50, Appendix B and the acpropriate regulatory guides and ANSI standards: B&R CP-QAP-12.1, Revisions 4 and 7, " Inspection Criteria and Docu-mentation Requirements Prior to System N-5 Certification" CP-QAP-17.1, Revision 4, dated February 2, 1983, " Corrective
Action" TUSI
- CP-EI-4.5-1, Revision 8, dated January 4, 1983, " General Program For As-Built Verift:ation."
16 TUGC0 CP-QP-11.13, Revisions 5 and 6, "As-Built Verification" QI-QP-11.13-1, Revision 8, dated January 4,1983, "As-Built Piping Verification Instructions" QI-QP-11.13-2, Revision 1, dated December 16, 1981, "Installa-tion Reinspection of Pipe Supports" QI-QP-11.16-1, Revision 14 and 21, " Installation Inspections of NNS Seismic Category II Supports For Class V Piping" CP-QP-15.0, Revision 3, dated July 19, 1983, " Tagging System" CP-QP-15.4, Revision 5, dated June 22, 1983, " Design Change Verification" CP-QP-16.0, Revision 10, dated June 10, 1983, "Nonconformances" CP-QP-17.0, Revision 1, dated June 10, 1983, " Corrective Action" CP-QP-18.0, Revision 12, dated July 19, 1983, " Inspection Report" The program represented by the above procedures s nspec-tion documentation. The licensee is also doing a complete review of CMC's to ensure that they have been incorporated into the installa-tion and documented by inspection reports. These programs essentially result in a re-inspection of all ASME Class 1, 2, and 3 and many non-ASME Class 5 supports. There was concern on the part of the CAT Team relating to missing hardware, documentation problems (NCR, IR, Punchlist), partial inspections instead of final inspections, design change verification, and drawing revision numbers. All of these concerns were originally addressed in the program, however, they were sometimes all encompass-ing type statements. The procedures are now more specific and The scope of these procedures has greatly broadened since their actual implementation began due to identified problems in previous inspections. A program that is able to detect previous short comings in inspections, ensure closed loop on design change implementation and verification, and provide final inspection documentation is a good program.
17 As pointed out by the CAT, they were unable to adequately assess the total program as it was just being implemented. Now with 2000 support packages completed and in the vault and the N-5 walkdown virtually completc in the fuel building, the program appears to be working. This program may still require some tuning, but the basic elements are there. The Region IV inspector reviewed nine completed supports and the document packages and found them to be adequate. Section III.B.3 - Identified a failure to properly install and inspect mechanical equipment. This potential finding was reviewed by the SRIC and determined to be a Severity Level V violation. This violation was identified in Appendix A of the Region IV letter to TUGC0 dated May 31, 1983. This item will be followed by the SRIC.
- 3. CAT Potential Findino " Contrary to 10 CFR 50, Appendix B, Criteria V and X, and FSAR Sections 17.1.5 and 17.1.10, installed and QC accepted heating and ventilation and air conditioning (HVAC) duct, supports, and equipment do not conform to design requirements. In addition, inspection procedures hav.e not been established or executed to verify conformance pf HVAC supports to design drawings (Section III.B.4)."
This potential finding.was reviewed by the SRIC and determined to be a Severity Level IV violation. This violation was iden'tified in Appendix A of the Region IV letter to TUGC0 dated May 31, 1983. This item will be followed by the SRIC. Welding and Nondestructive Examination
- 1. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion IX, and FSAR Section 17.1.9 certain special processes relative to the HVAC system wene not adequately controlled, including improperly qualified procedures; .
; improper certification of NDE personnel (Section IV.B. .
The failure to control special processes relative to welding of the i HVAC system and improperly qualified inspectors are addressed in the violation identified in CAT Potential Finding 3 of the Mechanical Construction section above. The inspection of welding relative to HVAC is visual inspection in accordance with AWS 01.1. The normal certification of welding inspectors related to NDE is SNT-TC-1A. SNT-TC-1A does not address visual inspection certification. This will be followed up as
" improperly qualified inspectors" as stated in the preceding para-graph.
i CAT report Section IV.8.3.C addresses improperly qualified weld procedures utilized by the HVAC contractor. G&H Specification 2323-MS-85, paragraph 2.14(a) " specifies that welding procedures and
l 1 18 l welders be qualified in accordance with AWS D1.1 Section 5 or ASME l Section IX." This statement was to allow the contractor to choose either code, but not necessarily both. DCA 9898, Revision 2, to paragraph 2.14(b) provides the following clarification:
"All welding shall be in accordance with ASME B&PV Code Section IX. Acceptance criteria shall be in accordance with AWS 0.1.1, " Structural Welding Code, Sections 6 and 8."
The Region IV inspector confirmed the selection of Bahnson to' qualify to ASME by a further review of the correspondence file that conveyed weld procedure qualification information to B&R and TUSI. This file covered submittals from February 1978 through July 1982. All of these letters indicated that the procedures were being qualified to ASME Section IX. The Region IV welding specialist reviewed the following Bahnson Service Company welding procedures and supporting procedure qualifica-tion reports:
. Bahnson Welding Procedure Specification BSC-10, Revision 2
. Bahnson Procedure Qualfication Report BSC-10.1.1.
. Bahnson Welding Procedure Specification.BSC-11, Revision 3
. Bahnson Procedure Qualification Report BSC-11.3
. Bahnson Procedure Qualification Report 302-A and W3.5A
. Bahnson Welding Procedure Specification BSC-12, Revision 3
. Bahnson Procedure Qualification Report BSC-12.1
. Bahnson Welding Procedure Specification BSC-13, Revision 1
. Bahnson Procedure Qualification Report BSC-13.4
. Bahnson Welding Procedure Specification BSC-14, Revision 0
. Bahnson Procedure Qualification Report BSC-14.1
. Bahnson Welding Procedure Specification BSC-15, Revision 0
. . Bahnson Procedure Qualification Report GM8-G-S
. Bahnson Welding Procedure Specification BSC-20, Revision 3
. Bahnson Procedure Qualification Report BSC-20.2
19
. Bahnson Welding Procedure Specification BSC-21, Revision 1
. Bahnson Procedure Qualification Report BSC-21.1 .
. Bahnson Welding Procedure Specification BSC-22, Revision 0
. Bahnson Procedure Qualification Report BSC-22.2 In the areas reviewed, the supporting qualifications for the welding procedures were consistent with the requirements of ASME B&PV Code Section IX, 1974 edition including Summer 1976 Addenda.
This item is consiuered closed. Civil and Structural Construction CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion V and FSAR Section 17.1.5, 16/ (see Attachment) civil construction test procedures were inadequate to ensure that mixer uniformity tests as required by the ASME-ACI-359 code were performed at the prescribed frequency (Section V.B.2)." Section V.B.2 - states, in part, "no mix uniformity tests had been performed to the commitments contained in Section 3.8.1.6 of the FSAR." This section continues on with requirements out of ASME-ACI-359 Sections CC-4223.2 and Table CC-5200-1. This section states, "The licensee could provide no evidence that these tests had been performed. Thus, tests to verify proper operation of the concrete mixers were not performed . . .. The Region IV inspector identified in FSAR Section 3.8.1.2.1 that the licensee was committed to the Proposed Standard Code ACI 359, 1973. The 1973 proposed code Section CC-4223.2 does in fact specify mix time of 1-1/2 minutes unless a shorter time is shown to be satis-factory by ASTM-C94 testing. ASTM-C94-74 was also directly committed to in FSAR Section 3.8.1.2.3 and in B&R Procedure 35-1195-CCP-10, " Concrete Batch Plant Operations." However, the actual performance of mixer uniformity test at CPSES was a requirement in National Ready Mixed Concrete Association's (NRMCA)
" Certification of Ready Mixed Concrete Production Facilities" check-list.
When the CAT inspector asked for the mixer uniformity tests, the persons being asked did not realize that these particular test reports were kept in a separate file as required by Procedure CCP-10. It took the document control people several days to locate them. 1 The Region IV inspector reviewed test reports HCP 13426 and HCP 26087 . performed by Robert W. Hunt Company and Test Reports TUGC0 646 and 647 performed by TUGC0 after TUGC0 assumed testing responsibilities.
. 20 These reports provided test data to meet the requirements of Section 3 of NRMCA Certification and Section 10.3 of ASTM-C94.
This item is considered closed. j Procurement, Storage, and Material Traceability CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion XIII, FSAR Section 17.1.13, CP-QAP-8.1, Revision 5, CP-CPM-8.1, Revision 1, and MCP-10, Revision 7, storage of certain safety-related equipment in outside lay-down areas and installed in the plant was not properly controlled (Section VI.B.2)." This potential finding was reviewed by the SRIC and determined to be a Severity Level V Violation. This violation was identified in Appendix A of the Region IV letter to TUGC0 dated May 31, 1983. This item will be followed by the SRIC.
- 1. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B,-
Criterion II and FSAR Section 1A(B), 17/ (see Attachment) individuals Section VII.B.2.a(2) - identifies four areas in which QC inspectors were certified with less experience than required. The specific areas of concern will be discussed later in this paragraph. CPSES FSAR Section 1A(B), page 1.A(B)-24 commits to implement Regula-tory Guide 1.58, Revision 1 for the construction phase. Regulatory Guide 1.58, Revision 1 is the regulatory guidance and NRC position on ANSI N45.2.6-1978, " Qualification of Inspection, Examination and Testing Personnel for Nuclear Power Plants."
"For inspection activities outside the scope of the ASME Code, inspection personnel are qualified in general compliance with the requirements of Regula-tory Guide 1.58, Revision 1 . . . .
"The capabilities of a candidate for certification shall be initially determined by a suitable evaluation of the candidate's education, experience, training, test results, or capability demonstration."
, "The following is the recommended personnel education and experience for each level.
These education and experience recommendations should be treated to recognize that other factors may provide reasonable assurance that a person can competently perform a particular task. Other factors which may demonstrate capability in a given job are l l l l
, 21 previous performance or satisfactory completion of capability testing."
en ht additional requirement of a high school diploma ca-tion used in lieu of education and experience should result in documented objective evidence (i.e., procedures and record of written test) demonstrating comparable or equivalent competence. The Region IV inspector reviewed the following B&R and TUGC0 proce-dures: CP-QAP-2.1 " Personnel Training and Qualification" QI-QAP-2.1-5 Training and Certification of Mechanical Inspection Personnel" CP-QP-2.1 " Training of Inspection Personnel" The above procedures meet the requirements of ANSI N45.2 6-1978 and Regulatory Guide 1.58, Revision 1. These procedures address the implementation of Sections 2.2 and 3.5 of ANSI N45.2.6 and Posi-tions C.6 and C.10 of Regulatory Guide 1.58, Revision 1. The Region IV inspector reviewed approximately 35 inspector certifica-tion records covering the four areas identified in the CAT report. The Region IV inspector then contacted the responsible CAT inspector to establish the exact concerns of the CAT and the names of the individuals that were considered to be inappropriately certified. The responsible CAT inspector provided the Region IV inspector with
- names and the respective area of concern for each individual. The Region IV inspector had already reviewed the records of five of the individuals and then reviewed the remaining certification records of the individuals identified by the CAT inspector.
The results of the above review are as follows:
- a. "Three individuals were certified Level II (L-II) mechanical inspectors having authority to witness pump or component dis-assembly and reassembly with qualifying experience only in welding and nondestructive examination."
Four individuals were identified that could be fitted into this catego ry.
l :
- , 22
- Individual 1 was hired by B&R in June 1979 as a welder and then was transferred to QC in March 1982. This individual was certified Mechanical Installation / Fabrication Inspector (MIFI)
Level II on July 9, 1982, based on an AAS degree in Welding Technology, 3 years welding experience, a 92.44 test score, and 36 hours of "on the job training" (0JT). This individual was certified Mechanical Equipment Inspector (MEI) Level II on June 22, 1983, based on the AAS degree, 1 year experience as an MEI Level I (January 20, 1983), a 88.71 test score, and 16.5 hours OJT. Individual 2 was hired by B&R in June 1982 as an inspector. This individual had a nonrelated AAS degree but had 2 years of verified previous experience in welding inspection and NDE. This individual was certified as MIFI Level II on March 24, 1983, based on 2 years 9 months experience, high school diploma, a 83.65 test score, and 55 hours OJT. 4 Individ'ual 3 was hired by B&R in May 1978 as a welder and then was transferred to QC in October 1982. This individual was certified MEI Level II on November 12 5 1982, based on 4 years previous experience, a GED certificate, a 84.78 test score, and 22 hours OJT. This individual was certified MIFI Level II on November 21, 1982, based on 4 years experience, a GED certificate, a 93.5 test score, and 20 hours OJT. Individual 4 was hired by B&R in May 1981 as a welder and then was transferred to QC in May 1982. This individual was certified
. MIFI Level II on June 25, 1982, based on 1 year experience, an AAS degree in Welding Technology, a 90.37 test score and 35 hours OJT. This individual was certified MEI Level II on March 10, 1983, based on 1 year 10 months experience, AAS degree, a 94.00 test score and 25 hours OJT.
With the exception of Individual 2 all the above met all of the requirements. Individual 2 by time requirements of N45.2.6-1978 was 5 months short on time between MIFI Level I (August 3, 1982) and MIFI Leve,1 II (March 24, 1983) and was 3 months short on MEI Level II. However, utilizing the capability testing requirements of ANSI N45.2.6, paragraph 2.2, and the documentation requirements of Regulatory Guide 1.58, Revision 1, Position C.10, the individual was certified to Level II based on documented performance,
- b. "One individual was certified Level II as a mechanical inspector having authority to witness pump or component disassembly and reassembly using education as a factor in the qualification process when the education was from a nontechnical, unrelated college degree."
f w - - , . - - . . . - - . .
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23 The CAT inspector interpreted ANSI N45.2.6 to require a technical or related degree if a 4 year college graduation is utilized as a factor for certification. ANSI N45.2.6-1978, paragraph 3.5.2(4) for education and experience recommendations for Level II states; "Four year college graduation plus six months of experience in equivalent inspection, examination, or testing activity." There is no requirement for a 4 year degree to be technical or related in ANSI N45.2.6-1978 or Regulatory Guide 1.58, Revision 1. There were two individuals with nontechnical 4 year Bachelor of Arts degrees. Both individuals had more than 1 year experience that was directly related and validated as equi. valent. The certification of these two individuals met the requirements of ANSI N45.2.6 and Regulatory Guide 1.58. For clarification, by review of QI-QAP-2.1-5 and the associated training outlines, the Region IV inspector determined that both MEI and MIFI inspectors can witness valve, pump or component disassembly and reassembly. However, f.inal alignment / installation of equipment is limited to MEI Level II by job description. c. Electrical inspector as a title is restricted to Level II certification at CPSES. There are no Level I " electrical inspectors" by certification at CPSES. However, the Region IV inspector did identify an individual who had received certifica-tion as Hilti bolt inspector Level I and conduit inspector Level I in less than a month after joining electrical QC. The CAT inspector confirmed the identity of the individual. This individual was hired by B&R in March 1981 in the QC docu-ment area and then transferred to QC on November 15, 1981. This individual was certified Level I Hilti bolt on February 15, 1982, with a test score of 80 and Level I conduit on December 8, 1981, with a test score of 91.66. This individual was a high school graduate with 1 year of college and had 8 years of previous QC experience. The passing test scores and 43 hours OJT was the basis of certification.
- d. "Two individuals were certified Level I anchor bolt inspectors with less than 1 month inspection experience."
The Region IV inspector identified two individuals who were high school graduates who did not have 6 months experience prior to certification as Level I anchor bolt inspectors. The CAT inspector confirmed the identity of the individuals. One of the individuals had a high school education, was hired in November I
24 1981, and was transferred to QA in December 1981. This individual scored 88 on his exam and had 206 hours of OJT for his certifica-tion as a Level I Hilti bolt inspector on February 24, 1982. The other individual had a high school education, was hired on December 1, 1981. This individual scored 80 on his exam and had 47 hours OJT for his certification as a Level I Hilti bolt inspector on December 22, 1981. Both certification folders also contained recommendations by Level II inspectors for Level I certification of the individuals based on high performance. On this basis the requirements for Level I certification were met. The certifications of the above inspectors is generally consistent with the requirements of ANSI N45.2.6-1978 and Regulatory Guide 1.58, Revision 1. The Region IV inspector recognizes that the provisions of capability testing in lieu of actual experience could definitely be taken advantage of and used to the extreme. However, the Region IV inspector did not consider this to be the case with the certifications reviewed during this inspection. This item is considered closed.
- 2. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion X, and FSAR Section 17.1, 18/ (see Attachment) inspection records were prepared and accepted by Level I inspectors as the ' inspector of record' rather than the required Level II ' inspector of record' required by ANSI N45.2.6 (Section VII.B.2.b(1))."
Section VII.B.2.b(1)b. - indicates that the intent of ANSI N45.2.6-1978 is that a Level II inspector be the " inspector of record." The CAT identified areas in civil QC, electrical QC, and non ASME mechanical QC where the " inspector of record" was a Level I inspector and there was no Level II sign-off for the validity of the results. The Region IV inspector, after a careful reading of ANSI N45.2.6-1978, does not agree with the CAT on their interpretation of the intent of the standard with respect to the " inspector of record." To have a Level II sign as the " inspector of record" when that inspector was totally removed from the inspection would not meet the intent of ANSI N45.2.6 or 10 CFR 50, Appendix B. 'The CAT found that inspection reports in most areas were being reviewed by Level II or a senior Level I inspector but were not being signed by the reviewer. With the exception of the civil area, the CAT inspector was satisfied that the reviews were in fact being accomplished, but the reviewer was not signing the inspection report. ANSI N45.2.6 does not require a Level II signature, nor does it require Level II evaluation on every inspection report.
1
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25 ANSI N45.2.6 and Regulatory Guide 1.58, Revision 1, establishes the qualifications (minimum) for inspectors but the documents do not address how to perform inspections, nor does it prevent a senior Level I that exceeds the minimum requirements from reviewing inspec-tion results on simple "go-no go" type inspections. ANSI N45.2.6 does not address when Level II inspections are required. This is done procedurally. The Region IV inspector reviewed several procedures and travelers that required Level II inspection. Some typical examples are solder-ing, NIS triaxial cable testing, penetration conductor inspections, and modification to panel CP1-EPBCED-06. The procedures and travelers identify the specific activities that are beyond the Level I capabilities and must be accomplished by a Level II inspector. The Region IV inspector concludes that the " inspector of record" and the level of review is appropriate for the type of inspections in ques-tion and meets the general requirements and commitments of ANSI N45.2.6. This item is considered cicsed.
- 1. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion XVIII and FSAR Section 17.1.18, QA audits have not been .
conducted at a frequency or at sufficient depth to identify and correct significant problems in various areas of construction; i.e., HVAC and electrical separation (Section VIII.B.2.b(5)(c))." Section VIII.B.2.b(5)(c) - indicates that areas of construction were audited; however, the audits did not identify major construction problems in HVAC and electrical areas. With regard to the major construction problem in HVAC, a violation was identified as discussed in the " Mechanical Construction" area, Potential Finding 3 of this report. The licensee's response is required to address the controls to be implemented to prevent recur-rence. With regard to the major construction problem in the elec-trical area, the electrical and instrumentation construction area of this report adequately addresses the three " CAT Potential Findings." The Region IV inspector was unable to identify a major construction problem in the electrical area. No new violations or deviations were identified in this area. This item is considered closed.
- 2. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criterion XVI and FSAR Section 17.1.16, audit findings related to maintenance instructions identified in 1979, 1981, and 1982, were not resolved in a timely manner (Section VIII.B.2.b(5)(C))."
9
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. 26 l
, Section VIII.B.2.b(5)(c) - indicates that three audit reports, TCP-5, TUG-5, and TUG-14, had identified a common problem concerning the incorporation of vendor / manufacturer maintenance requirements into the ongoing maintenance instructions. "The problem was identified in
, 1979, and it has not been resolved as evidenced by the 1982 audit.
Thus, the effectiveness of the corrective action system for audits is not effective." The Region IV inspector reviewed the following audit reports:
- a. TCP-5 This audit was performed on B&R construction and storage mainten-ance during the period of August 27-30, 1979. Deficiency 1 identified a problem where the manufacturers long-term storage maintenance instruction had not been properly incorporated into the B&R maintenance Procedure MCP-10. The particular problem identified involved the rotation of shafts on large motors.
The followup and close out of this deficiency was documented in Audit Report TCP-9. This audit took place during the period of March 24-April 1, 1980. The corrective action implementation and follow up audit were accomplished within 6 months after the initial audit report was available.
- b. TUG-5 This audit was performed on TUSI startup maintenance activities during the period of June 22-July 2, 1981. Deficiency 4 identified a problem where the manufacturer recommended. operational mainten-ance requirement for maintaining the qualified life of the equipment was not incorporated into the operational maintenance procedure, as required by the maintenance program. This particular problem identified the cycling of the circuit breakers for the Class 1E battery charges under load every 6 months.
The followup and close out of this deficiency was documented in Audit Report TUG-8. This audit took place during the period of February 22-26, 1982, here again the corrective action implementa-tion and the follow up audit were accomplished within 6 months after the audit report was available.
- c. TUG-15 This audit was performed on TUGC0 operations maintenance activities during the period of September 27-October 1, 1982.
This audit did not have a deficiency relating to failure to ,
-incorporate manufacturers instructions into site procedures.
However, 'the audit did identify a proceoural keakness in that the Procedure MDA-301 did not make specific references to , responsibility assignment and documentation of manufactures 1 l r . - - _ _ . - _ _ . , . _ - . . . . . . . _ _ _ _ _ . - . _ . _ , . . . _ , . _ . . .
27 instructions into the actual operational maintenance instruc-tions. The audit identified that implementation of the manufacturers instruction was adequate, but the procedure was weak. The follow up audit for this concern has not been completed yet, but it will be documented in Audit Report TUG-35. The above three audits pertain to maintenance and maintenance instruc-tions, however, they do not identify the same problem or a failure to provide prompt or appropriate corrective action. All three audits were of three different organizations at three different time periods. The first two audits have some similarity, but involve two different types of maintenance procedures and two different disciplines. The third audit and the identified concern is totally unique to itself. This item is considered closed.
- 3. CAT Potential Fin. . . " - - -
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This potential finding was reviewed by the SRIC and determined to be a Severity Level IV Violation. This violation was identified in Appendix A of the Region IV letter to TUGC0 dated May 31, 1983. This item will be followed by the SRIC. Design Chance Controls and Corre'ctive Action Systems
- 1. CAT Potential Findino " Contrary to 10 CFR 50, Appendix B, Criterion V, and FSAR Section 17.1.5, procedures were not adequate to assure design changes were properly transmitted to the quality control organization such that an appropriate inspection could be performed (Sections IX.B.l.b and IX.B.l.c)." 19/ (see Attachment)
The Region IV inspector has already addressed many of the concerns with design change and corrective action in previous areas of this report. The Region IV inspector has attempted to identify the specific concerrs of the CAT inspectors in Section IX of the CAT report. Many of the concerns the CAT inspectors had with regard to design change and corrective action were related to potential hardware problems in trical and mechanical areas and their concern with the ' ilized at CPSES. The ot unique with CPSES. It has been utilized at many other nuclear projects. The implementation of the iterative design process is somewhat unique at CPSES due to the type of design drawings provided by the architect engineer, G&H and by the actual time frame in which the design and construction was initiated (construction permit issued December 1974).
, 28 Because the type of drawings furnished by G&H in the pipe support and raceway support areas involved the use of an inordinate number of
" typical type" support drawings many of the design change documents have become free standing documents unique in themselves and are, therefore, design documents. This is somewhat unique to.CPSES.
However, the actual reasons and complexity of the changes at CPSES are not unique, i.e., changes in location, interferences, loading changes, etc. Another factor was the time frame in which the original design effort for CPSES was initiated. With an actual construction permit date of December 1974 and much of the preliminary design preceding that by 2 or 3 years, the commitments and requirements for design and construc-tion have changed drastically. The evolution of ANSI Standards, Regulatory Guides, NF Section of ASME, IEEE 323 and 384, and post TMI changes have caused continuous design changes. This is not unique to CPSES, but they are one of the lead facilities on the implementation of IEEE 323 and NF Section of ASME Code. Section IX.B.1.b - identifies the following concerns:
- a. Adequacy of design review (four to five thousand DCA's and CMC's) considering a Sep'tember 1983 fuel load date. '
ThehicenseehasnowextendedthefuelloaddatetoDecember 1983. The extension was partially based on their inability to complete the design review. This indicates the importance that the licensee has assigned to this activity. At this time, the completion of design review of DCA's and CMC's is not considered by the licensee to be on the critical path based.on a fuel load in December 1983.
- b. Inspected Item Removal Notice (IRN) is not required for addition of items to an inspected component.
The IRN was developed for the express purpose of tracking items that had to be removed or disassembled for construction reasons. The IRN provides a control mechanism and alerts QC that a reinspection is required. Design change ~ documents provide the tracking and control mechanism for the addition of items to an inspected component.
- c. Construction and/or engineering rather than QC determine when inspection or reinspection is required.
It would be virtually impossible for QC to try to determine the status of construction completion to schedule inspections. Construction must provide the information to QC. QC hold point and inprocess requirements are appropriately indentified by QC, but construction must provide the completion status to QC.
,k 1 .' 29 '
The various areas of QC have determined what design / construction documents they will maintain. All changes to those documents including DCA's and CMC's are sent to QC. Design changes to supports are in the support package which QC uses to inspect the item or component. The normal process is to engineer the item, build or construct the item, and then inspect the item, and the information normally follows the same process path. J d. Final inspections in the vault when there are outstanding design p changes. The Region IV inspector addressed this concern in the electrical
,,f, h
fr and instrumentation construction area of this report. The total CPSES post construction verification, design change verification, d final inspection, and document review will close the loop.
- e. Design changes are not being submitted to QC per the requirements of ANSI N45.2.11, Section 8. " ANSI N45.2.11, Section 8, states that 'p'rocedures shall be provided which assure that the impact of the change is carefully considered . . . and information concerning the change is transmitted to all affected persons and organizations.'"
ANSI N45.2.11, " Quality Assurance Requirements for the Design of Nuclear Power Plants," is limited in scope and applicability to design functions and design organizations by Section 1 of the standard. Section 8 of this standard pertains to communication , with the licensee and design organizations. It does not preclude or prohibit communications outside the design organizations, but it does not require direct communication with construction QC. QC at CPSES is on the distrubution list for the design documents that they deem necessary. QC procedures appropriately address the required document search required for the performance of inspections,
- f. CP-QP-15.4, Revision 3 " Design Change Verification," only requires a verification of 30 percent or a minimum of three design changes per drawing.-
The Region IV inspector reviewed Revision 5 of CP-QP-15.4, the 30 percent is no longer applicable. All design change documents are now being reviewed by the Design Change Verification Group (DCVG). To assure incorporation of the design change in the physical plant. This does not mean that every design change receives actual physical verification. It means that a review is made of all design changes to determine if the change would affect the physical condition of the item, then determine if the change has been implemented and inspected. This review utilizing CMC's,
30 DCA's, IRN's and all available inspection records, should indicate whether all attributes of a design change have been complied with and appropriately inspected. If further inspec-tion is required, the DCVG will transmit a copy of the open design change document to QC for further inspection.
- g. It was unclear as to who was to perform the required inspection a DCVG individual or a QC inspector, and if the individual or a QC inspector, and if the individual would meet the requirments of ANSI N45.2.6.
CP-QP-15.4, Revisicn 5, Paragraph 3.3 clarifies this concern by requiring the DCVG to notify QC of all open inspection items required. This does not preclude physical verification by DCVG, but does require a ANSI N45.2.6 certified inspector from the appropriate QC group to perform all inspections of record. The Region IV inspector was unable to identify any inspection records that were not performed by the appropriate QC group.
- h. RFIC's Versus NCR's Versus Punchlists The item is addressed in the electrical and instrumentation construction area of this report. The Region IV inspector was unable to identify any area where a discrepancy, deviation, or nonconforming condition, once identified, regardless of method of documentation, that was not properly tracked, corrected, and inspected.
- i. Design change verification does not include energized or turned over systems.
No systems are exempted from design verification, however, if the preoperational checks or testing have provided the assurance and inspection records to verify the compliance and incorporation of the design changes, then further verification is not required.
- j. QC inspection results are causing changes to design documents rather than changes to the hardware.
the hardware was changed to agree with the drawing even though the engineering evaluation indicated that the "as-built" hardware was adequate. The important objective is to ensure that the physical "as-built" hardware is adequate and meets all the requirements and that the design document reflects the "as-built" condition.
l 31
- k. VCD walkdowns do not ensure that components have been installed to requirements.
This item is addressed in the Mechanical Construction area of this report. None of the supports reviewed by CAT had received final inspection at the time of the CAT inspection. These items are considered closed.
- 2. CAT Potential Finding " Contrary to 10 CFR 50, Appendix B, Criteria II and XV, and FSAR Sections 17.1.2 and 17.1.15, nonconforming conditions identified relative to some safety-related hardware installations are not being properly documented, evaluated, and dispositioned through the corrective actiu program (Sections III.B.2 and IX.B.2) 20/ (see Attachment)
This item has been addressed to some degree in almost every other section of this report. RFIC's are no longer used to document separation problems, even though the Region IV inspector coul identify any gross misuse of RFIC's. This item is considered closed.
- 3. Summary of Region IV Followuo to CAT Inspection This report plus the Region IV letter of May 31, 1983, and the attached Notice of Violation document the Region IV followup to the CAT inspection and their potential findings as documented NRC Inspection Report 50-445/83-18 and 50-446/83-12.
The Region IV inspector did not expand the sample size of the CAT. The potential findings and specific concerns provided a sufficient sample size. The Region IV inspector reviewed the specific concerns and then l made an indepth look at these concerns. Four violations were identified by the SRIC and will be followed by the SRIC. One new unresolved item is identified in this report.
- 4. Unresolved Item Unresolved items are matters which require more information to ascertain whether they are acceptable items, violations, or deviations. One unresolved item is identified in the Electrical and Instrumentation Area of this report pertaining to the installation of NIS conduits in close proximity to fluorescent light fixtures. (8328-01) l
32
- 5. Exit Meetina The Region IV inspector and R. G. Taylor, Senior Resident Inspector-Construction for CPSES, met with R. G. Tolson and other licensee and contractor personnel on September 16, 1983, to discuss the scope and findings of this inspection. The unresolved item identified in this report was acknowledged by the licensee representatives.
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. .' l Attachment Texas Utilities Generating Company 50-445/83-28 Comanche Peak Steam Electric Station 50-446/83-14
- 1. Cable E0102532 - This cable was identified as E0120532, which is a nonexistent number at CPSES. The Region.IV inspector utilized location information to determine identification.
- 2. Battery Charger BC1ED1 Assuming the above cable number is correct, the appropriate suffix for the battery charger is -2.
- 3. Cable EG102595 - This cable was identified as a Type W-216 cable.
- 4. Tray Section T120CBD36 - Cable E0102534 does not pass through Tray Section T120CBD31.
- 5. Conduit C12030560 - Identified as T12030560.
- 6. Cable EG100032 - This cable was identified as a Type W-206 cable.
- 7. 4/0 6.9KV SHLD Cable - Identified this cable as 750 mcm. CPSES does not have any 750 mcm 6.9KV SHLD cable.
- 8. Junction Box JB1A-915Y - This is the designation for a junction box, not a conduit.
- 9. Tray Se: tion T13GACZ92 - This was identified as T13GACZ79, however, that tray section ends several feet away from the area of concern.
- 10. Sectio- II.B.3.e - Was shown as Section II.B.3.C in the CAT report.
Section II.B.3.C pertains to motor control centers.
- 11. Anchor bolt - Referred to as a Richmond insert anchor bolt. It is actually an imbedded anchor bolt. There are no 2 inch Richmond inserts at CPSES.
- 12. Support 3 - This support was identified as H-
- 13. his support was identified as i
- 14. Suppor m This support was identified as
- 15. Support S This support was identified as his support was subsequently changed to a shared Liass a support.
- 16. FSAR Section 17.1.5 - This was identified as FSAR Section 15.1.5.
l l
E Texas Utilities Generating Company
- 17. FSAR Section lA(B) - This was identified as FSAR Section 3.8.
Section 3.8 does not address Regulatory Guide 1.58.
- 18. FSAR Section 17.1 - This was identified as FSAR Section 3.8.
- 19. Sections IX.B.1.b and IX.B.1.C - This section reference also included
.IX.B.4 which did not pertain to this potential finding.
- 20. Sections III.B.2 and IX.B.2 - This section reference was identified as III.B.8, IV.B.2, and IX.B.2. Sections III.B.8 and IV.B.2 did not pertain to this potential finding.
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-k q 4'A LBP-84-8
( UNITED STATES OF AfdRICA NUCLEAR REGULATORY COMMISSION Before Administrative Judges: Peter B. Bloch, Chairman Dr. Kenneth A. McCollom Dr. Walter H. Jordan Docket Nos. 50-445 In the Matter of 50-446 TEXAS UTILITIES GENERATING COMPANY, et al. (Application for Operating License) (Comanche Peak Steam Electric Station, Units 1 and 2) . January 30, 1984_ MEMORANDUM (Records Retrieval) This memorandum discloses the Board's thinking about the adequacy of the record concerning the computerization of certain deficiency records for construction and the adequacy of the system for retrieving and utilizing these deficiency records. The purpose of this disclosure is to assist the parties in focusing on matters the Board considers important when they file Proposed Findings or submit additional relevant proof. Since the findings in this memorandum are preliminary, tentative and non-binding they may not be referenced as authority for filings and they'are not subject to motions for reconsideration. I. Computerization
~
During the June 16, 1983 hearing session, Mr. Stuart Treby, Staff Counsel, conducted a cross-examination of Mr. Ronald Tolson, designed to F01A-85-59 x M1p
e e Memorandum: 2 ascertain whether Applicants' use of punchlists (Deficiency Listings), attached to Inspection Reports (irs) constituted compliance with Part 50, Appendix B, Criterion XVI. The cross-examination begins at Tr. 8537. At Tr. 8537, line 13, Judge Bloch asked how the deficiency listings would be used to track a separation problem that affected two adjoining electrical conduits. Mr. Tolson answered that the computer "would automatically show it against both conduits." Following the hearing, the Board took a site visit to the Comanche Peak plant. Our first stop was the computer center, where we asked the operator to pull onto the screen an unresolved nonconformance that had been detected in a component by an inspection report. When the operator was unable to do that, we were directed to a second location, where the operator of the second computer also could not do it. Upon arriving in Washington after the hearing, the Chairman re-quested an explanation from Applicants. The first response, a letter of September 14, 1983, was that the report the Board had sought at the site could not be obtained from the particular system but could have been obtained from other systems. Because the Board was not fully satisfied by this answer, an affidavit was requested. In the responsive affida-vit, filed on October 11, 1983, Mr. Tolson stated that "most site groups" use the computer system for tracking open irs that require action by these groups. He also stated, however that " prior to - mid-September, 1983" . . . "some open items were entered into and tracked with the computer system." Affidavit at 2 [ emphasis added). 9
o . Memorandum: 3 We note that in his initial testimony Mr. Tolson relied on the computer system as part of his explanation of how irs were used by Applicants. However, the October 11, 1983 affidavit produced the clarification that only some deficiency listings were available in the computer at the time of Mr. Tolson's initial testimony. Hence, we conclude that Mr. Tolson's initial explanation was incomplete. II. Records Our concerns go beyond the completeness of Mr. Tolson's testimony, houever. Applicants were installing a computer system for irs for some purpose, although the purpose does not seem to appear in our record. Presumably, a computer tracking system was considered to be helpful in handling the complex mass of documents being generated by construction. Considering the lack of success in using the computer system for that l purpose , the Board is concerned about whether the manual system is adequate. Our concern is heightened by Staff documents and testimony. Mr. Compton testified that 1 Inspection Report 50-445/83-24; 50-446/83-15, ff. Tr. 8917 at 12 . concludes that computer-based data on deficiencies was inadequate to conduct as-built inspections as of April 4, 1983, and that no further inspections using computer-based data were planned. Tr. 8160. i
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'emorandum:
M 4 Ford 4tated that irs "are not dispositioned."3 Mr. Ford stated that hold tags might be needed for irs but that he did not believe they were used.4 Mr,. Beach stated that he was concerned about whether irs are properly dispositioned.5 Mr. Beach also stated that engineering approv-als ,of "use as is" for an IR item would not be stated in writing.0 " Although the staff subsequently seemed satisfied with Applicants' explanation about these matters,. we do not think our record satisfacto-rily reveals how the staff arrived at that position or whether it did an empirical check on the adequacy of the Applicants' answers. The need for a further empirical explanation is heightened by the Staff's report of the final walkdown inspection of the Fuel Building, filed October 12, 1983 (dated July 27, 1983) at 17-18, 19, finding that there was no procedural control or historical record for punchlists. Mr. Taylor, the Resident Inspector, heightened our concern by testifying that he does not know of any study of the reliability with which manual records are being used.7 Furthermore, it was t}1e subjective and undocu-mented view of the Resident Inspector that the error rate in Brown & 3 Tr. 8160. 4 Tr. 8162. 5 Tr. 8164. 6 Tr. 8180. 7 Tr. 8976-78.
1, ' . - f P Memorandum: 5 Root work, at Comanche Peak, is double the error rate on a typical nuclear plant.8 We note that our concern about the adequacy with which manual QA 9 records is being used extends also to CMCs and other design-deficiency documents that have not been computerized. Assuming that non-conformances are carefully recorded, difficulty in retrieving the documents could lead to an unacceptable level of uncorrected deficiencies. Indeeh, given the massive size of a nuclear plant, the lack of availability of an adequate report on open deficien-cies also could make adequate final walkdown inspections hard to make or to trust. FOR THE ATOMIC SAFETY AND, LICENSING BOARD Peter 8. Bloch, Chairman ADMINISTRATIVE JUDGE Walter H. Jofd,en ADMINISTRATIVE JUDGE h%Q 0 he % Kennetn A. McCollom ADMINISTRATIVE JUDGE Bethesda, Maryland 8 Tr. 8968. See also Tr. 9005. See Tr. 8955-8959. L_
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%D COMANCHE PEA OR THE. COMPLETION OF OUT. STANDING REGULATORY ACTIONS l , , . u :... l ' '. ~
MAY 1984 -
, g Acoroval:
R. DA g - Tt b % t.Th'ir, c,/s /s t. R.DeYoung',Djrector,IE Date
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H / W. Denton, Director, NRR
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Date ' R. h u.t h,.\Tvi % r. r.T, 5, g /g/ge J. T.* Coll 4ns,' Administrator ~ Date Region IV \ l F01A-QS-59 ps#$po
COMANCHE PEAK PLAN FOR THE COMPLETION OF OUTSTANDING REGULATORY ACTIONS I. PURPOSE AND SCOPE - On ifarch 12, 1984, the EDO directed t!RR to manaae all necessary NRC actions leading to licensing decisions for Comanche Peak and Waterford. A copy of that directive is included as Attachment 1. This plan establishes the program for Comanche Peak. , The purpose of this plan is to assure the overall coordination and integration of the outstanding regulatory actions regarding Comanche Peak, and achieving their resolution prior to a 11 censing decision. This plan-encompasses all licensing, insp_ection, hearing, and allecatiqndssues. Further, this plan'TdTr~ esses the scUpe of the work needed, specifles the critical path issues, identifies the responsible line organization, the schedule for completion, and (where applicable)'the need for additional resources to meet the schedule. The planned completion date for all reguLlatory actions is assumed to be _0ctober 1.1984, and resource needs are predicated on that assumption, a status report will be issued to management every two weeks starting two weeks after the approval of the plan. II. BACKGROUND - Comanche Peak Steam Electric Station Unit 1 is' in the final stages of the ' operating license. review process. The Construction Permits for Unit I and 2'were issued on December 19, 1974. Texas Utilities docketed their application for operating licenses on April 25, 1978. The Final Environmental Statement was issued September 24, 1981. The Safety Evaluation Report (SER) was issued on July 14, 1981. Because of the large number of outstanding issues identified in the SER, the staff recommended delaying the ACRS review. SER Supplement No. I was issued on October 16, 1981, and the ACRS meeting was held on November 13, 1981. The ACRS, by letter dated November 17, 1981, supported issuance of an operating license. The latest SER supplement was issued on November 23, 1983. Comanche Peak has been in a heavily contested hearing for over two years. AU but one contention have been dismissed. The remaining centention
~iiuestions the ability of the appTfHnPs Quality Assurance / Quality Control Program to prevent deficiencies in the design and construction of the plant.'
The Licensing Board has admitted many allegations of design,and censtruction deficiencies into the hearing as relevant to this contention. -
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The Applicants are currently pro.iecting a fuel load date for Unit 1 to be in late September 1984 The basis for this prn,iection was provided to the staff on May 7, 1984. This fuel load date appears achievable but allows no flexibility for unexpected avents in a very tight schedule. The number of hearing issues and uncertainty regarding the timing of the Licensing Board's initial decision may impact the fuel load.
, l. '
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III. PLAN FOR THE COMPLETION OF OllTSTANDING REGULATORY ACTIONS This plan describes the method in'which coordinated regulatory actions are to be taken by the staff to be ready to support an NRC decision
; regarding Comanche Peak licensing. As stated in the Purpose, the plan i encompasses all licensing, hearing, inspection, and allegation issues.
This summary addresses the scope of work needed, identifies the responsible line organization, the schedule for completion, and the resource needs to meet the schedule. The management organizational arrangement responsible for directing the overall effort and coordinating actions by the various involved
, offices is shown graphically in the enclosure to the EDO memorandum of March 12,1984 (Attachment 1). The management is under the overall direction of T. A. Ippolito, who reports to the Director of the Division of Licensing. The managbrs responsible for implementing and directing this organization are the following individuals:
Project Director (T. A. Ippolito)
- - - - -4 0 ELD Contact (J. Scintoll
- - - - -{ 0! Contact (8. Haves)I NRR Action Region IV Actions IE Actions (T. Novak) (R.Bangart) (R. DeYoung)
The line offices will continue to manage their own responsibilities regarding Comanche Peak in accordance with the schedule and objectives of this plan. Line office activities are to be coordinated with the program management organization via their representative as identified above. Additional resources are expected to be necessary to support licensing, hearing, and inspection issues, and substantial resources are necessary to respond to the approximately 400 allegations regardino Comanche Peak. This plan proposes the formation of an Technical Review Team (TRT) to evaluate and resolve a number of technical . issues, including i allegations, presently identified. A proposed organizational chart of the TRT is shown below. The groups identified will be assigned to evaluate and resolve technical issues and allegations that have been grouped into five technical areas: QA/QC, Electrical / Instrumentation Civil / Mechanical, Coatings, and Test Programs. The groups will be comprised of a group leader and reviewers that are specialists in the. particular technical area. -
-- - - , ~ , - . ,-- e
( . 7 L*
- t J C
^
CcoanchePeakTechnjgalReviewT$am(TRT) (' Pro.iect Director J Deputy Pro.iect Director s i L---- Office of 1 Investigations GROUP LEADERS QA/0C Electrical / Civil / Mechanical Cos' tings Test Program Records Instrumentation and Operational Readiness The staffing of these groupsfwill be drawn from the various NRC offices and/or contractors as arranged between the Project Director and line management. The TRT may be called together for a specified period of time, dispersed b'ack to the individual's parent office, and then reconstituted in whole or in part as needed to complete resolution of like issues. , The TRT will be under the direct supervision of the Pro,4ect Director. In accordance with the EDO memorandum of May 25, 1984, the TRT . organization is scheduled to be in place and functioning by June 4, 198a. Detailed guidance will be issued by the Pro,iect Director to the Technical Review Team and other participants in this effcet. This guidance will address the following: Method and approach for identification and disposition of allegations - Tracking System . Preparation of Documentation and Records Protection of Individuals Initiation cf Special NRC actions, such as Confirgation of Action Letters or 50.54(f) letters c _ Manpower accountino The basis upon which the schedules and resource estinates have been developed is that the Comanche Peak fuel load date is October 1,1984
, Fioure 1 is an overall scnedule and' Figures 2 throuch 5 are individual
-/, schedules for the resolution of Licensing, Hearing, Inspection, and Allegation,s Regulatory Actions 5 respectiively.
6
., g e '
_ . ~ _ -
The major issues, schedules, and resource estinates needed to meet the schedules are summarized as follows: A. Licensino Reculatory Actions Licensing Actions are those things resulting frem the design review of the FSAR. NRR is responsible for the resolution of these action items. The total number of outstanding action items is 37. Four of these action items are considered to have the potential for impacting the schedule. These items relate to 1) the adequacy of the TDI diesel generators, 2) the the Applicants' exemption request for relief from GDC-4, 3) review of the Cygna Rannet nf Ln indelendent assessmant ofldesion and construction a and 4) electrical equipment environrnental qualification. NRR experience with other facilities involved in complex licensing reviews (Diablo Canyon, Seabrook, and Shoreham) indicates that additional project management resources are necessary. Two additional project managers for the period from ilune-September will be needed, for a total of 8 man-months of additional effort. The technical resources presently assigned by NRR to evaluate and resolve the remaining open licensing actions are sufficient to meet the schedule shown in Figure 2. Additional IE resources are not expected to be required as the CAT inspection is complete and 0A/QC reviews and emergency preparedness reviews are essentially complete. B. Hearing Reoulatory Actions Hearing Actions are those issues in contention before the ASLB. There are three major issues each wi+.h a number of sub-issues. The three major issues are Design Adequacy and Quality Assurance, Construction Adequacy, and Construction Ouality Assurance. There are two critical path actions: Design Adecuacy-and Construction Adeouacy. The design adeouacy action concerns an IDVP being performed by the Applicants at the staff's request. CYGNA is~ performing the review for the Applicants. This is currently under review by the 9 staff. Cygna personnel actions may have contributed to be prenatifi- C-cation of inspection areas to Applicant 0A/QC personnel. Resolution /- l of this concern may make it necessary to request additional / i independent assessment activities. The critical path issue concerning Construction Adecuacy is containment liner coating (painting).
. l The res'ources presently available are sufficient to resolve all hearing actions with the exception of the critical path issues.
It is estimated that 10 man-months are required to resolve the Desion Adequacy Action, and 6 man-months to evaluate the Construction Adequacy Action (painting). The design adequacy review will reouire a team composed of IE and NRR personnel, similar to the Cygna IDVP effort. Coordination of Hearing activities is expected to be extensive and involve integrating the activities of NRR, OELD, and Region IV with the Technical Review Team. An additional senior manager (SES-level) is needed to manage this effort as it is expected that the Project Director will devote full time effort to management of the technical review team actions commencing June 4,19M. These estimates assume that the reviews will conclude that the existing circumstances are acceptable to the staff and/or no major corrective actions are required of the Applicants. Should this prove.. otherwise, additional resources will be required for resolution. See Figure 3 for Hearing Testimony Completion Schedule. C. Insoections Regulatorv Actions Inspection actions are those that assure that adequate completion of plant construction and the readiness of the Applicants to operate the plant. These actions are the responsibility of Region IV. . The total number of outstanding action items is 377. These may be grouped as follows: SER verification: 30 actions Rnutine construction inspections, preoperational test program and operational readiness inspections and startup test program: 121 actions Operating Licensing: 20 actions Open items inspections (unresolved items, violations, 50.55(e) items, inspector follow-up items and Part 21 items: 201 actions Room inspections: TBD CAT follow-up: 5 actions - All the inspection items require resolution prior to 0L issuance. Many require applicant actions prior to inspection or relate to hearing issues. Particularly significant is the retest inspection effort as the applicant plans to re-run approximately 25 preoperational tests to confirm system readiness subsequent to various modifications and design changes. Many of these tests will be witnessed by the NRC and test results will be evaluated as appropriate. Systems involved include safeguards systems, reactor protective system, service water, component cooling water, and the diesel generator. The number of inspection items represents a sizeable effort that could impact fuel load.
.~ . .
- 6-Some additional resources will be reouf red to complete the routine inspection program and resolve the many open items. It is expected that this area could require approximately a6 man-months. Considering the number of items and based on Waterford experience, the Region estimates that much of this effort can be handled with existing resources but that aoproximately 18 man-months additional resources will be required. See Figure 4 for the Inspections Schedule.
D. Alleaations Reculatory Actions The Allegation Actions are those concerns reported by various individuals, intervenors and action groups regarding the safety of construction of the plant. Concerns regarding wrongdoings, intimidation, etc. are r.et included in the. technical review team effort but are referred to OI or OIA as appropriate. To date the number of individual actions is approximately 400. These actions are grouped into specific categories to facilitate their resolution. Resolution of these actiens will involve the Technical Review Team, NRR, 01, and Region IV. The organizational group with primary responsibility for resolution of these actions is the Technical Review Team (TRT). The resources required to resolve these actions are identified below according to the Team functional group: Resource , Functional Group No. of Allegations ' Estimate (man-months) QA/QC Records 180 17 Electrical /Instrum. 5 2 Civil / Mechanical 97 17 Coatings 11 4 Test Programs 14 2 Estimated Totals 707 T The TRT effort is expected to require additional administrative support (secretarial) of approximately 3 man-months. Hence, the total TRT resource needs are 45 man-months. The total proaram for resolvino the allecations actions is a critical path item. See Figure 5 for the schedule for completion of the review of these allegations.
--- ,r-. .s , ,-----~n,. w,, w,,,,-. ,,.a
[ In addition, 97 allegations will be handled by the following offices: Responsible Functional Group No. of Alleoations Head Office 1 Intimidation 30 OI Design Pipe / Pipe Supports 19 NRR Vendor / Generic 18 NRR/IE Independent Assessment Program 7 NRR 1 Miscellaneous 23 RIV Design of pipe and pipe supports, and the Independent Assessment Program allegations will be dispositioned by NRR personnel that are handling these issues for the hearings. Intimidation allegations will require additional 0I resources, as discussed later in this section. Existing resources in the Vendor
; Inspection Branch, IE and NRR will disposition the vendor / generic allegations. Existing resources in Region IV will be responsible for the miscellaneous allegations.
E. Office of Investigation Actions 01 actions are those actions necessary to support the resolution of allegations. They involve issues where wrong-doing, intimidation, or harrassment may be involved. It is clear that with the present resources assigned to the Comanche Peak investigation (one investigator) the schedule for resolving the allegations and wrongdoing issues will not be met. We estimate several additional investigators will be required on full time basis from June through September, for a total of 12 man-months of effort. During this 4-month period OI will require the full-time support of
'one individual with a technical background, as many allegations are a combination of technical and wrong-doing issues, for a total of 4 man-months.
The NRC staff effort to complete the actions in the licensing, hearings, inspections and allegations areas will be substantial and the impact will be felt by several Offices. The foregoing summary lists a total of 821 separate actions requiring approximately 100 man-months of effort above the existing (budgeted) resources. Personnel for much of this effort will be obtained from contractors. It is estimated that approximately $1 million will be 4 necessary to fund contractor assistance in support of Comanche Peak reviews during the remainder of FY 1984 The estimates are somewhat fragile and assume that no major new issues are raised, that the Applicants meet their projected schedule, and that staff review of the identified issues will conclude that the existing circumstances, or the resolution, is acceptable.
Attachment 1
+*g.t** ****%*,,
~~ UNITED STATES ~ ~ ~ ~
! , a NUCLEAR REGULATORY COMMISSION --
. a wasHmoTom. o. c.20sas
,I MAR 121984 l
9 MEMORANDUM FOR: John T. Collins, Regional Administrator Region IV . Harold R. Denton, Director .4 '3h Office of Nuclear Reactor Regulation Richard C. DeYoung, Director -
' Office of Inspection &, Enforcement FROM: William J. Dircks .
Executive Director for Operations
SUBJECT:
COMPLETION OF OUTSTANDING REGULATORY ACTIONS ON CDMANCEE PEAK AND WATERFORD Construction of the Comanche Peak and Waterford facilities is nearing completion. There remain a number of issues that need to be reso7ved before the staff can make its licensing decisions. The issues remaining for these plants are quite complex and span more than one Office. In order to assure the overall coordination / integration of these issues and to assure issues are. resolved on a schedule to satisfy hearing and licensing decision needs, I am directing NRR to manage all necessary NRC actions leading to prompt licensing decisions. Darrell Eisenhut, Director, Division of Licensing, NRR is being assigned the lead responsibility for this activity. He will coordinate the efforts of NRR, IE, and Region IV, and will coordinate this activity with OI and OELD. Prior to any of the affected Offices undertaking major activities (e.p., inspections) or making decisions on these plants, that activity should be concurred in by NRR. We are presently in the process of assigning a dedicated senior manager to assist Mr. Eisenhut in the management of these activities. The first phase of this program wf11 be the identification of issues needed to be resolved for each plant prior to hearing and licensing decisions. Once the issues have been identified a Program Plan for resolution of each item should be developed and implemented. The Program Plan should address the scope of the work needed, the identification of the responsible line organization, and the schedule for completion. In principle, this effort will therefore be similar to the effort undertaken regarding the allegation review on Diablo Canyon except that this effort should encompass all licensing, inspection, hearing, ar.d allegation issues.
. . 1 Each affected Office will assign a full time senior manager to work with NRR to define, schedule and complete the issues. I expect these managers to be identified by each of you within a few days. All affected offices should )
t provide dedicated resources and give their full support to this effort, to assure that all existing issues are expeditiously handled and all new issues are prcmptly provided to NRR so as not to delay the licensing decisions. In addition, copies of all information, documents, depositions, etc. should be promptly provided to NRR to ensure a coordinated approach. I anticipate that the approach utilized here will be necessary for a number of upcoming OL projects, and am directing NRR to take the lead for carrying out this activit,y. Will J. Dircks Executive
- Director for Operations cc: G. Cunningham, El.D
- 5. Hayes, OI O
5 9 D l
. . ~
I c 1r IOVERALL MANAGEMENT - COORDINATION (NRR) a . ( F 01 Contact ,
. j 4,
. i E ELD Contac:
, r Overall Review of Review of Construction QA Adecuacy* (CAT) ~ '
Review of Desion ~ and Operation Issues (IE) Issues (Reg IV) . (NRR) 6 e S 6 e S e O e S e 0 4 6 o 4 O
O' FIG)RE 1
. COMANCHE PEAK Major Regulatory Actions Schedule REGULATORY '
- ACTIONS Hay June- iluly August September October Licensing (SERs) X--------------------------------------------------------------X Hearing Testimony Suhmittals to ASt.B X------------------------------------------------------X Inspections X--------------------------------------------------------------X
~
Allegations X-------------------------------------------X
- w
Fir,URE ? COMANCHE PEAK . LICENSING REVIEW SCHEDULE' Regulatory Responsible - Action Organization May June July August . September Dctober I i i i i i l l l l , I .
- l
- 1. FSAR Open Issues NRR/DE, DSI (------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - --------------
-------------------J
- 4. Issues Remain DHFS, IE (TMI issue; 3
, suh-items) i' 9L -
- 2. f
- Confirmatory HRR/DE,DSI (------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - --------------
-------------------)
/ Issues Required For Unit 1 License; 10 '
Issues Remain ' (TMI Issue; 12 sub-items)
- 3. License Conds. NRR/DE (------------- - - - - - - - - - - ' - ---------------- ,-----X
- 2 Issues Remain A. Technical Specif. NRR/DE, DSI t------------- - - - - - - - - - - - - ---------------- --------------
-------------------)
5 Issues Remain *
- 5. New Issues Raised NRR/DSI, NMSS J------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - --------------
-------------------)
by NRC; 7 Issues , , Remain. (includinn Financial Dualifications) - i !6 New Issues Raised NRR/DHFS, DE .(------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - -
------X by Applicant DL, DSI
{, 1 6 Issues Remain , 7. Dustanding Gen. HRR/DS!, IE l------------- ------------ ---------------- --------------
-------------------)
Ltrs; 3 Remain 4
FICilRE 3 - - Clif4TNCllE PEAK 2l
*01 Liaiting ** Technical Review Team (IRT) .
HEARING TESTIMDNY PREPARATION AND SilRMITTAL SCHEDULE . 1 Regulatory Responsible Action Organization May June July August September Octohe:- l l l 1 I I I I I I I I I I I I I I I
- 1. Independent .
! Assessment
- a. IAP-Cygna NRR/DE, DL, IE X----- -------------- ---X t
- b. Applicant's Plan NRR/DE DSI, IE (-------------- -------------: :
(W/D concerns) i
- 2.
- Intimidation / NRR/DL,OELD (-------------- -------------- ---------------: :
Harassment Issues ,._ . TRT (---------. -------------- -------------- ---------------:: 3.**0A/QCProgramh( Issues 4.** Pipe Support TRT (--------- -------------. -------------- - X s, ()
- Issues b tyr -
l ( Q / p*
- 5. Helding Issues NRR/DE, RIV, IE X-------
(J -- % i (Followup) (
'f 97 -
6.** Coating Issues RIV (LANL), NRR/DE (--------- -------------
------X
- 7. Vendor Generic
- a. Transamerica TDI Task Force (--------------
Delaval Diesel
------X
--)-
Generators 1 8 Miscellaneous a.** Staff Walkdown TRT t-------------:t (includes Cable Tray) i
- b. Polar Crane RIV (---------. ------X Reinspection 4 c. Mirror Shield RIV (--------- (
s FIGURE 4
, COMANCHE PEAK SCHEDULE-i
^ INSPECTION AND REGIONAL RESPONSIBILITIES 1 April May June July . August September i i i i i i i i i i i i i i i October-i i i l, Construction Inspec. X---------- -------------------------- --- .- *------ --X N , HC 2512 s!
- . Construction Inspec. X---------- --------------------------------------- -----------X i' Items Followup i 50.55(e) Items, X----------- --------- ,--- ------------- -----------X i
Part 21 Items Followup ! Primarily preoperation X Primarily Test Wit nessing,-Proc edures Review -------------------x Test - j Inspections MC 2513 Results i Review i PreoPerational 1 X----------- ------------- "------------ --------------------) j Inspec. Items i Followup i Startup Test X----------- -Procedures
----- ,----- Review
-------------.-----X X Post-OL-----------X Program Inspec. Witnessing MC 2514 4
- Operational K K X (Final Rpt)
- Readiness Report
); per Module 94300 I SER Verification XItgms_frge, }SERjs,1-4___ X X -Items from SSER 5
------------------- (
; IE Bulletins Inspec. X------------ ------------- -------------
------------x 4 Hiscellaneous X------------ ------------- ------------- ------------- ---------X Allegations Resolution 2 Operator Licensinq 'X----------------.X
FICCE 5 ' COMANCilE PEAK Allegations Resolution Schedule April May June July August September October i i i i i i i i i i i i i i i i i i i i i
- 1. Plan Approval X
?. Personnel X-- --- X Arrangemen.t
- 3. Obtain Logistic X------ --- X Support
- 4. Develop Review X------------- X Packages ,
5 Task Force X Briefing & Assignments ,
- 6. First Site X- y.--s Review Period ,
- 7. Second Site X-----X YL d
Review Period X-M)
'~
8 Third Site X X
- Review Period
- 9. Prepare Final Draft SSER Xr8 X-p 10 itanagement Review ThX TM
- 11. Final Report X----X
- 17. SubmitRe/r X to ASLR
N F I k, x i
,g '
/
( / In Reply Refer To: f Dockets: 50-445/84-40 M gg g 50-446/84-15 pi Texas Utilities Electric Company ATTN: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Gentlemen: This refers to the inspection conducted under the Resident Inspection Program by Mr. J. E. Cumins and NRC contract personnel during the period October 21, 1984, through December 18, 1984, of activities authorized by NRC Construction Permits CPPR-125 and CPPR-126 of the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. J. T. Merritt, and other membets of your staff at the conclusion of the inspection. Areas examined during the inspection included plant statu:;, action on previous NRC inspection findings, action on licensee identified design construction deficiencies (10 CFR Part 50.55(e)) reports and plant tours. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection report. During this inspection, it was found that certain of your activities were in violation of NRC requirements. Consequently, you are required to respond to t51s violation, in writing, in accordance with the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this letter, g<)
/
RPB2/8h C/RPB28 ' C/RPB ' NR JECumin's/vs OHunnic t 0Hunte VNkha == i',, /3 /M ' i /O' / '" t /A /p \ A /e 1 is# 4
.m.
I4 . ( .- Texas Utilities Electric Company Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely,
" Original signed M 0.R. HUNTER
- D. R. Hunter, Chief Reactor Project Branch 2
Enclosures:
- 1. Appendix A - Notice of Violation
- 2. Appendix B - NRC Inspection Report 50-445/84-40 50-446/84-15 cc w/ enclosure:
Texas Utilities Electric Company ATTN: B. R. Clements, Vice President, Nuclear Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Texas Utilities Electric Company ATTN: J. W. Beck, Manager, Licensing Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 bec to DMB (IE01) bec distrib. by RIV:
"RPB1 *RRI-OPS TX State Dept. Health
*RPB2 *RRI-CONST. Juanita Ellis
*EP&RPB *R. Bangart, Task Force Renea Hicks
*J. Gagliardo (CPSES) R. Martin, RA (1tr only) Billie Pirner Garde C. Wisner, PA0 *0. Hunnicutt, Task Force
- MIS System R. Denise, DRSP V. Noonan, NRR "RSTS Operator
*RIV Flie S. Treby, ELD T. Westerman, E0
*w/766
f~ O APPENDIX A NOTICE OF VIOLATION Texas Utilities Electric Company Dockets: 50-445/84-40 Comanche Peak Steam Electric Station, 50-446/84-15 Units 1 and 2 .. Construction Permits: CPPR-126 CPPR-127 Based on the results of an NRC inspection conducted during the period of October 21, 1984, through December 18, 1984, and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 49 FR 8583, dated March 8, 1984, the following violation was identified: Failure to Assign a Unique Identifier to a 10 CFR 50.55(e) Item 10 CFR 50, Appendix "B", Criterion V requires that, " activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings." Non-ASME QA/QC Procedure CO-QP-16.1, Revision 5, Section 3.4 has been established in accordance with Criterion V and requires that a unique sequential identifier be assigned to reportable 10 CFR Part 50.55(e) items. - Contrary to the above, on October 26, 1984, the NRC inspector observed that the identifier CP-83-17 had been assigned to two different items. The first item, ID number 117, was identified July 6,1983. This item involved potential tiltout subassemblies by RTE-Delta. The second item, ID number 118, was identified July 12, 1983. This item involved relief valves for containment cooling water side of the spent fuel pool cooling heat exchangers which were set at an incorrect back pressure. This is a Severity Level V violation. (Supplement II.E) (445/8440-01) Pursuant to the provisions of 10 CFR 2.201, Texas utilities Electric Company is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations: and (3) the date when full compliance will be achieved. Consideration may be given to extending you response time for good cause shown. Date: g A w, h
r APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-445/84-40 50-446/84-15 Dockets: 50-445; 50-446 Construction Permits: CPPR-125 CPPR-126 Licensee: Texas Utilities Electric Company (TUEC) Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 Inspection at: Glen Rose, Texas Inspection conducted: October 21, 1984 through December 18, 1984 Inspector: . k' > urb d /2/h 7/I'/ k.*J. E. Cumins, Senior Resident Reactor Otte '
/g Inspector. Construction 4
NRC Contract Personnel: R. P. Evans, Project Engineer, EG&G Idaho Approved: b 'E ? h n> N a :l t '~ /2/1'.T/9'l D. M. Hunnicutt, Section Chief Date Reactor Projects Branch 2 Inspection Summary Inspection Conducted October 21, 1984, through December 18, 1984 (Report 50-445/84-40) Areas Inssected: Routine, announced inspection of plant status, action on previous 4RC Inspection findings, action on licensee identified design / construction deficiencies (10 CFR 50.55(e) reports), CPSES inspection force personnel meeting, and plant tours. The inspection involved 144 inspector-hours onsite by one NRC inspector and one NRC contract person. Results: Within the five areas inspected one violation was identified (failure to follow procedural requirements for assigning a unique identifier to a 10 CFR 50.55(e) item, paragraph 4), gy-( tI
- 4 Inspection Sunnary Inspection Conducted October 21, 1984 through December 18, 1984 (Report 50-446/84-15)
Areas Inspected: Routine announced inspection of plant status, action on licensee identified design / construction deficiencies (10 CFR 50.55(e) reports), CPSES inspection force personnel meetings and plant tours. The inspection
. involved 17 inspector-hours onsite by one NRC inspector.
Results: Within the four areas inspected, no violations or deviations were identi fied.
, DETAILS
- 1. Persons Contacted B. R. Clements, Vice President, Nuclear Operations TUGC0
*J. T. Merritt, Assistant Project General Manager "A. Vega, Site Quality Assurance Manager R. Baker, Staff Engineer, W. Katness, Quality Engineer, Brown & Root (B&R)
- 0. 8. Jones, Unit 2, TF Engineer, TUGC0 C. R. Hooton, Civil / Structural Engineer Supervisor, TUGC0 P. Chang, Pipe Support Engineer Supervisor, TUGC0 J. Wythe, I&C Engineer, TUGC0 J. D. Hicks, i
The NRC inspectors also contacted other plant personnel during this inspection period.
- Denotes those attending one or more exit interviews.
2, Plant Status , Unit 1 At the time of the inspection, construction of Unit 1 was 98 percent complete. The fuel loading date for Unit 1 is pending the results of ongoing NRC reviews. The licensee continues to complete and turnover systems and areas from construction to operations. The turnover process is accomplished in two phases. The first phase is accomplished when construction completes a system or area and turns that system or area over to the startup group. The turnover process is completed for a system or area where operations makes final acceptance of the system or area from the startup group. The table below shows the status, as of December 13, 1984, of the 422 distinct areas identified by the licensee for turnover from construction to operations: Total number of areas 422 Number of areas submitted to startup 403 Number of areas accepted by startup 403 Number of areas submitted to operations 403 a
- . - - , ..-,---,----e, -,n, em. , - , - - -,-,,,www_>ae_- r--, ,,,w-a-- ..n-+ ,,-e --,- - , , - -mm,-.,,-,---,,- ,-,, . - -- - - - - ., , -,- -m
Numbers of areas accepted by operations 117 The table below shows the status, as of December 13, 1984, of the 332 distinct subsystems identified by the licensee for turnover from construction to operations: Total number of subsystems 332 Number of subsystems submitted to 332 startup Number of subsystems accepted by 332 startup i Number of subsystems submitted to 307 operations Number of systems accepted by 253 operations Unit 2 At the time of this inspection construction of Unit 2 was approximately 68 percent complete with fuel loading scheduled for approximately 18 months after Unit 1 fuel loading.
- 3. Action on Previous NRC Inspection Findings
- a. (Closed) Unresolved Item 445/8408-04: Tack welding of polar crane seismic connection shins.
Licensee's DCA 9872 was revised (Rev 4) to modify a note allowing the tack welding on one or both sides of the shims in the seismic connection. The licensee actions appeared to be acceptable,
- b. (0 pen) Severity Level IV Violation (Supplement 11.0) 445/8408-01:
Gaps on Unit 1 polar crane bracket and seismic connections exceed design requirements. Licensee's DCA 9872 was revised (Rev 4) to add the following statement:
"A review of these existing conditions has been performed and the gaps are acceptable without further action. Observations show the gaps tend to have marginal movement. These changes in gap width and location are minor and are able to open only as wide as the bolted connection allows." The NRC inspector reviewed Gibbs & Hill, Inc.,
letter (GTT-10457) dated August 3, 1984, which substantiated the OCA revision to the extent that the existing conditions were acceptable. This item will remain open pending further review during a subsequent inspection. _ .-,c_ - . - _ - . _ _ , . ~ _ . , , . ......,._,.,___,_._.._,.,,.,_--_,.._%-_,~,.__.__.-.,_,__,,r,,.,__,___.~.
. . . .. - -. . ~ . .. . . . - _ - . .. - - - ..-.
4 j 5-
- c. (0 pen) Deviation (50-445/8408-03): Failure to Implement any Locking Device to Prevent Nut Backoff - The NRC inspector reviewed DCA 1090 Rev 1, which allows the use of multi process epoxy coating to prevent nut backoff. An in place inspection of Platform OP-11 by the NRC inspector was performed. This item will remain open pending further review during a subsequent inspection.
- 4. Action'on Licensee Identified Design / Construction Deficiencies 1
(10 CFR Part 50.55(e) Report
- a. The 10 CFR Part 50.55(e) reports discussed below were reviewed for content, compliance with NRC requirements for reporting, appropriate evaluation, and adequacy and implementation of corrective action.
Each 10 CFR 50.55(e) report is identified and tracked by the unique licensee assigned number shown at the beginning of each discussion. (1)CP 84-09 Insufficient lubrication of Delaval turbocharger thrust bearings could cause potential probisms. Licensee letter TXX-4227 dated July 12, 19P4, reported to the NRC that evaluation of.this deficiency had - determined that it was not reportable. Corrective action was to rework the oil drip system in
! , accordance with Transamerica Delaval's Orawing 102675.
Inspection Report I-1-0054304 documents completion of this work. (2)CP 84-16 Several of the main steam relief valves may not have been qualified for actual loading conditions. Licensee letter TXX-4330 dated October 8, 1984, reported to the NRC that evaluation of this deficiency had determined that it was not reportable. The licensee had subjected the valves to an operability I test, which the valves passed. This testing was
- reported in Valve Testing Test Report #A-655-84 (September 14, 1984.)
(3)CP 84-17 Potential flooding problems due to overflow of the sumps. Licensee letter TXX-4263 dated August 16, , 1984, reported to the NRC that evaluation of this deficiency had determined that it was reportable. The licensee's corrective action was to modify the sumps such that sump drains were isolated to preclude . backflow. The corrective action is documented by , DCA 21075, Revision 0, and Startup Work . Authorization 23205. In addition, a complete ! reanalysis of flooding as a result of backflow from i sumps is being performed by the licensee. 4
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t (4)CP 84-18 Electric motors on.four motor operated valves were not as specified nor were they qualified to the' required specifications. Licensee letter TXX-4308 dated September 28, 1984, reported to the NRC that evaluation of this deficiency had determined that it was reportable. The licensee had replaced the unqualified motors with qualified motors. NCR E-84 100064 acts to track this replacement. (5)CP 84-19 Transamerica Delaval, Inc., reported to the NRC on a Part 21, the failure of a valve spring in a diesel generator on a non-nuclear, marine engine application. Licensee letter TXX-4270 dated August 20, 1984, reported to the NRC that evaluation of the Part 21 informatien had determined that it was not reportable
, since the failure of the spring was an isolated case ,
of material surface damage and not generic. (6)CP 84-20 Transamerica Delaval, Inc., reported to the NRC on a Part 21, the failure of a fuel injection pump, a component of a standby diesel generator, at another nuclear site. Licensee letter TXX-4309 dated September 21, 1984, reported to the NRC that evaluation of this deficiency had determined that it was not reportable. Failure Analysis Report FA 84-007 indicated that this was an isolated case not affecting the ,
. Itcensee. *
(7)CP 84-21 Some piping was found by the licensee not to meet the required specifications in that insufficient ultrasonic testing (UT) had been performed by the vendor. Licensee letter TXX-4292 dated August 30, 1984, reported to the NRC that evaluation of this deficiency had determined that it was not reportable. The licensee conducted the required additional UT, documented on NCRs M-14730N through M-14741N, on the installed piping. The UT indicated no problems in the installed piping and therefore would not have affected adversely the safety of operation. (8)CP 84-25 Certain of the air operated valve assemblies had natural frequencies less than that specified by the equipment specification. Licensee letter TXX-4318 dated September 27,1984, reported to the NRC that evaluation of this deficiency had determined that it was not reportable. The licensee reanalyzed the stress problems documented in letters CPPA-40664 through CPPA-40668, in which these valves were used and concluded valve operability was not impaired.
l (9)CP 83-11 Component Cooling Water System was determined not to j have auto isolation. Licensee letter TXX-3690 dated June 21, 1983, reported to the NRC that evaluation of this deficiency had determined that is was reportable. Licensee corrected action, documented in OEI-I-251, Revision 0, was to add controls to sense depletion of the surge tank and to upgrade components to maintain seismic integrity.
- b. . Selected NRC inspector findings in the area of 10 CFR 50.55(e) reporting documentation are discussed below:
The NRC inspector determined from a review of TUGC0 documentation that on June 15, 1983, Texas Utilities received notification from Transamerica Delaval, Inc. , of a potential problem with the RTE-Delta potential transformer tiltout subassembifes which are used in the emergency generator control panels. RTE
- Delta had originally reported the problem as a 10 CFR Part 21 report to the NRC on April 21, 1982. On July 6, 1983. TUGC0 issued a significant deficiency analysis report" (SDAR),numberCP-83-17,totrad'thispotentialdeficiency.
On July 11, 1983, SDAR CP-84-17 Vas voided and SDAR number CP-84-17 was later assigned to a different unrelated deficiency. Section 3.4 of TUGC0 procedure Cp-QP-16.1, Revision 5, Significant Construction Deficiencies, states that the SDAR number is a " unique sequential identifier". The ' reassigning of SOAR number CP 83-17 to a second SOAR is an apparaa+, W htina of this nrocedura_1 requirement. (445/8440-01). in addition, there was no documentation saae avaiianie so sne m'* NRC inspector to indicate that TUGC0 had taken any subsequent action to evaluate the potentially deficient condition or that ~. any corrective action had been taken. 4n October 1984 Tvhen the Nis inspecsar pursuea sne w.;t;r,-the item was added to the '-- Master Data Base. This is an unresolved ites (445/8440-02) pending the determination as to what corrective action is required (if any) to correct the potential deficiency.
- 5. Meeting of CPSES Inspection Force Personnel On December 14, 1984, the NRC inspector attended a meeting between
- 8. R. Clements, TUGC0 Vice President for Nuclear Operations, and CPSES inspection force personnel. Mr. Clements had held similar meetings in an attempt to discuss and clarify the reporting of deficiencies in the plant.
Mr. Clements stated the following to the attendees:
- a. They should write a nonconformance report (NCR) if they identified anything wrong and should question any NCR evaluation if they did not think the evaluation adequately addressed the problem.
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- b. There is,an ombudsmarvon site to discuss any problem they may have, and his pffice ha&b'een relocated to make it more accessible and to provide more 1Wfvacy to any employee wishing to talk to him. The
- ombudsman's name and his office location were provided.
- c. They could contact the Nuclear Regulatory CommissioMNRC) and 7 discuss any problems they have. . ,L/_
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- d. A " safe team" will soon be onsite. T " safe team"A s going to be an independent group whose function will[he toJvestigate a concernswhileprovidingconfidentialithotheemployee.
The attendees were provided an opportunity to ask questions and to discuss any concerns or problems they had. After the meeting, the NRC inspector toured the offices that have been built for the " safe team". The offices are located adjacent to the route the majority of the construction employees have to use when entering and leaving the CPSES site. No violation or deviations were identified. i 6. Plant Tours (Units 1 and 2) l At various times during the inspection period, the NRC inspector conducted general tours of the reactor building, fuel building, safeguards building, electrical'and control building, and the turbine building. During the l tours, the NRC inspector observed housekeeping practices, preventive i maintenance on installing equipment, ongoing construction work, and discussed various subject with personnel engaged in work activities. No violations or deviations were identified.
- 7. Exit Interviews The NRC inspectors met with members of the TUEC staff (denoted in paragraph 1) at various times during the course of the inspection. The scope and findings of the inspection were discussed. The licensee acknowledged the findings.
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2 TABLE OF CONTENTS I Executive Summary and Conclusion II Background III Review Approach IV Review Findings A. - Management Organization B. - Quality Assurance / Quality Control C. - Equipment Turnover and Preoperational Testing D. - Electrical E. - Design Activities / Control F. - Installation of Safety Related Fluid Systems
-G. - Civil Construction H. - Heating, Ventilation, and Air Conditioning Systems I. - Formal L terviews with QA/QC Personnel (General) 9 .s 6
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- 1 I
- 3 I. EXECUTIVE
SUMMARY
NRR in coordination with the Director of IE and the Region II & IV Admini-strators formed a team to perform a limited unannounced review of Comanche Peak. The purpose of the review was of 1) evsluate the current implementa-tion of the applicant's management control of the construction, inspection and test programs, 2) provide an indepth understanding and background information to the NRC new management team established by the Executive Director for Operations memorandum of March 12, 1984, and 3) obtain informa-tion necessary to establish a management plan for resolution of all out-standing licensing actions. The team consisted of eight reviewers, a team leader and team manager. The reviewers 'and team leader were selected from the Region II staff. The manager'was the NRR Comanche Peak Project Director. The team was assembled in Region II headquarters where it was briefed by NRR, IE and ELD. . The team conducted its review from April 3 to April 13, 1984, The review consisted of an audit of significant elements and processes of the appli-cant's management control in construction, inspections and testing of systems important to safety. These included:
- 1. Component and material receipt inspection and control.
- 2. Structure, systems, and component fabrication and installation.
- 3. Structure, system, and component acceptance, and preoperational testing.
- 4. Quality assurance and control documentation and procedures to effect
, items 1 through item 3 above.
The portions of the system evaluated included piping, pipe and ecmponent supports, instrumentation and control, electrical cable separation and cable tray supports, component qualifications, and allegations relating to these areas. The reviews also included briefings from the Applicants' management and interviews with QA/QC, Document Control, and craft personnel. The total effort was conducted with little or no advance notice of areas, personnel or documentation to be reviewed. Each member of this team was chosen because he had both many years experi-ence in the discipline he was reviewing, and he had performed evaluations at a wide range of nuclear facilities. The team spent over 800 hours per-forming this review. The following is a list of the special review team members, their positions, and field of expertise: Paul Bemis, Section Chief, Management Organization, Qualification and Training Paul Fredrickson, Project Engineer, Quality Assurance / Quality Control, Bill Orders, Senior Resident, Preoperation and Startup Kim VanDoorn, Senior Resident, HVAC and QC inspector interviews 4 m
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- 4 Al Ruff, Reactor Inspector, Electrical Louie Jackson, Reactor Inspector, Quality Assuran:e/ Quality Control Winston Liu, Reactor Inspector, Design Activities / Control Ed Girard, Reactor Inspector, Welding and Metallurgy Joseph Lenahan, Reactor Inspector, Civil and Structures The teams findings indicated that the applicants management control over the construction, inspection, and testing programs is generally effective and is receiving proper management attention. The findings identified three potential enforcement actions (See Sections B&E); two areas of weakness requiring Applicants management attention; (See Section B) and seven areas where Applicants activities exceeded normal and accepted practice (See Sections A, B & E). The team also found improvements in the relationship between the current QA/QC management and inspectors which in the past has caused communication problems (See Section I). The team believes that the results of this limited review reveal the plant is being built in a safe manner.
The findings and conclusions of this report of the teams review should not be construed as resolving any of the issues identified by the ASLB hearings, allegations, or staff concerns of the design adequacy of the plant. II. Background On March 17, 1984, the EDO dir.cted NRR to manage all NRC actions leading to licensing decisions for Comanche Peak and Waterford. The purpose is to assure the overall coordination and integration of the outstanding regula-tory actions and achieving their resolution prior to a licensing decision. This effort is to encompass all licensing, hearing, inspection and allega-tions issues. Soon thereafter, the newly established Comanche Peak project team found that there was a need to 1) obtain current information relative to the management control of the construction, inspection and test programs and 2) obtain information necessary to establish a management plan for resolution of all outstanding licensing actions. To help achieve this objective expeditiously and objectively it was decided that an unannounced review of Comanche Peak plant was necessary. As a consequence, NRR in coordination with OIE and the Region II and IV Administrators formed a review team. Because of resource limitations in Region IV, the team was staffed with Region II personnel. The team was assembled in Region II Headquarters on April 2, 1984. The team was briefed on significant issues raised as a consequence of the licensing review, the hearing contentions and the allegations. The team leader and the reviewers were not provided with the names of the allegers in order to assure their confidentiality. The team conducted their review from April 3 to April 13, 1984. ~ _-__1_._ __:..._. ._______.___._2
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III. Review Approach The teams' review approach was to first obtain an understanding of Comanche Peak management ,and management control systems. This was accomplished by
, briefing from the Applicants management.
With this understanding, the team reviewers commenced their efforts. These included examinatten of appropriate documentation, formal and informal interviews of plant personnel, and specific technical allegations related to their areas. The allegations were not reviewed separately but were subsumed in the total review in order to provide further assurance of alleger confidentiality and not compromise any on going or future investigations. l
;In addition to the review of the Quality Assurance program, from a program- , matic point of view, each of the reviewers examined the implementation of the QA/QC program in their individual areas of expertise in an attempt to identify any breakdowns that could exist in a narrow area.
IV Reiliew Findings The team conducted it.s review of the following areas: A. - Management Organization B. - Quality Assurance / Quality Control C. '- Equipment Turnover and Preoperational Testing D. - Electrical
- E. - Desigit Activities / Control F. - Installation of Safety Related Fluid Systems
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G. - Civil Construction ' l H. - Heating' Ventilation, and Air Conditioning Systems l I. - Formal Interviews with QA/QC Personnel I !_ The review, findings and conclusions in each of these areas are provided below: _ A. Management Organization The construction and operations organization were reviewed to insure a working relationship between the organizations as well as functional relationships within each organization. The qualifications of the individuals in positions of authority were reviewed against regulatory standards and the applicant's commitmsnts. In addition to qualifica-tions, a review was made of the interface between all levels of the command chain. The limited review revealed that in all areas, individual qualifications appear to meet requirements, the interface between construction and operations appears to be functioning in a workable manner and interface between all levels of the management chain appears to be functioning in an acceptgble manner. There appears to have been a cormunication
6 problem in the onsite QA/QC chain in the past, but according to interviews conducted during this review the problem has and is being corrected. This review found the management and craft at Comanche Peak appear to be competent and management to possess a positive attitude which is a strength at this project. Management exhibited a sufficient level of consciousness for both safety and employee concerns. These management attitudes were confirmed by the attitudes they manifested in their employees and the attention to detail in the required quality of work. B. Quality Assurance / Quality Control The following areas were reviewed primarily from a programmatic point of view: nonconformance control; training, audits; records (maintain-ability and retrievability); document control; receipt, storage and handling of materials; and procurement. Within the areas reviewed,.there were several findings identified. The following is a brief description of each according to category:
- 1. Potential Enforcement Issues.
a) ASME record packages were not being maintained in a fire proof container. b) At least two vendor audits had not been performed within the required time period. .
- 2. Weaknesses a) Certain drawing packages issued to the field contained non-applicable DCAs and/or CMCs, which had been deleted by engineering.
b) Many non-ASME Section 3 drawings contained a large number of DCA's and CMC's (over 300 in some cases) outstanding without being incorporated by revision.
- 3. Strengths a) The QA/QC training program is extensive and comprehensive, b) The use of a recently established computer system drawing control instead of stamped drawings referencing design changes.
c) The vendor witnessing program is extensive in its audits and source inspection of purchased materials
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, 7 d) The ability to expeditiously locate and retrieve records, i without prior notice, from permanent records vault.
C. Equipment Turnover and Preoperational Testing The processes of turnover of safety related equipment from construction to startup as well as pre-requisite and pre-operational tests of the equipment were reviewed to determine adequacy of: methodology employed in turnover of equipment to startup, return equipment to construction for rework, and ultimate release of equipment to operations; technical and administrative controls over preoperational testing; and preopera-tional test procedures, both technical content and administrative control. This review found the majority of the tests to be performed are retests or reperform's and could be conducted in parallel with the remaining initial test. The performance of the remaining test should not impact an October 1984, fuel load date. In addition, the turnover methodology and control of the preoperational test program appears adequate. D. Electrical The assessment in this area was to determine acceptability of the safety related electrical equipment installed and inspected in accord-ance with NRC requirement and applicant commitments. A review was made of the overall program to include: drawings, procedures, quality control inspections, and records. E. Design Activities / Control This review focused on the following areas: requirements of IEB 79-02; IEB 79-14: Alternate Analysis for small bore piping system, rigorous analysis for safety related piping systems; review of design calcula-tions for pipe supports; review of stress analysis for piping systems; field inspection and verification; and the iterative design process. I 1 nere was a so a strengtn ioenti- l fied in that the applicant was found to have used conservative considerations in many areas of design and analysis for the safety related piping systems and pipe supports. l l i
l-8 The review concluded that the design program and its implementation appear to meet or exceed requirements, except as noted above. F. Installation of Safety Related Fluid Systems The review of this area was directed towards assessing the adequacy of installation of safety related fluid systems used for safe operation and shutdown of the plant. This review contained: first hand observa-tion of systems by the reviewer; examining control of welding materials; examination of piping supports, weids and records. The reviewers concluded that the applicants program appears to assure compliance with requirements, commitments and good engineering practice. G. Civil Construction Activities Examination of site civil design activities, including design change process, procedures and QA records of completed work activities (such as the SSI dam, cable tray supports and whip and moment restraints), and procedures and work activities for ongoing work (such as applica-tion of protective coating) was performed. The limited review found that the applicant was meeting requirements in these areas. Two areas of note: (1) protective coatings and (2) thermo lag, appear to be progressing in a manner such that they will not impact an October fuel load. H. HVAC This effort followed up on previously identified discrepancies at Comanche peak and other sites which used the HVAC vendor. In all areas reviewed where discrepancies had been identified the applicant appears her through rework or reanalysis. The
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I. Formal Interviews with QA/QC Personnel Formal interviews of five (5) management / supervisory personnel and twenty-eight (28) inspectors were conducted to assist the team in assessing quality of work and management support of quality. It was felt discussions with inspection personnel would give a conservative insight into the quality of site construction. The major thrust of the interviews was to determine if; (1) the . personnel had any plant safety or quality concerns; (2) intimidation was experienced; (3) training was adequate; (4) inspectors could freely talk to NRC; (5) management supported problem identification; (6) was there feedback on identified problem evaluation.
, 9 With the exception of two inspectors who were " unsure" due to lack of knowledge, all personnel interviewed felt the plant was being built in a safety and quality manner. There were some concerns raised which will be forwarded to the Comanch Peak Project Director for evaluation; in some cases, Region IV was already aware of the concerns and performing followup. The major problem in the past appears to have been communication between inspectors and their supervision, but it is apparent that for the past couple of months and presently, this problem is being addressed properly.
In addition to formal interviews, each reviewer performed numerous informal interviews to determine problem areas. The overall conclusion from all interviews was that the Comanche Peak Project is being built safely and with quality. V. Conclusion The purpose of the special team review has been met in that (1) an assessment of the applicant's current management control of the construction, inspection and test programs has been made; (2) an in depth understanding has been achieved and (3) information has been obtained to establish a management plan for the resolution of all outstanding licensing actions. With respect to the assessment of the applicant's management control of the construction, inspection and testing programs, the special review team has determined that based on the number and significance of the strengths vs weaknesses identified in this review, that the applicant's programs are being sufficiently controlled to allow continued plant construction while the NRC completes its review and inspection of the facility. Further, the review provided a sufficient understanding of these programs and their strength and weakness to assist in the development of the
" Comanche Peak Plan for the Completion of Outstanding Regulatory Actions."
This plan was approved for implementation on June 5,1984. O
t A. Management Organization
- 1. Entrance Meeting The afternoon of April 3 the special review team arrived onsite unannounced. The team spent the afternoon of April 3 and the morning of April 4th meeting with the applicant's Senior Corporate Management, Site Management, Site QA Management, and Document Control Supervision
< being briefed on the organization, functions, and location of areas under their control.
- 2. Management Organization f
The nuclear portion of Texas Utilities Generating Company is organized l in the following manner for its senior management: a) The highest level executive is the President of the company. The ! President has recently turned over all possible non-nuclear duties to his Executive Vice President-Plant Operations. The President's primary responsibility is to complete the Comanche Peak Steam Electric Station as safely and expeditiously as possible. b) Reporting. directly to the President are the Executive. Vice President . Engineering and Construction and the Vice President Operations. Even though there are fossil plants presently being built in the system and the licensing organization reports to the Executive V.P. Engineering and Construction he spends between 60-80% of his time at the Comanche Peak Site. He has also delegated his non-nuclear responsibilities in an effort to focus on the nuclear station completion. The Vice President-Operations
-(V.P. OPS.) spends approximately 80% of his available time on site directly observing the operations group preparation to take over the . plant upon construction completion. He is also an active participant in construction and startup meetings and the decision making process. A few months ago the V.P.0PS. was moved from his normal reporting path to Executive V.P.-Plant operations, directly reporting to the President.
c) Reporting to the Executive Vice President Engineering and construction is the Vice President Engineering and Construction (V.P.E.&C.). The V.P.E&C. has been located on the Comanche Peck site since 1977 and during the same year he assumed the additional title of Project General Manager for Comanche Peak. In January 1984 he delegated his non-nuclear responsibilities in order to devote his full attention to Comanche Peak completion.
'd) The Assistant Project General Manager (APGM) reports to both the V.P.E&C and the V.P.0PS. He reports to the V.P.E&C. in the areas of construction and onsite engineering and to the V.P.0PS for startup (S/U). This position is where the common tie between
. 2 construction and operations is most decisive. The APGM has been
- on site since 1977.
e) In addition to the APGM, the V.P.0PS has reporting to him: The , manager of Nuclear Operations, who is located at the site, and the Manager of Quality Assurance who is located in the corporate office but has a Quality Assurance / Quality Control Manager on site who is responsible for all QA/QC on site. 2 The current positive management attitude is a strength exhibited at Comanche Peak from both the . operations and the engineering and construction sides of the company. This positive attitude appears to manifests itself in the attitudes of the workers, the training, and in its consciousness for quality. i One additional strength was noted in that the applicant is using operations' maintenance procedures to perform periodic maintenance on equipment in the plant, and the applicant is using full Anti-C dressout i and respirators for the craft (for training) to perform maintenance activities so when the equipment becomes contaminated the workers will be use to the confining clothes and equipment. This practice should significantly reduce exposure and therefore dose received by these individuals after the plant is operational, i 3. Project Management Meeting . Every Saturday morning a project management meeting is held, wherein ,* work activities, . progress, startup and test problems, and QA/QC coverage is discussed. -This meeting is attended by Senior Corporate Management; including the . President of Texas Utilities Generating Company, and the Senior management from construction and operations; it is also attended by the site management of construction and startup. Several members of the review team attended this meeting on April 7, 1984. The meeting appeared to be well managed, with problem areas being openly discussed (even though senior company management and NRC were in attendance,- the dialogue between individual managers and supervisors was not toned down). An example of an area of concern which was discussed was the completion of the application of protective coatings in the containment. It was the general consensus- that additional manpower was required to complete the work effort. An additional 100 people were authorized with the expectation they would be available within one week. During .this meeting it was decided to change the concept that was presently being used for plant completion. The applicant had been using a Building completion methodology, but after consultation and reviews by an acknowledged industry expert is was decided to prioritize , systems completion, with buildings to follow, or run in parallel where possible. l t
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. 3 The highest level of the Company's management in attendance at this meeting allows for immediate decisions to be made for the next weeks priorities for plant completion. This method of holding project meetings appears to have kept the applicant in position to meet their projected fuel load date.
B. Quality Assurance / Quality Control
References:
CP-QAP-16.1, R20, Control of Nonconforming Items CP-QP-16.0, R13, Nonconformances CP-QP-16.1, R5, Significant Construction Deficiencies CP-QP-17.0, R3, Corrective Action CP-QP-15.7, R2, Tracking of Audit Reports /Correc-tive Action Reports
- a. General This portion of the review was performed to verify that:
nonconformances are being identified items were considered for reportability to NRC corrective action prevented recurrence the licensee has an adequate trending program
- b. Review Effort The reviewer selected NCRs from various safety related systems to verify the following:
logged numerically for control maintained even when later cancelled
- considered for reportability to NRC corrective action initiated which prevented recurrence considered in a trending program The following NCRs were reviewed:
C-84-01030 M-83-01162, R2 M-84-00965 M-11678N M-82-01528, R2 M-11660N M-83-01454, R1 M-11675N M-04729, R1 M-11687N M-05689, R0 E-84-01031 M-06244, R1 M-01695N M-09765 M-01692 M-09766 M-09812S, R1
4 The responsibility for closing NCR M-09812 S, R1, has been transferred to TUGC0 startup because these Westinghouse valves are required to be disassembled during system flushing. The valves are to be reassembled under a startup work authori ation (SWA). Valve stroke time testing of these valves will be verified under the SWA. The relief valves listed on NCRs M-09765 and M-09766 were required to be reset because the vendor had not been furnished the correct back pressure information to set the valves. l
- c. Conclusion
References:
CP-QAP-2.1, RIO, Personnel Training and Qualifica-tion QI-QAP-2.1-1, R6, Nondestructive Examination Personnel Certification QI-QAP-2.1-5, R5, Training and Certification of Mechanical Inspection Personnel
- a. General The purpose of this part of the review was to verify that the licensee has:
a formal training program conducted required training to qualify personnel ( requirements for on-the-job training ( objective evidence of personnel qualifications evaluated the candidate's education, experience, and training V k I). i prior to certification reevaluated personnel on a periodic basis records of personnel qualifications
- b. Review Effort N
A review was made of the documents listed above, and the reviewer held discussions with responsible corporate and site personnel to verify that procedures are consistent with regulatory require-ments. A review was made of General Examination Tests, RT-II-G-A, UT-II-G-B, PT-II-G-B, and MT-II-G-F; also Practical Examinations MT-II-P-04 and PT-II-P-07. These examinations confirmed the tests to meet the reouirements o N ecommended Practice. The records of M inspectors were reviewed. The records contained objective evidence of QC inspectors qualifications by
- l. ~
e 5 general and practical examination, on-the-job training, classroom, specialized training, education, and work experience records were availabl meet the requirements of Confirmation of annual documented eva uar.1ons o qua ifications of inspectors was verified.
- c. Conclusion T '
mplete. When personnel were questioned ne training they were actually receiving
- 3. Audits
References:
QI-QAP-2.1-4, Auditors Certification DQI-CS-4.6, R6, Conduct of Internal, Prime and Subcontractor Audits
- a. General The TUGC0 QA audit program is based on FSAR Section 17.1.2 which addresses ANSI N45.2.12, Draft 3, Rev. O. TUGC0 Corporate Office is responsible for audits both internal and external. The audits spanned contractors, engineering, construction and corporate.
Audits are listed in five areas, Site Construction / Engineering / Quality Control, Operations /Startup, Vendor, Pre-award Surveys, and Vendor Surveillance. Audits scheduled in the five areas were 107, 158, and 80 during 1982, 1983, and 1984, respectively.
- b. Review Effort A review was made of the licensee's implemented audit program to verify whether it meets the requirements of the accepted QA Program and ANSI N45.2.12 (Draft 3, Revision 0 - 1973) as endorsed by the QA Program. The reviewer also verified the following aspects of the audit program:
The scope of the audit program has been defined and is consistent with FSAR commitments Responsibilities have been assigned in writing for the overall management of the audit program Methods have been defined for taking corrective action when deficiencies are identified during audits The audited organization is required to respond in writing to audit findings
6 Distribution requirements for audit reports and corrective action responses have been defined Checklists are required to be used in the performance of audits . The reviewer selected audits TPC 40, 43, 56, 57, 61, 69, 70, and TUG 22 performed during 1982 and 1983 for review. The audits were preplanned to cover specific functions and were comprehensive. The reviewer noted that some audits had not been distributed in accordance with ANSI N45.2.12-1977; however, proper corrective action had been taken by QA audit supervision and was documented by memorandum dated August 16, 1983. Subsequent reports were distributed in a timely manner. Review of the vendor audit program is discus-sed in paragraph B.7. The records of four lead auditors and two auditors were reviewed. The qualifications of auditors and lead auditors were verified to be in accordance with the requirements of ANSI N45.2.23-1978. Confirmation of annual documented evaluations of qualifications of auditors were verified.
- c. Conclusion ,
References:
(a) CP-QP-18.2, R2, Implementation of the Permanent Plant Records Management System (b) CP-QP-18.3, R2, Permanent Plant Records System Organization
'(c) CP-QP-18.4, R2, Permanent Plant Records Receipt Control and Storage (d) CP-QP-18.5, R2, Automatic Records Management System Implementation (e) CP-QP-18.6, R0, Record Turnover to TUGC0 Operations Group (f) CP-QP-18.7, R0, N-5 and N-3 Code Data Reports (g) CP-QP-18.8, R1, Records Verification
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(h) CP-QAP-11.1, R3, Fabrication and Installation Inspection of Components, Components Supports, and Piping
-(i) CP-QAP-16.1, R20, Control of Nanconforming Items (j) CP-QAP-12.1, R8, Inspection Criteria and Documentation Requirements Prior to System N-5 Certification (k) CP-QAP-18.1, R2, Processing QA Records (1) CP-QAP-18.2, R4, QA Review of ASME III Documentation
- a. General The quality assurance records program is based on FSAR Section IA (B) which addresses ANSI N45.2.9 (Draft 1, Rev. O,1973) for the
-design and construction of Comanche Peak. The site records program is managed under the control of the Site QA Manager. The Permanent Plant Records Vault (PPRV) houses most of the design and construction records for completed work and have had final review performed. Completed records are being turned over to the control of the operations records control system on a regular basis.
Temporary storage of records is also ongoing at several working locations at the site utilizing one-hour fireproof cabinets. Records, where possible, are filed, by system and component. The PPRV uses smoke detectors tied into the site fire station for records fire protection; a water hose adjacent to the main PPRV door provides fire extinguishing capability, as do portable fire extinguishers in the area. A computer is used to aid record retrievability, but is not essential, as records are maintained in hard copy. Records flow to the PPRV through both a regular site construction /QC path and an ASME path,
- b. Review Effort A review was made of various procedures to verify that provisions had been made to maintain various types of quality records, and that responsibilities had been assigned to carry out the records storage requirements. Records storage procedures were also reviewed to ensure that they described the storage facilities, the filing systems used, methods of receipt, and handling and disposal of the records. The Brown and Root (B&R) program for flow of ASME
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Section III records to the PPRV was reviewed. The reviewer also verified retrievability of records from the PPRV. To verify general record retrievability, the reviewer selected several general construction and inspection packages such as weld data, concrete placements, equipment packages, and equipment travelers. All records were retrieved in a short period from the PPRV. During the review, other records were retrieved of specific design / construction / inspection activities. No significant difficulties were identified during these real-time challenges to the records retrievability system. The ability to expeditiously locate and retrieve records is identified as a strength. This ability appears to be primarily due to indexing and storage of records by component or material, when possible, instead of by record type. To review the B&R ASME records flow, the records associated with safety injection isometric SI-2-RB-13-4; Core Spray CS-1-SB-032; Chemical and Volume Control CT-1-SB-14; Component Cooling CC-2-SB-042; Boron Recycle BR-1-SB-05 Spool IQ3; BR-1-SB-004 Spool 103, BR-1-SB-006, and Main Steam MS-1-SB-050 were reviewed. These records contained the inspector's identification, the type of inspections, the acceptability, verification of review and approval, and were readily retrievable. Heat numbers on materials installed in the field were recorded during a site tour. Certified Materials Test Reports (CMTR).were requested and i furnished which verified traceability for those items recorded i during the' tour. Al so CMTRs, for selected subassemblies were { verified to meet ASME code requirements. " *-
- - d # -
~
~
This program of records review, approval and turnover from B&R, the ASME "N" stamp holder, to TUGC0 appears to be very thorough, though complex. Records for l work performance by B&R are assembled, reviewed, and aporoved, l then submitted to the Authorized Nuclear Inspector (ANI) for l review, then submitted to TUGC0 for filing. A task force i comprised of B&R and TUGC0 personnel, then make another review of these records. Any discrepancies noted are then resolved between B&R and TUGCO. These records are then red labeled, and can not be removed from the vault without written approval of QA management-thereby, preventing loss of QA records. l A review was made of the temporary storage of records in the field. Although records are best protected in the PPRV, record storage in adequate fire proof cabinets is allowed based on the record storage equipment qualification in NFPA No. 232-1975, which bases fire protection on exterior fire load calculations. Although the reviewer did not check any fire load calculations J
9 justifying the use of one-hour fire cabinets, those cabinets observed appeared to be adequately protected. During this review, the observation was made that several completed ASME moment restraint record packages being maintained in a non-fireproof cabinet in the ASME Safeguards Building QC trailer. This failure to store quality assurance records in a fireproof cabinet is a potential enforcement issue. Prior to conclusion of the review, these records were relocated to fireproof cabinets. Based on the above problem, the reviewer noted some confusion at the site on the control of " documents" as they progress through design / con-struction/QC and as to when they become " records." This was evident as little distinction appeared to be made for the storage of " documents" or " records" in the field. Working " documents" were provided equal to or better protection than " records" in some instances. Other than the example stated, no other storage problem was identified. Comanche Peak had established, on March 30, 1984, records monitoring teams to review the records flow program. The clarification of the document / records interface for storage control is a weakness and is to be addressed by the monitor teams. This weakness is considered part of the potential enforcement issue addressed above. The physical construction of the PPRV was reviewed. The construc-tion of the PPRV is satisfactory for protection from exterior fire damage. For inside originated fire damage, the PPRV has a fire detection system but does not have the industry standard water or halon automatic fire suppression system. The system for unattended PPRV fire control was reviewed. With the fire detec-tion alarms annunciating in the close-by fire station, the fire station personnel having ready access to the PPRV and the location of a fire hose reel outside the PPRV door, the fire protection appears adequate. Verification was made that the operations vault, into which all the PPRV records will be transferred, contains an automatic fire suppression system.
- c. Conclusion The records control of the PPRV appears to meet all requirements, with sufficient staff to control the activity.
1
- 5. Document Control
References:
(a) DCP-3, R17, CPSES Document Control (b) DET-12, R0, DCC/ Task Force Interface
t 10
- a. General Controlled documents, primarily drawings, specifications, and procedures are maintained and controlled by the site Document Control Center (DCC). The p.edominance of document control within the sphere of the DCC relates to drawing control and changes to those drawings. The DCC has established satellite document control centers which control and distribute most of the working documents. These satellites provide controlled document copies to crafts and the Unit 1 Task Force Paper Flow Groups (PFG).
Controlled documents and changes are provided to the satellites from the DCC. -- -- -
~ ~ -
. , ,e g vices
- e . . . . . . . e a w rung in at specific building task force. Revisions to controlled drawings and documents that affect controlled drawings, such as design change authorizations (DCAs) or component modification cards (CMCs) are distributed upon receipt to the satellites and controlled number recipients. For drawings, a computer system keeps track of drawings and the DCAs and CMCs that affect those drawings. When new drawings, drawing revisions, DCAs, or CMCs are generated the computer is updated.
When the satellites receive a new drawing revision, CMC or DC/.., any controlled drawings checked out to .the crafts or under the control of the PFG are updated by the satellite DCC personnel. This maintains current the controlled drawings in use by insuring that drawing packages contain the correct revision with applicable DCAs and CMCs. Drawings checked out to the craft from the PFGs or directly from the satellites are returned at the end of the working day. Prior to checking out drawings from a satellite directly to the craft, a computer run is made to insure that drawing packages contain the appropriate revision and applicable CMCs and DCAs. When craft personnel return drawing packages to the satellite or PFG, a drawing, CMC and DCA check is again performed to verify return of the controlled documents.
- b. Review Effort A review was made of the references listed to verify they met the requirements of the accepted QA Program. The reviewer also verified that administrative controls have been established for the control of drawings and that indices are maintained for drawings, manuals, specifications, and procedures which indicate current revisions.
In order to verify the control of drawings, the reviewer selected several drawings to determine if the current drawing revision with applicable DCAs and CMCs located in the DCC, was also onhand in the control and auxiliary building PFGs. Two drawing di scre-pancies were noted. Drawings 2323-El-2011, R8 and 2323-El-0900, Sheet 1, R6 maintained in the PFG had several DCAs in the package
s 1 l, f 11 that were missing from the current drawing package computer printout. The verification was performed on April 12, 1984, using a current drawing status. This problem appears to be from engineering eliminating CMCs and DCAs from its data base applic-able to particular drawings without informing DCC of the change. Although the computer change keeps satellite issues current, no
" trigger" device causes satellite personnel to remove tne CMCs and DCAs from the PFG drawing package. A review of the engineering mechanism for updating the data base found the procedure satis-factory and a review of having non-applicable CMCs or DCAs in the drawing package revealed that while possibly confusing, the practice is not a technical problem. As the working contrclled drawing packages are expected to be current at all times, this mechanism whereby non-applicable CMCs and DCAs remain in con-trolled drawing packages is identified as a weakness.
The computer assisted drawing control program was reviewed. Specifically, with the sole reliance on the current computer printout to determine drawing package adequacy, the controls of computer input and changes were reviewed. Access codes have been established so that a limited number of engineering and DCC personnel have access to affect their respective data base. A procedure and training exists to define appropriate computer changes authorized for each grcup. The system appears to be adequately controlled and use of a computer system versus stamped drawings referencing DCAs and CMCs is identified as a strenath. During this review, a frequent observation from all r M personnel revealed that other than the inconvenience of the sheer volume of a large number of CMCs and DCAs in a package, they had not encountered construction errors due to accumulation of DCAs and CMCs. In that no problem appear to be developing, but the potential to lose control is high when drawings are not revised periodically to keep outstanding drawing changes reasonably low, the maintenance of working drawings with a large number of completed CMCs and DCAs without a drawing revision is identified as a The applicant does have a program under way which began $ s ago to update those drawings identified by ation.
c. Conclusion
The limited review revealed that the current document control system appears to be functioning satisfactorily. All DCC and PFG personnel interviewed were aware of their responsibilities and how their job was performed. The DCC, satellites, and PFGs reviewed appeared to be adequately staffed. _ .~
./
12 The use of the drawing control computer appears to keep craft personnel up-to-date in an expeditious manner.
References:
(a) CP-CPM 8.1, RI, Receipt, Storage, and Issuance of Items (b) CI-CPM 8.1, P1, Color Coding of Piping Material s (c) CI-CPM 8.2, R5, Control of Spare Parts (d) MCP-10, R7, Storage and Storage Maintenance of Mechanical and Electrical Equipment (e) ICP-5, R3, Control of Permanent Plant Instrumentation (f) CP-QAP-8.1, R7, Receiving Inspection (for ASME items) (g) CP-QP-8.0, R2, Receiving Inspection
- a. General Warehousing activ'i ties are managed under the Project Support Services organization. Safety-related material is stored in several warehouses and also in an outside laydown yard. All material is received at one warehouse and then moved to the appropriate storage location. Shipping damage inspections are conducted by warehouse personnel and receipt inspections are performed by QC inspectors. Environmentally sensitive material is stored in a temperature and humidity controlled storage location.
A preventive maintenance program exists to insure that mechanical and electrical equipment is maintained in an operable condition while in storage,
- b. Review Effort A review of the licensee's program for the receipt, storage, and handling of equipment and material with respect to selected elements of the licensee's accepted QA Program was performed. The review was to verify that administrative controls had been established concerning receipt inspection of safety-related
\, materials, preparation and retention of required documentation, t control of nonconforming and conditional release items and control of items in storage. Implementation of the program was reviewed et
13 by selecting several safety-related items in storage and verifying document and item control to be in accordance with the program. The reviewer also toured the warehousing locations. Storage discrepancies were not identified. The QC receipt inspection program was also reviewed. QC inspections appeared to be conducted in a satisfactory manner.
- c. Conclusion Based on the limited review of the warehousing and receipt inspection program and implementation, both programs appear adequately managed. Storage locations appear adequately staffed.
Warehousing and QC personnel were knowledgeable and professional in their respective areas.
- 7. Procurement
References:
(a) CP-EP-5.0, R7, Procedure for Field Procurement (b) DQP-CS-2, R6, Procurement (c) DQP-CS-4, R9, Procedure to Establish and Apply A System of Pre-Award Evaluations, Audits, and Surveillances (d) DQI-CS-4.1, R3, Vendor QA Manual Reviews (e) DQI-CS-4.2, R3, Generating and Maintaining the TUGC0 Approved Vendors List (f) DQI-CS-4.3, R4, Vendor Performance Evaluation System (g) DQI-CS-4.4, R4, Conduct of Vendor Pre-Award Evaluations (h) DQI-CS-4.5, R6, Conduct of Vendor Audits (i) DQP-VC-1, R7, Final Inspection and Release for TUGC0 (j) DQP-VC-2, R7, Witnessing Trip (k) DQP-VC-3, R3, Initiating Yellow Flag Sheets (1) DQP-VC-4, R6, Guidelines for Certifying Vendor Compliance Inspection Personnel
e 14 (m) CP-QP-5.0, R1, Quality Assurance Review of Site Generated Procurement Documents
- a. General Safety-related purchase requisitions are generated by TUGC0 engineering at the site and are converted to purchase orders by the site procurement and subcontracts section. Technical and QA requirements are determined by engineering. A QA review of all safety-related purchase orders is conducted on site to verify QA requirements and use of an approved vendor. Each purchase order requires the vendor to inform TUGC0 when a product is ready to ship. TUGC0 QA determines whether to perform a pre-shipment inspection at the vendor's location or to waive this inspection.
Approximately one-third of all safety-related shipments are source inspected. TUGC0 also maintains a vendor audit program to insure that vendors can meet the requirements imposed by the purchase orders. The vendors that are satisfactorily audited are placed on the approved vendors list. TUGC0 has also initiated an annual review of supplier performance.
- b. Review Effort A review was made of the licensee's procurement program with respect to selected elements of the accepted QA Program. The review was to verify that administrative controls had been established for the preparation, review, approval and revision of procurement documents. A review of the licensee's procedures to verify that acceptable methods were being used to qualify vendors which provide quality goods or services; that these procedures required the maintenance of records of supplier qualifications and audits; and that responsibilities have been assigned to perform the vendor qualification program was performed. Several purchases orders at the site and at the TUGC0 offices in Dallas were reviewed. Purchase orders, based on the limited review, appeared to be handled satisfactorily.
Also reviewed was the source inspection or witnessing program implemented from the TUGC0 QA office. The program is quite extensive and appears to be very effective at performing material inspections at the source and identifying potential problems difficult to detect by a receiving inspection alone. A portion of this program, though, needs clarification. Although, the witnessing procedures describes how to perform the source inspection
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p/ i his YV hpd, is considere ure weakness, ut not a program weakness. The entire witnessing program is a strencth. IA/ 0
\ Also reviewed was the vendor audit program, which is used to maintain the approved vendors list. The reviewer selected several vendors on the current list and reviewed their most current audits. All audits reviewed were considered satisfactory. Two of g the vendor audits, Dresser Industries and Forney Engineering were fi v last audited in 1978. The licensee, through the FSAR, utilizes
, /# velop the audit program, a part of which is the vendor audit program.
g / .
= e
. es this annual requirement with respect to vendor audits, in that vendors may be audited triennially providing that annual evaluations continue to show the vendor performing satisfac-
[t)6ggg torily. The TUGC0 vendor audit program does not provide for an annual, triennial or any periodic vendor audit schedule. Vendors are reaudited primarily on a usage and performance history basis. This failure to establish meas to audit vendors at least , triennially is considered a The i inspector found no indication that a o e cally , resulted in maintaining an unsatisfactory vendor an the approved l vend st. ! ute for a ' TUGC0 audit, the large numoer of source inspections would mitigate the possible consequences of not performing periodic vendor audits. j
- c. Conclusion The procurement program appears to be satisfactory. The vendor witnessing program is an asset and appears well managed. Other than the missing timetable for the vendor audit program, the conduct of audits and vendor annual evaluations appears to be well managed. Personnel in the procurement QA staff appear to be 1 knowledgeable and professional in their work.
C. Equipment Turnover and Preoperational Testing
References:
CP-SAP-3, Custody Transfer of Station Components I STA-802, Final Acceptance of Station Systems, Structures, and Equipment CP-SAP-21, Conduct of Testing l 1 l
16
- a. General The processes of turnover of safety related equipment from construction to startup as well as pre-requisite and pre-operational testing of said equipment were reviewed in order to determine if:
(1) The method employed for transferring custody of components, partial subsystems, subsystems or systems from construction to ~ startup; the return of equipment to construction for rework or modification; and the ultimate release of custody from startup to operations are' technically and administratively adequate. (2) The administrative controls over preoperational testing are technically and administratively adequate. (3) The preoperational test procedures both performed and yet to be performed are technically viable and administratively sufficient.
- b. Review Effort (1) Equipment Turnover
- The turnover of safety related equipment from Construction to Startup is administratively controlled by Startuo Administrative Procedure CP-SAP-3, Custody Transfer of Station Components. This procedure establishes the requirements and responsibilities fo'r ~
transferring custody of components,' partial subsystems, subsystems or systems from:
.(a) Construction to Startup (b) Startup back to Construction for rework or modification (c) Startup to' Operations
- The Startup group determines the turnover boundaries necessary to perform pre-operational testing activities. The Completions Group (a subgroup of Startup) assembles the turnover packages consisting of equipment, valve, piping and instrument lists, drawing lists such as flow, instrumentation and control, and auxiliary one-line diagrams as required to sufficiently describ? the content and boundaries of the turnover.
The Completion Group is also responsible for initiati ng and processing turnovers consistent with established schedules in the turnover package, such as to: (a) identify the equipment (b) indicate the scope of the turnover .. 1 l 4
.i 17 (c) assemble the late revisions of the appropriate diagrams /
prints and applicable design change documents (DCA's) (d) list deficiencies, including design changes that have not been implemented The Completion Group coordinates all required pre-turnover walkdowns and punchlist activities for the purpose of establishing the status of remaining work to be done prior to turnover of that equipment to startup. Startup personnel review the packages and perform a walkdown of the equipment / system to determine if the equipment identified in the package is ready for turnover. Any deficiencies requiring resolution prior to turnover are resolved prior to transfer; those deficiencies not requiring pre-turnover resolution are added to the Master Data Base (a computerized tracking system) to facili-tate future disposition. Upon completion of the startup walkdown and correction of required deficiencies, custody / turnover of the equipment is transferred to startup. Custody of station components may be returned to construction for performance of work such as major modifications, repair or clearing of construction deficiencies. The return of equipment to construction voids all preoperation testing on said equipment. After the completion of applicable prerequisite tests, (construc-tion tests), including initial operation of the equipment, startup may relinquish " operational control" to Operations yet maintains custody of the equipment pending completion of preoperational testing. The turnover packages for the following systems were reviewed: (a) Component Cooling (b) Auxiliary Feedwater (c) Containment Spray (d) Chemical and Volume Control (e) Residual Heat Removal (f) Safety Injection (g) Hydrogen Recombiners (h) Reactor Protection System im..
. . ~ .. - - - - .. - -- .- . - .. . . .
18 The turnover of equipment from Startup to Operations is detailed in Station Administrative Procedure STA-802 Final Acceptance of
. Station Systems, Structures and Eouipment. Pursuant to that Procedure, Operations initiates a detailed review of the turnover package and walks down the applicable equipment. Following successful completion of the reviews and walkdowns, Operations
- accepts the equipment / area. At this time all responsibility for
..that equipment lies with operations.
There has been no safety related equipment transferred to operations, thus the. review of the process was in terms of programmatic sufficiency. (2) Preoperational Testing Program 4 The preoperational test program was reviewed in order to verify that the tests to be performed have been identified and that each of the identified tests entailed at a minimum, test objectives, summary of the test, necessary prerequisites, and acceptance criteria. The test organization was reviewed in order to verify that the lines of authority and responsibilities of test personnel are specified and that where interfaces exist between organizations involved in the test program, that organizational responsibilities are clearly established. The administration of the test program was reviewed 'in order to verify that methods are established to receive (from construction) the jurisdiction over systems before commencement of testing.
! The administrative mechanisms established for jurisdiction control of systems before, during, and after testing were reviewed in order to verify that those mechanisms adequately provide for:
control of system status before preoperational testing including 1' the completion of adequate prerequisite (construction) testing; the return of systems- to Construction if necessary to support modifications and/or reports; the control of system status subsequent to testing including measures necessary to prevent invalidation of test results; the control of the system during testing; only the assigned System Test Engineer or his designate may conduct system testing. The conduct of testing was reviewed in order to verify that adequate administrative measures provide for: methods to change a test procedure during the conduct of testing; the criteria for interruption of 'a test and continuation of an interrupted test; methods to coordinate the conduct of testing; methods to document significant events, unusual conditions or interruptions to
,_ testing; methods for identifying deficiencies, documenting their
-m-J
- ~ ,a y, -m e , . - - - - ~ - - -
y - - -- , - -
- . - , - , , ---e e
. 19 resolution and documenting retesting; methods for providing the current test procedure to operations and coordinating test activities with the shift supervisor; methods to ensure that the systems test engineer has the appropriate latest revision of the required documentation / references.
The program for evaluation of test results was reviewed in order to determine that: deficiencies are clearly identified and appropriate corrective action proposed, reviewed and completed; subsequent to corrective actions or modifications have been completed, tests or portions of test have been rerun as necessary to ensure that tests of the as-built system are adequate; the results of the evaluations were reviewed by the appropriate licensee personnel responsible for approving the original proce-dure. (3) Prerequisites Tests Selected prerequisite tests were reviewed in order to determine if the tests provide and adequate mechanism of accomplishing vital testing and operation of the associated equipment. The tests reviewed appeared technically and administratively sufficient. The prerequisite tests when performed in compliance with Startup Administrative Procedure CP-SAP-21, Conduct of Testing, and as required by the applicable preoperatidnal tests, appear to provide an. adequate mechanism for initial equipment checkout and operation. (4) Preoperational Tests
- Selected preoperational test procedures for tests which are yet to be performed, were reviewed in order to ascertain adequate implementation of the following
(a) Management review and approval (b) Procedure format with emphasis on clarity of testing required (c) Clarity of test objectives t (d) Pertinent prerequisites identified, e.g.-
- . 1) required plant systems are specified
- 2) proper facility procedures and other references are
, specified and uniquely identified -
- 3) completion of calibration checks, limit switch setting protective device setting, included where applicable
- 4) special supplies, and test equipment specified.
, . _ . . . . . . - _ . _ _ _ - _ _ , - , .- _.,,,,y.-
20 (e) Special environmental conditions, if any, identi.fied. (f) Acceptance criteria are clearly identified and the procedure requires comparison of results with acceptance criteria. (g) The source of the acceptance criteria is identified, i.e., FSAR, T/S, Reg. Guide, engineering drawing, etc. (h) Initial test conditions are specified
- 1) Valve line-ups
- 2) Electrical power and control requirements
- 3) Temporary installations (instrumentation, electrical, and piping)
- 4) Temperatures, pressures, flows (1) The procedure includes reference to appropriate FSAR sections, T/S, drawings, specification, codes and other requirements.
(j) Step-by-step instructions for the performa.nce of the proce-dure are complete to the extent necessary to assure that test objectives are met. (k) Provisions are available for documenting that all items, including prerequisites, are verified as having been per-formed. (1) Provision is made for recording details of the conduct of the test including observed deficiencies, their resolution, and retest. (m) Procedure requires that temporary connections, disconnections or jumpers be restored to normal or refers to another procedure. (n) Procedure provides for identification of personnel conducting the testing and evaluating the test data or refers to another procedure. (o) Procedure provides for independent verification of critical steps or parameters, including QA holdpoints. These procedures included but were not limited to the following: 1-CP-PT-11-01 Component Cooling 1-CP-PT-29-2 D/G Control & Functional
.e e ,
21 1-CP-PT-48-01 Containment Spray 1-CP-PT-49-02-RT-1 CVCS - Seal Water & Letdown Performance Retest 1-CP-PT-49-03-RT-1 CVCS - Chemical Control Purification and Makeup Retest 1-CP-PT-57-01-RT-1 SI Pump Performance Retest Selected completed preoperational procedures were reviewed in order to ascertain, at a minimum that: (a) The licensee is performing an adequate evaluation of test results. (b) All test data are either within previously established acceptance criteria, or that deviations are properly dispositioned. (c) The licensee's methods for correcting deficiencies and for retesting are adequate. (d) The adequacy of the licensee's administrative practices in maintaining proper test discipline concerning test execution, test alteration,'and test records. (e) The licensee is following his procedures for review, 4 evaluation, and acceptance of test results. These procedures included, but were not limited to: 1-CP-PT-57-06 RHR - ECCS 1-CP-PT-67-01 Hydrogen Recombiner 1-CP-PT-64-02 Reactor Protection System 1-CP-PT-57-02 Centrifugal Charging Pump 1-CP-PT-57-01 SI Pump Performance 1-CP-PT-48-01 Containment Spray 1-CP-PT-29-04 D/G Sequencing 1-CP-PT-02-08 Class I-E Switchgear (5) Systems Status System walkdowns were performed in order to determine the current status of safety related components / systems. The following systems, among others were selectively reviewed in that assess-ment: (a) Residual Heat Removal (b) Chemical Volume and Control (c) Safety Injection (d) Containment Spray (e) Auxiliary Feedwater (f) Component Cooling 4 _ . _ . _ , ,_ y ,. y - , _ . _ . _ ,_ - - . - . . _ _ , , _
22 Preoperational test status reports were also reviewed and inter-views conducted in order to assess the current status of completed and remaining testing. The review revealed that of the 198 original preoperational test procedures, 45 have yet to be performed; of the 34 preoperational/ retest procedures, 33 have yet to be performed; that of the 39 preoperational/reperform proce-dures, 37 have yet to be performed. Thus of 271 total procedures, 115 or 42% have yet to be performed. It should be noted however that the~ " Retests" and "Reperforms" are, as a general rule, much less in scope than the original preoperational test and as such should require less time to complete. Further the " Retests" and "Reperforms" will be run on essentially " debugged" systems, thus should run much smoother than the original tests. (Note: The retests and reperforms were necessitated by extensive electrical rework and station modifications.) There is no preoperational testing currently ongoing, nor has there been any significant testing in the past 10 months, the result of the aforementioned electrical rework and other modifi-cations. Plans are currently underway to recommence preopera-tional testing during the month of April 1984. A statistical analysis of the preoperational testing which has been performed, spanning the period of July 1982 to June 1983, in essence. the period immediately proceeding a virtual shutdown of testing necessitated by the modifications as aforementioned, revealed that in that 11 month period,177 of the 198 original tests were performed. This calculates to be an average of.11 tests completed per month. Applying this rate to completion of the total testing remaining,115 tests, it would take approxi-mately 10 months to complete the preop program. If, however, one assumes that rate would apply only to the original preoperational tests, not the retests or reperforms, and a valid assumption that the retests and/or reperforms can be run in. conjunction with or at ' least during the time frame of the preop tests, then the 45 remaining original preops can be run in 4 months. Assuming preop testing resumes in April 1984 as planned, preop testing could conceivably conclude by August 1984, if no major undisclosed problem is identified. It should b'e noted that a mechan' ism /meth'od now embraced by the utility to facilitate turnovers, is that of room / building turn-overs in conjunction with the equipment inside. This is cumber-some and could impact preoperational testing. Preoperational testing is performed on a system related basis, thus if a system is complete, yet the room in which the system is placed is not (i.e., painting, etc.), preoperational testing may be, and is under the current program, delayed until room turnover. (Note: See Section A for changing completion methodology).
, . , , . .___ym. y - --
, - , _ , , ., - .a,
23
- c. Conclusion Based on the above limited review, the following conclusions were formed:
(1) The administrative process of custody transfer of systems appears to be adequate. (2) The preoperational test program appears to be intact, viable and adequate. (3) Preoperational tests appear to be technically and administratively adequate. (4) Preoperational testing could conclude by August 1984.
References:
QI-QP-11.2-3, Torquing and Spacing of Concrete Anchor Bolts QI-QP-11.3-23, Class 1E Conduit Raceway Inspection QI-QP-11.3-26, Electrical Cable Installation Inspection QI-QP-11.3-27, Class IE Power Cable Meggering QI-QP-11.3-28, Class 1E Cable Terminations QI-QP-11.3-29.1, Verify Electrical Separation QI-QP-11.3-38.1, Installation of Class 1E Electrical Equipment QI-QP-11.3-40, Post Construction Inspection of Electrical Equipment and Raceways QI-QP-11.3-42, Electrical Inspection of Seismic Category 1 Instrumentation Rack Assemblies QI-QP-11.10-1, Inspection of Seismic Electrical Support and Restraint Systems QI-QP-11.3-50, Cable Grip Support Installation Inspection
- a. General The assessment in this area was to determine if safety-related electrical equipment'was being installed and inspected in accordance with NRC requirements and licensee commitments and to determine if 0I Texas Utilities Services Inc., (TUSI) programs which includes drawings, procedures, quality control and construction inspections, and quality records are adequate to accomplish work in this activity.
Discussions were held with craftsmen and other Comanche Peak Steam
\ Electric Station (CPSES) project eersonnel to determine their ability and knowledge to carry out their individual responsibilities and to evaluate their morale and opinion with regard to the Comanche Peak e nuclear project. No adverse comments were made by the Comanche Peak
~
l 24 project empicyees and all considered the project to be of high quality construction.
~
The licensee recently organized his manpower into a Building Management Organization (BMO) to make the most efficient use of aroject resources. There are four main BMOs - Containment Building, Safeguards Building, Auxiliary Building, and Electrical Control Building. Each organization is an integrated group of engineering, construction, and QA personnel. This group supports the effort to complete the construction in their area of assignment under the direction of a Building Management Director. The department supervisors are responsible for the technical direction of their personnel, and QC personnel report to the applicable QA Department manager. There is an exchange of problems and resolution of problems among the project personnel and bi-weekly BM0 meetings. As a room or area is considered nearly complete an Electrical Separa-tion Verification (QI-QP 11.5-29.1) is performed on the room and/or area. ' When both these procedures are complete, or essent- y complete, and/or at the discretion of the BM0 director, the room and/or area becomes controlled. Access is, limited to correct minor outstanding deficien-cies or complete other' known outstanding work. The BM0 Director determines when this room and/or area is to be turned over for an inspection and acceptance by the Stations Startup and Test Group. This turnover usually follows the inspections and completion of most of the deficiencies found during the performance of QI-QP,11.3-29.1 and QI-QP 11.3-40. An inspection walk down was performsd on many of the rooms / areas that the BM0 Director considered to be essantially complete. This walkdown showed that the rooms / areas were clean, that electrical / mechanical separation, including barriers, cable tray attachments, identification of cable trays, conduits, and cables, cable t-ay fill and cable spacing (where applicable) in trays, and cable v toorts (Kellen grips or equivalent) were satisfactory.
- b. Review Effort (1) Review of Quality Assurance Implementing Procedures The referenced procedures were examined to assure that FSAR requirements and commitments were being complied within the areas relating to the installation and inspection of electrical equip-I ment and components.
l These procedures provided check lists and acceptance criteria for QC inspector.
25 (2) Electrical Cable Installation The following installed safety-related (S/R) electrical cables that had been accepted as satisfactory by site constructi were examined. A physical examination was mace to te rmine compliance with applicable design and installation criteria relative to type, location / routing, identification tags at termination points, minimum bend radius (where applicable), cable color compatible with designated raceways and separation of trains, excluding barriers, which are performed prior to or concurrent with QI-QP-11.29-1, " Verify Electrical Separation." The routing was checked by using a signal generating device. Cable No. Tyoe From To EG100483 3/C No. 10AWG MCC1EB2-1 MOV 1HV5540 EG113626 9/C No. 12AWG MCCIEB2-1 CPIECPRTC08 EG113646 9/C No. 12AWG MCCIEB2-1 CP1ECPRTC05 EG112219 2/C No. 12AWG MOV IHV4759 CPIECPRTC05 EG100497 3/C No. 8AWG MCCIEB2-1 MOV IHV4759 EG112216 5/C No. 12AWG MCCIEB2-1 MOV 1HV4759 E0100009 1/C No. 4/0AWG SWGR1EA-1 TBXCSAPCH01 E0112206 5/C No. 12AWG MCCIEB3-1 MOV 1HV4758 E0112207 7/C No. 12AWG MCCIEB3-1 CPIECPRTC04 E0112209 2/C No. 12AWG CPIECPRTC04 MOV IHV4758 ~ The cable identification is accomplished by an alphanumeric coded tag and by the color of the cable jacket. The first character of the alphanumeric code indicates whether the cable is safety or channel oriented (E), associated train (A) or non-safety (N). The second character identifies the color of the cable jacket and with respect to safety-related (S/R) applications they are "0" (Orange), "G" (Green), "W" (White), "B" (Blue), "R" (Red) and "Y" (Yellow). All cables are to be tagged with their unique alphanumeric number at termination points in equipment and junction boxes. Cables that enter and leave a junction box but are not terminated in that junction box are not required to be identified in that box with their alphanumeric number. All of the j above cable were properly identified. The routing of the above cables was checked with signal tracers. Using this method, junction box covers, cable tray covers, fire barriers and other items did not have to be removed. This check showed that cable tray systems and conduits appeared to be properly installed with proper attachments and supports, that l these systems were properly identified, and that the cables
- travelled the route indicated on the cable pull cards.
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26 (a) QI-QP-11.3-26, Electrical Cable Installation Inspection (b) QI-QP-11.3-27, Class IE Power Cable Meggering (c) QI-QP-11.3-28, Class 1E Cable Terminations i (3) Electrical Cable Termination I A physical examination was made on terminations of selected class 1E electrical cables in the Hot Shutdown Panel on elevation 832' of the unit 1 safeguards building. The examination verified that terminations were in compliance with requirements, including proper lug material and size, accurate location, and identifica-tion of terminal block and conductor. The cable wiring diagram was used to determine the proper termination points and conductors identification. Cable Terminations that were checked were for cables EG104556, EG111148, EG104551, EG139204, E0104791, E0104740, E0122101, E0104742, E0130596, and E0122103. The QC records showed that inspections were made on these termina-tions in accordance with QI-QP-11.3-28, " Class IE Cable Termina- l tions." redundant Class IE cables and Class 1E/Non-Class 1E cables within a cabinet shall be maintained in accordance with the equipment specification. If the specification gives no separation require-ments, the minimum separation distance between redundant Class 1E and Class 1E/Non-Class IE cables shall be greater than or equal to 6 inches. In cases where the above separation criteria -cannot be maintained, barrier shall be installed between the cables." Acceptable barriers include the following: (a) Metallic conduit; including Servicair Company FC 33 flexible conduit (b) Two sheets of fire retardant material separated by a minimum of " of air space or thermal insulating material (c) A single barrier with a 1" maintained air space or thermal insulating material between the components or devices and the barrier During the cable termination inspection in the Hot Shutdown Panel, f it was noted that barriers were installed but there still existed p s. - - i.n .r.blems. . _ __ j ,, .y .
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and that the remaining barriers would be insta e'c as needed to e meet the separation criteria before QI-QP-11.3-40 was signed off f for that room or panel.
l 27 To insure that internal electrical separation in panels was being adhered to, several panels in which QI-QP-11.3-40 was essentially complete were examined. These panels were located in the cable spreading room and control room. The panel examined included termination cabinets TC-22, 23, Auxiliary Relay Panels 1, 2, and
- 5. These panels showed that internal separation was satisfactory even though work was still in process in some of these panels.
During the inspection for electrical separation in the above panels it was noted that some cables in the panels were being spliced. This was was determined to be satisfactory and meets FSAR commitments which state in paragraph 8.1.5.2.5. , " Wire splices are used in limited applications on field cables that terminate in certain Class IE panels, cabinets or racks. T.he normal design is to terminate field cables without the use of wire splices. The wire splices are only used where additional length is required for the field wire and it was not judged reasonable to pull a new field cable. The use of such wire splices has been minimized. The wire splices are butt splices. The crimping technique, device and materials used for the splices are identical to those used for the terminal lugs in that panel. The wire splices are only allowed on low power applications such as control cables. Since previously accepted crimping methods and materials are used, the splices are limited to low nower circuits and to field cables that already terminate in the panel, and the required wire separation and wire bundles support is maintained..." Interviews with CPSES project personnel which were conducted by other members of this review team indicated that there may be a problem with cable terminations to Weidmuller Terminal Blocks. These terminal blocks employ a screw clamp connection. The manufacturer's literature for these terminals blocks states, "The screw clamp" refers to a connection in which the wire is stripped of its insulation to a recommended length and clamped without any further preparation. A screw clamp and current bar are used to insure the connection; and since the clamping screw does not make direct contact on the wire, damage is prevented." As inspectors 1 were making inspections for QI-QP-11.3-40, " Post Construction J Inspection of Electrical Equipment and Raceways," they would tug and flex the conductor to insure that the connection was tight. This action caused the conductor wire strands to slightly spread and thereby reducing the tightness of the screw clamp connection. Since these connections were previously verified as satisfactory per QI-QP-11.3-28 " Class 1E Cable Termination" inspections and the fact that equipment may be energized, the licensee now calls for a visual inspection with regard to QI-QP-11.3-40 termination checks. The Weidmueller Terminal Blocks used at CPSES are qualified per the manufacturer's literature for nuclear applications including environmental qualification. Tests for this qualification were
3
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28 performed by Franklin Research Center and are documented on their I reports FC 4959 and 5205. j (4) Electrical Conduit and Cable Tray Installation Conduit and cable tray raceway systems were inspected in rooms - and/or areas in which both QI-QP 11.3-29.1, " Electrical Separation Verification" and QI-QP 11.3-40, " Post Construction Inspection of Electrical Equipment and Raceways," were essentially completed and access to these rooms and/or areas were controlled. This inspection was to verify completeness of work in the electrical area, including electrical separation, power cable spacing in trays and cable supports on vertical runs of cable systems. All items were condidered to meet construction criteria. Specific Conduit System checked including support and spacing were: Conduit No. Location Remarks C13005319 Safeguards Bldg #1, Elev 773, Room 56S Access Controlled C1304036 Safeguards Bldg #1 Access Controlled Elev 773, Room 56S 1 C13012998 Safeguards Bldg #1 Access Controlled Elev 773, Room 565 C13010777 Safeguards Bldg #1 Access Controlled Elev 773, Room 565 C14013679 Safeguards Bldg. "I Access Controlled Elev 773, Room 54 C22G08188 Aux. Bldg. Eley. 790 Only Room 170 was Various Rnoms Access Controlled C22G08189 Aux. Bldg. Elev. 790 Only Room 170 was Various Rooms Access Controlled The inspection of these conduits showed that they were installed to the construction requirement and that electrical separation was satisfactory. QC records for these conduit systems showed that applicable inspection were made in accordance with the following procedures: QI-QP- 11.3-23, Class IE Conduit Raceway Inspection QI-QP- 11.2-3, Torquing and Spacing of Concrete Anchor Bolts
29 QI-QP 11.10-1, Inspection of Seismic Electrical Support and Restraint Systems QI-QP 11.3-29.1, Verify Electrical Separation [For Room 56S, Inspection Report (I.R)# E-1-0013485/3-84; for Room 54, IR# E-1-0013480/3-84;- for Room 170, IR# E-1-0017514/1-84] Several additional conduit runs were examined in the field to verify electrical separation. These conduit runs were located in the cable spreading area and are identified below: (a) Conduit C12019632, orange safety train, goes under ladder
. tray T16GCCM02, green safety train, at one point and separa-tion is approximately 6 inches. At another point it goes over ladder tray T14GCDH41, green safety train, and separa- i tion is approximately 2" with a barrier installed between the tWo.
(b) Conduit C15R10537, red protection channel, at one point goes under ladder tray T13GCCM15, green safety train, and separa- ! tion is approximately 2 inches. (c) Conduit C15B11396, blue protection channel, at one point goes under ladder tray T130CCM0, orange safety train, and separa-tion in approximately 2 inches. , (d) Conduit'C12G21191, green safety train, goes under solid tray T140CDJ31, orange safety train, and separation is approxi-mately 3 inches. The above are acceptable per QI-QP 11.3-29.1 " Verify Electrical Separation" and Gibbs and Hill Specification 2323-ES-100 Section 4.11.3.2. , Spacing of power cables in trays is to follow requirements of Gibbs and Hill Specification 2323-ES-100 section 4.2.1.4., which in essence, states that minimum spacing between power cables shall be a minimum of one quarter of the diameter of the largest cable. The spacing of cables in the 'following trays and rooms were considered to meet this requirement: Electrical Separation Tray Numbers Location Verification per QI-0P-11.3-29 T120ABA05-12 Room 174, Aux. Bldg. Not complete
~T120ABB01 Room 174, Aux. Bldg. Not complete T110AA01-05 Room 174, Aux. Bldg. Not complete T110SAA30 Room 54, Safeguards Bldg. Complete
30 T120ABA96* Room 219, Aux. Bldg. Approx. 90% complete T11GAAB11* Room 214, Aux. Bldg. Complete T120ABA98* Room 241, Aux. Sidg. Approx. 90% complete T120ABA47-50 Room 241, Aux. Bldg. Approx. 90% complete T120ABB93 Room 219, Aux. Bldg. Approx. 90% complete
- Asterisked trays contained vertical runs of cable. Cables were supported properly by Kellem Grips in accordance with QI-QP-11.3-50, " Cable Grip Support Installation Inspection."
A review of some of licensee Inspection Reports (irs) that were performed for QI-QP-11.3-29.1 " Verify Electrical Separation" showed that I .R E-1-0024985 of 2/28/84 and IRE-1-0036072 of 4/12/84 applied to the same room (room 219) in the auxiliary building. Neither of these reports indicated that they were performed as a result of a specific job or Inspection Item Removal Notice (IRN). Both were designated as final inspections. It is recognized that the licensee can perform re-inspection as deemed necessary; however, it is considered that there should be only one final inspection for post construction work. If additional final inspections are required in this area for IRN's, Design Change Authorizations (DCA), etc., they should be referenced in the remarks section of the IR. The one " final" electrical separation inspection, which .could be performed concurrent or before QI-QP-11.3-40 " Class 1E Electrical Post Construction Verifica-tion," would indicate that electrical work in this area in almost complete and would aid in triggering the performance of QI-QP-11.3-40. The licensee stated that this area would be reviewed to see if the " final" inspection in this area could be clarified. t
- c. Observation and Conclusions
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References:
QI-QAP-11.1-28, Rev. 23, Fabrication, Installation Inspections of ASME Component Supports, Class 1, 2, and 3 i ' I
7 31 QI-QAP-11.1-28A, Rev. 5, Installation Inspections of ASME Class 1, 2, and 3 Snubbers Procedure AB-5, Rev. 5, A Simplified Method for Design and Analysis of Small Size Piping TUSI Engineering Guidelne, Section IV, Base Plates, Rev. 11 CPSES,~XCP-ME-10, Rev. 1, Pipe Support Adjustments TUSI CP-EI-4.5-1, Rev. 9, General Program for As-Built Piping Verification TUSI Engineering Guideline, Section II, General Engineering Sectopm II Criteria for Pipe Support Design, Rev. 8 Specification 2323-MS-46A, Nuclear Safety Class Rev. 5 Pipe Hangers and Supports ,
-Construction Procedure Field Surveys 35-1195-CCP-9, Rev. 4, TUSI CP-EI-4.6-9,~Rev. 1, Performance Instruction for Piping Analysis by SSAG TUSI Engineering Guideline, Hilti Concrete Anchor Section V, Rev. 3 Bolts ADLPIPE, Static and Dynamic Pipe Design and Stress Analysis, Arthur D. Little, Inc., May 1981
- a. General The organization of the general site engineering, construction, and procurement efforu were defined in procedure CP-EP-3.0. By this procedure, the Project Manager is responsible for the Comanche Peak Steam Electric Station (CPSES) design and engineering. These activi-ties are normally delegated to Gibbs and Hill, Westinghouse, and other organizations. However, the licensee, (TUGCO) retains overall respon-sibility for design activities and performs design functions as
. necessary. The TUGC0 Engineering Manager is responsible for the general direction of engineering activities. FSAR Chapter 3 provided the licensee's requirements for the design of structures, components, equipment, and systems. The reviewer selected samples'in pipe support design, piping stress analysis, and design
32 procedure applications to verify program implementation, to ensure site procedures, site interface procedures, and design interface procedures satisfy NRC requirements and licensee commitments. 4
- b. Review Effort The reviewer held discussions with the design engineering personnel in the pipe support group to determine whether they understood the applicable design control procedures; whether they were 'able to verify 4
design parameters that were within the applicable criteria and/or design specifications; and whether the person doing the design review was independent from the individual who performed the design. The reviewer also held discussions with the engineering personnel in the piping system Site Stress Analysis Group (SSAG) to determine whether they performed their work activities in accordance with established instructions, procedures, and specifications. The seismic response spectra with respect to operating basis earthquake (OBE) and safe shutdown earthquake (SSE) were discussed with the responsible engineers. It was noted that these seismic response spectra were furnished by the A/E's (Gibbs and Hill, Inc.) home office to the site stress group to be used for the piping system analysis. The followhg major areas were reviewed to determine a conclusion: (1) IE Bulletin 79-02, Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts, Requirements (a) Factor of Safety for Concrete Expansion Anchor. Bolts Design A review of the Pipe Support Engineering Guidelines Manual, Section V, revealed that a factor of safety of five (a more conservative value) has been used for establishing the allowable loads (tension and shear) for the wedge bolt calculation. In accordance with the vendor (Hilti) design manual and the NRC IE Bulletin 79-02 requirements, the factor of safety of four could be used (Comanche Peak pipe support installations use Hilti wedge bolt only). As noted above, the safety factor used exceeded the requirement. (b) Pipe Support Base Plate Design IE Bulletin 79-02 states that pipe support base plate flexibility be accounted for in the calculation of anchor bolt loads. Discussions with the responsible engineers indicated that the pipe support group personnel do consider base plate flexibility into their design calculations. Finite element method (base plate flexibility consideration) has been used for non-typical (other than four anchor bolts in one plate) base plate analysis. FUB II base plate program has been utilized for all typical (four anchor bolts in one plate) base plate analysis. The FUB II program generally
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- 33. I produces loads which are about 25% higher than the loads i generated by the Finite Element Method. In fact, many base 1' plates were analyzed by the more conservative program, FUB II computer application (developed by ITT Grinnell Corp.). This approach exceeds the NRC requirements.
(c) Anchor Bolt Tension - Shear Interaction IE Bulletin 79-02 permits a formula to be used for calcula-tion of bolt-tension-shear interaction. This formula can be interpreted from a linear distribution to an elliptical distribution. Comanche Peak pipe support group has elected to use a linear distribution (a conservative approach) for all concrete expansion anchor bolt cafEFationsh (2) IE Bulletin 79-14, Seismic Analysis for As-Built Sa'fety-Related Piping Systems, Requirements \v' This bulletin states .that the seismic analysis input information conforms to the actual configuration of safety-related piping systems. Licensees are requested to verify: pipe run geometry; i support and restraint design, locations, function and clearance; embedments; pipe attachments; and valve and valve operator locations and weights. To accomplish the above requirements, the site pipe support group and the site stress analysis groups are responsible for verification based on as t built configuration. The
.' as-built configuration is identified by'a'TTeld survey team. This
, consists of three surveyors and one OA field surveyisteam, inspector, which, field measurements by utilizing equipment to perform such as transits, levels, theodolities, etc. The high accuracy of the information obtained through the field survey is a highlight for implementing the IE Bulleting 79-14 requirer.ents.
(3) Alternate Analysis for Small Bore Piping Systems The reviewer examined portions of procedure AB-5, A Simplified Method for Design and Analysis of Small Size Piping, Rev. 5, May 1982. It was noted.that the procedure was developed by Gibbs and Hill, Inc., in a very conservative manner in terms of thermal load and seismic load calculations. Furthermore, approximately 30% of small bore (2 inches and under) low energy pipe lines in Unit 1
, and 10% in Unit 2 are analyzed by the Alternate Analysis Method (i.e., a simplified method for design and analysis of small size
- ^
piping). The balance of small bore piping is analyzed by the computer application. (4) Rigorous Analysis for Safety-Related Piping Systems Most of the safety-related piping systems are analyzed by the rigorous analysis method. The computer program involved in the analysis is one of the typical programs being used in the _, _ _ ._. _ . . ~ _ _ _ _ _ _ _ - _ _ . _ _ _ . _ _ __
f 34 industry. This computer program, ADLPIPE Static and Dynamic Pipe Design and Stress Analysis, has been developed and uodated by Arthur D. Little, Inc. , since the early 1960s. (5) Iterative Design Process The reviewer held discussions with responsible licensee represent-atives in the area of safety-related pipe supports and piping systems. It was noted that the Iterative Design Process was utilized for implementing the design of pipe supports and the analysis of piping systems. In accordance with the licensee's description: "the process for the design of piping and supports is iterative in nature. It is unrealistic to expect to design piping and supports to satisfy all applicable requirements the first time through the process. Such an iterative design approach is employed throughout the nuclear industry, and is utilized in (6) Review of Design Calculations for Pipe Support Support No. Pipe Size pipino System AF-1-002-705-S33K, Rev. 3 10" dia. Auxiliary Feedwater
~
CC-1-158-701-A43R, Rev. 2 16" dia. Ccmponent Cooling SI-1-031-709-A32R, Rev. 2 12" dia. Safety Injection SI-1-029-702-532R, Rev. 2 24" dia. Safety Injection BR-1-AB-001-005-3, Rev. I 2" dia. Boron Recycle The above design calculations were randomly selected and were partially reviewed for conformance to analysis criteria, applic-able codes, NRC requirements, and the licensee commitments. Furthermore, these calculations were evaluated during the review for thoroughness, clarity, consistency, and accuracy. Deflection criteria used for support design were discussed with the respon-sible engineers and were partially verified. Weld size calcula-tion and snubber size determination were also verified for adequacy. In general, the design calculations appeared to be adequate in terms of using design input, reference, units (dimension, force, and moment), equations, tables, and sketches.
I 35 (7) Review of Stress Analysis for Piping Systems Calculation Nc. Piping System AB-1-19A Safety Injection AB-1-30 Containment Spray AB-1-69 Residual Heat Removal and Safety Injection AB-1-135E Auxiliary Steam and Main Steam AF-1-SB-006 Auxiliary Feedwater AF-1-SB-007 Auxiliary Feedwater The above piping stress analyses were partially reviewed for conformance to design specification, applicable code, NRC require-muits, and the licensee commitments. Thest analyses were also evaluated for thoroughness, clarity, consistency, and accuracy. The NRC reviewer examined portions of the seismic inputs to be used in the stress analysis. These seismic inputs in terms of pe-iods versus accelerations from the corresponding floor response spectra curves under OBE and SSE conditions were partially verified for accuracy. Furthermo're, the reviewer held discussions with' the responsible engineers to ensure that seismic anchor movement, nozzle ther, mal movement, and valve orientations were properly considered in the stress analysis. During the review the reviewer examined piping system AF-1-SB-006. This 3/4" diameter vent and drain pipe was analyzed for support requirements. Results from the analysis revealed that no pipe supports were needed for the pipe. However, the reviewer noted that a Component Modification Card (CMC) No. 90567 was issued to the pipe in that a piece of tee (pipe) was added to the vent and i drain system. The pipe support group accepted this CMC without 1 performing detailed evaluation. The responsible engineer stated that this CMC was reviewed by a well qualified engineer. Based on f # fv his engineering judgement, no detailed calculations were required. [ l The inspector indicated that a detailed evaluation for this CMC g0 } j [i was needed. In addition, a sampling program should be initiated to ensure that no other similar CMCs were accepted without I performing detailed evaluation. The responsible licensee repre-sentative took immediate action to perform detailed calculations for the vent and drain piping system due to the addition of the CMC (No. 90567). Furthermore, a sampling program was immediately initiated to review 50 other similar packages. This matter will be identified to the Comanche Peak Project Director for followup.
M Yll 36 Results from the detailed calculations revealed that no pipe supports were required for the vent and drain piping system as the original evaluation indicated. Results from the sampling program showed that no discrepancies were identified for the 50 other similar packages. Piping system AF-1-SB-007 was partially reviewed.
- reviewer. The pipe support group reanalyzed this 3/4 inch piping system by hand calculations (alternate analysis) and also by computer application (rigorous analysis). Results from the two analyses were consistent and conservative. Four pipe supports were required by the analysis. Loads used for support design were verified and were found conservative. This matter will be forwarded to the Comanche Peak Project Director for followup.
(8) Field Inspection / Verification The NRC reviewer performed a field walkdown at the l'n i t I containment building area and noted the following discrepancies: Support No. Status CC-1-218-012-C53K Snubber connection cotter keys missing CC-1-295-005-C53R Sway strut installed over 5 tolerance CT-1-038-436-C62K Snubber connection cotter keys missing; no washers in rear bracket CT-1-117-405-C62K Snubber connection cotter key missing CT-1-117-415-C62K Snubber safety wire broken CT-1-053-444-C62K The south snubber was installed improperly DD-1-046-020-C65R Snubber cotter keys missing FW-1-096-705-C62K Snubber safety wire broken FW-1-102-002-C62k Snubber cotter key missing; needs relative adjustment on snubber FW-1-102-003-C62K Snubber cotter keys not bent MS-1-151-025-C52K Snubber installed over 5 tolerance CC-1-RB-056-008-3 Snubber cold setting over the limit
&WO W 37 CC-1-RB-066-007-3 Snubber cold setting over the limit CC-1-RB-068-007-3 Spring hanger cold setting incorrect (15 lbs. versus 11 lbs.)
The above pipe supports discrepancies were verified with the licensee's OC inspector in accordance with detailed drawings. All The licensee representa-f f tives stated that a final walkdown inspece'on/ verification for all [/Mp pipe supports is to be implemented in accordance with procedure V CP-QAP-12.1, Inspection Criteria and Documentation Requirements Prior to System N-5 Certification. The majority of the discrepancies appeared to be minor problems which could be easily repaired during the final inspection prior MS-1-151-025-C52K, Rev. 3 and CC-1-295-005-C53R, Rev. 4, w -h were not installed in accordance with the detailed drawings. (9) Design Consideration for Piping Systems Between Safety-Related and Non Safety-Related Buildings The NRC reviewer held discussions with the licensee represen.ta-tives in the area of piping stress analysis and pipe support design. Stress Analysis No. AB-1-135 E for the Auxiliary Steam and Main Steam System was partially reviewed and discussed with respect to design considerations between safety-related and non safety-related buildings. The piping system was classified as high energy line and safety-related. The pipe run starts from the Turbine Building into the Electrical Control building. Since seismic classifications for the two buildings are different, the criteria used for the piping system analysis should also be different. The failure of the pipe in the Turbine Building may impose a damage to the pipe inside the Electrical Control Building if the piping system was not properly analyzed and designed. The responsible licensee representatives agreed to performed further evaluation with regard to the above concerns. This matter will be identified to the Comanche Peak Project Director for resolution. (10) Interpretation of Tolerance for Snubber Installation [ During the field review, three reviewers interviewed the licensee's QC inspectors with respect to their interpretation of f.i.g_ degrees tolerance requirements for strut and snubber
38
' The reviewers held discussions with the licensee representatives with regard to the above concerns. It was determined that the rh- licensee will revise the inspection procedure to clari fy the strut / snubber installation tolerance and will conduct a training
[d
, for all QC inspectors who are involved pipe support inspections.
This matter will be identified to the Comanche Peak Project Director for followup. (11) Final Adjustments for Spring Hangers and Snubber Settings The reviewers held discussions with the responsible licensee representatives with regard to implementing the final adjustments for spring hangers and snubber settings. It was determined that, after the fuel loading, the licensee QA startup group will perform the final walkdown inspection to ensure that all spring hangers and snubbers be adjusted to proper position. This matter will be brought to the attention of Comanche Peak Project Director for followup. (12) Technical Training The reviewer held discussions with the responsible pipe support engineering (PSE) personnel to determine whether they performed their work activities in accordance with established procedures A review of the training record revealed that since 1980, the PSE personnel have received extensive training activities in terms of technical applications and code interpretations. Portions of the training courses are listed as follows: Date Course Attendance (Engineers) (a) 06/16/80 Introduction to Nuclear All Codes and Standards, QA for Engineers (b) 10/13/80 ASME Code Seminar All 10/14/80 (NF Design Philosphy) (c) 04/13/81 Alternate Analysis Method 26 for Small Size Piping (d) 06/21/81 Vent and Drain Piping 8 Seismic Qualification (e) 05/11/82 Design Verification 34 05/13/82 Process
39 (f) 07/14/82 Pipe Support Snubber 24 07/15/82 Installation (Instructed by Manufacturer) (g) 07/ ~/82 Analysis of ASME Class 16 2 and 3 piping (h) 11/12/82 Seismic Analysis of 65 11/16/82 Pipe Supports 11/17/82 (i) 06/14/83 Finite Element Method 19 thru (including ASME 1, 2 & 3 08/06/83 pipin analysis) (j) 06/29/83 Current Version of ADLPIPE 9 Computer Code (Stress Analysis) (k) 11/17/83 Quality - It's Your Job All (1) 03/08/84 Snubber Reduction Program 6 (m) 03/19/84 Stability Problem in the 26 Design of Pipe Supports The above training activities in the area of pipe suppor't designs appeared to be effective and well administered. This observation was supported by the extensive discussions with the responsible engineering personnel and by reviewing the procedures and results of the design calculations.
c. Conclusion
Discussions with the responsible personnel revealed that the enginee-ring personnel involved in the area of stress analysis for piping systems and pipe supports appeared to be knowledgeable. A review of portions of the alternate analysis criteria and related documents was performed. It was noted that the methods and procedures used in the criteria were conservative. A review of the eleven calculation packages indicated that computer applications were extensively used in the stress analyses, pipe support designs and, base plate and concrete expansion anchor bolt calculations. Design calculations, in general, were good. During the review, the NRC reviewer noted that conservative considera-tions were found in many areas of design and analysis. These conserva-tive considerations included: factor of safety used for concrete expansion anchor bolt calculation, computer program (FUB II) used for base plate analysis, weld stress allowables for welding connections,
/W 40 alternate analysis for small bore piping, and seismic loads used in design and analysis. These consecutive design considerations are considered strengths in the applicants program. Finally, the reviewer noted that the geographic location of Comanche Peak site has the lowest seismic risk in the United States in accordance with the criteria specified in Uniform Building Code.
r@ M s sequen ~Towup 1 to ensure t at safety-related pipe supports are installed in accordance with design drawings and to verify that cerrective actions with respect to the aforementioned discrepancies are ad.equately implemented in accordance with established procedures.
- 14. Installation of Safety-Related Fluid Systems
References:
(a) QA-QAP-11.1-26, Rev. 14, "ASME Pipe Fabrication and Installation Inspec-tions" (b) QI-QAP-11.1-28, Rev. 23 " Fabrication, Installation Inscection of ASME Component Supports, Class 1, 2, and 3" (c) QI-QAP-11.1-28A, Rev. 5, " Installation Inspections of ASME Class 1, 2, and 3 Snubbers" (d) CP-QAP-12.3, Rev. 3, " Testing Phase Quality Assurance Functions Prior to ASME Code Certification and Stamping" (e) CP-QAP-12.2, Rev. 7, " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing"
- a. General The review of this area was directed to assessing the adequacy of the licensee's construction program as it pertained to installation of safety-related fluid systems required for safe operation and shutdown of the plant. The assessment was undertaken through selective examina-tion of installed systems and installation related activities to determine whether they were accomplished in accordance with good engineering practice and with licensee commitments and NRC require-ments - including the requirements of the applicable code, ASME
~
41 Section III. The review in this area did not undertake to evaluate the
-licensee's final : hecks and analysis of system piping in accorda344,
% a-
- b. Review Effort (1) Tour of Areas Containing Safety-Related Fluid System Components The reviewers toured the Safeguards, Auxiliary, and Reactor Buildings and the Service Water Pumping Station to observe installed safety-related fluid system comp;nents for any visually apparent signs of unsatisfactory or questionable items such as visual weld defects, undersize welds, improperly or insufficiently supported piping, damage to more susceptible support components (e.g., snubbers), corrosion, missing or loose fasteners and spacers, etc. Only one item of concern, requiring follow-up, was identified during the tour. A spring can piping support was found to have a significant buildup of rust inside the can on the spring. The licensee was informed of this spring can, which was identified Serial No. 942-12. The rusting in this item did not appear to be so severe as to significantly impair its function but the course of the rusting and its significance to the functioning should be evaluated further by the licensee.
(2) Control of Welding Materials
, The reviewers examinco the licensee's control of the welding materials used in installation of safety-related piping system components at the issuance stations to verify compliance with code requirements and good practice. Specific attention was directed to the adequacy of the licensee's:
segregation, identification, and control of filler metals', including consumable inserts oven storage of low hydrogen electrodes to limit moisture pick-up preparation of issuance records handling of returned filler metals documentation of current welder qualification limitations The reviewers also observed areas toured in the plant, as des- l cribed in (2) above, and plant areas entered for specific item ! inspections for evidence of inadequately controlled filler I materials. No evidence of uncontrolled or improperly controlled l welding materials was observed. The licensee welding material 1 1 P
4 42 controls observed by the reviewers met or exceeded code require-ments and good practice. (3) Piping and Supports The reviewers visually examined examples of installed runs of safety-related piping and associated supports to verify they were in accordance with good engineering practice and that they were in compliance with code requirements and with licensee drawing and procedure -requirements. Three runs were selected which had most or all of their final acceptance inspections completed. Two of these were nearly ready for the final code review required for ASME certification '(referred to as N-5 certification) that the installations were in accordance with the code. The third had the certification complete. The licensee contracted the piping and support installation work to Brown and Root, Inc. This contractor was responsible for assuring compliance with code requirements, including obtaining code inspector certification therefor (on N-5 Data Reports). Licensee procedures applicable to and utilized by the reviewers in the examination of piping and supports were examined for compli-ance with code requirements. The procedures were as follows: , (a) QA-QAP-11.1-26, Rev. 14,' "ASME Pipe Fabrication and Installation Inspections"
*(b) QI-QAP-11.1-28, Rev. 23 " Fabrication, Installation Inspection of ASME Component Supports, Class 1, 2, and 3" (c) QI-QAP-11.1-28A, Rev. 5,
" Installation Inspections of' ASME Class 1, 2, and 3 Snubbers" (d) CP-QAP-12.3, Rev. 3, " Testing Phase Quality As:ur- ;
ance Functions Prior to ASME
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Code Certification and Stamping" (e) CP-QAP-12.2, Rev. 7, " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" i i
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43 The. runs of. piping and supports ' installed that were examined by the reviewers were described on isometric drawings. The runs examined, ~ identified by the drawing numbers, and the examination - checks made by the reviewers are as follows: Run: 3" Containment Spray (ASME Section III, Class 3), Drawing BRP-CT-1-SB-019, Rev. 6 The reviewers visually selectively examined the installed safety related piping to -verify the following in accordance with the drawing, code, procedures; and good engineering practice: configuration apparent pipe size valve identification visual appearance of welds heat numbers on pieces 2,10, and 18 and serial number on valve . piece 14' were traceable through installation- 1 records to original receipt and acceptance records The reviewers examined the records for the above piping to verify the following in accordance with code and procedural requirements: proper installation and inspection steps completed for all components mill test, reports for all materials hydrostatic testing Run: 2" Reactor Coolant (ASME Section III, Class 1), Orawings BRP-RC-1-RB-10, Rev. 8 and BRHL-RC-1-RB-10, Rev. 2. The reviewers visually examined the ir. '.alled piping and supports to verify the following, in accordance with the - drawings, code, procedures and good engineering practice: configuration apparent pipe size snubber and spring can sizes offset for snubber RC-1-015-707-C41K spring can settings visual appearance of welds size of piping welds support serial numbers 19050, 17791 and 17789 traceable to installation and receiving records heat numbers on material pieces 1 and 12 that were traceable to acceptable mill test reports serial numbers on valves 1RC-8057A and -8058A that were traceable to installation and acceptable receiving 1 inspection records ' g-w- --%, m.-- - - y,--~.-w,. -,,..yo,-,- w,-ym_.,,,,,,gw--,..mm--,,-*ww----.v--e
.\ s 44 visual appearance of fasteners snubber pins and washers evidence of damage to or deterioration of any components The reviewer examined the records for the above piping and supports to verify the following, in accordance with code and 4- procedural requirements:
proper installation and inspection steps completed for piping hydrostatic testing Run: 8" Auxiliary Feedwater (ASME Section III, Class 3), 10" Drawings BRP-AF-1-SB-006, Rev.17 and BRHL-AF-1-SB-006, Rev. 3 The reviewers visually examined the installed piping and supports to verify the following, in accordance with the drawings, code, procedures and good engineering practice: configuration apparent pipe size snubber sizes and settings - visual appearance and size of welds serial number on valve IAF-031 traceable to acceptable receiving records .
,sn'ubber pins and washers evidence of damage to or date,rioration of any components Note: Heat number traceability could not be checked on the materials and weld quality could not be checked entirely satisfactorily as most of the components were painted.
The reviewers examined the records for the above piping and supports to verify the following in accordance with code and procedural requirements: proper installation and inspection steps completed for piping hydrostatic testing The licensee's procedures and installation appeared to generally meet or exceed the applicable requirements and were in accordance with good engineering practice. Records proved readily retrievable and complete. Licensee QC inspectors who accompanied the NRC reviewers in their examinations of the installations appeared knowledgeable. One item of concern was noted - it was not clear what tolerance was applied to snubbers and sway struts that were installed with offsets or
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e 45 angles specified by drawing. This concern is discussed in paragraph E.b(10). (4) Residual Heat Removal Heat Exchangers (RHR Hxs Supports) The reviewers requested the licensee to identify and provide for review the bolting requirements, the drawings and the installation records for the RHR Hxs. The . drawings and some of the installa-tion records were provided. The bolting requirements were not identified and the welding records - were not provided by the completion of the inspector's visit. The records and information had been requested about 1 to 2 days before the end of the visit and licensee personnel indicated insufficient time was allowed to provide all of what was requested. The r examined the RHR Hx supports for visual weld quality nd installation of bolting. The weld quality appeare (in accordance with code iI [] requirements). se, with many
$ h[ySM' yhreads exposed ween tne nuts an The status of the final inspections to be performed on the Hxs was tc su , gainst which k
- As already indicated above,
- 1To is appears to be radictory to t ngs o the genera ding of the team.
- c. Conclusions p Based on their examination and findings described above, the reviewers generally concluded that the licensee's program for installation of y/[t< jW' safety-related fluid system components assures compliance with require-ments, commitments and good engineering practice. As their assessment
{, l was incomplete relative installation of the Hxs described above, the ( reviewers recommend additional evaluation to complete the review relative to such components. This will be identified to the Comanche Peak Project Director for followup.
,g 1 46 G. Civil Construction Activities
- a. General The objective of this portion of the review was to determine the adequacy of the implementation of the licensee's quality control /
quality assurance program for civil construction activities. During the review selected quality assurance records were examined to verify the records were complete and retrievable. Emphasis was also placed on examination of the document control system. The reviewer examined site civil design activities, including the design change process, proce-dures and QA records for completed work activities such as the SSI dam, selected cable tray supports, and whip and moment restraints; and procedures and work activities for ongoing work including application of protective coatings and testing of Richmond inserts. The reviewer also interviewed QC inspection personnel.
- b. Review Effort (1) Safe-Shutdown Impoundment Dam, Units 1 and 2 (a) Review of Construction and Quality Control Procedures The reviewer examined specifications, drawings, and quality l control procedures for construction of the safe shutdown i impoundment (SSI) dam. Acceptance criteria utilized by the reviewer appear in FSAR Section 2.5.4.5 and NRC requirements.
Construction of the SSI dam was completed in Spring of 1977. The dam was designed by Freese and Nichols, consulting engineers, and was constructed by Brown and Root. The onsite
- quality control inspection activities were performed by i
Freese and Nichols and the firm of Mason-Johnston and Associates. Quality. assurance was provided by Brown and Root site quality assurance group and the Texas Utilities Services, Inc., (TUSI) site QA surveillance group. Documents examined were as follows: Freese and Nichols drawing numbers FN-SSI-3 through FN-SSI-7, Safe Shutdown Impoundment Dam Freese and Nichols specification FNSSI-1, Contract Specification for Safe Shutdown Impoundment Dam Brown and Root Construction Procedure numbers 35-1195-CCP-2 through CCP-8 l - Brown and Root Quality Control Procedure CP-QCP-7.1, Surveillance of SSI Dam Activities I.
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i .g >~ . l 47 l The Mason-Johnston and Associates Corporate QA Manual and Mason-Johnston field and laboratory testing proce-dures ( (b) Review of Quality Records The reviewer examined selected records which document quality control inspection and quality assurance activities during construction of the SSI dam. Acceptance criteria utilized by ! the reviewer are the procedures listed above. Records l examined were as follows: Records of QA workshops conducted by Freese and Nichols and Mason-Johnson and Associates. These workshops were conducted to provide training for field inspection personnel. l Weekly field corrective action reports for April - July l 1976 and January - March, 1977. L Results of quality control tests performed on filter materials, and impervious core materials placed between ! April and July 1976. These records includ'ed results of Atterberg Limits, field density' tests, and proctor tests performed on the imprevious core materials, and results of field density, relative density and mechanical analysis tests performed on the Type A and B filter materials. l l Stop work orders 1 Brown and Root QA Audit Reports Training records of QC inspection personnel Design Change / Design Deviation request numbers FN-81, i FN-82 and FN-84 Based on review of the records, the reviewer concluded that the dam was constructed in accordance with the requirements I ( of the construction drawings and specifications and as !- stipulated in the FSAR. The records were neat, legible, complete, and retrievable. f (2) Unit 1 Reactor Building Internal Pipe Whip Restraints (a) Review of Quality Control and Construction Procedures The reviewer examined specifications, drawings, and quality control procedures for construction and inspection of the pipe whip restraints in the reactor building. Acceptance
y S 48 criteria utilized by the reviewer appear in :>ection s.8 of the FSAR. The pipe whip restraints are non-ASME since they are not attached to the piping. The restraints are treated as part of the reactor building internal structure and are constructed in accordance with the American Institute of Steel Construction (AISC) Standard Practices, as is all other non-ASME structural steel members (cable tray supports, structural steel building frames, stairwells, non-ASME equipment supports) in the power block. This is standard industry practice. The whip restraints were fabricated by the Chicago Bridge and Iron (CB&I) Company. Onsite installa-tion was performed by Brown and Root. Documents examined by the reviewer were as follows: Gibbs and Hill Specification 2323-55-168, Structural Steel (Category I) Gibbs and Hill Drawing numbers 2323-51-0581, 0581-01, 05.84, and 0585, Reactor Building Internal Structure, Pipe Whip Restraints TUGC0 Instruction Number QI-0P-11.14-1, Inspection of Site Fabrication and Installation of Structural and Miscellaneous Steel The re' viewer also examined the outstanding (unincorporated) design changes against the above specification and drawings. There were 29 DCAs against the specification, 12 against drawing number 0581, 3 against drawing number 0581-01, 11 against drawing number 0584, and 11 against drawing 0585. The reviewer examined the document packages maintained in DCC Satellite 306 for the above specification and drawings and verified that they were complete and contained the latest (current) revisions of the drawing and design changes. (b) Field Inspection of Whip Restraints The reviewer, accompanied by a QC inspector, examined pipe whip restraint numbers M-22 and M-25 which are located in steam generator compartment numbers 4 and 1, respectively, on elevation 900 of the reactor building. Acceptance criteria utilized by the reviewer are those documents listed above. Examination of these and other restraints on the 900 eleva-tion, and discussions with the QC inspector and design engineers, disclosed the following problem. OCA number 14,813, Rev. 2, against drawing n mber 2323-S1-0581 revises the erection notes for the whi Discussions with various design engineers and the
C' 49 inspector disclosed that there was some confusion as to where the use of Jam nuts was required. In addition, the reviewer observed several locations where jam nuts had not been installed on anchor bolts where nuts had only been installed hand tight. This item will be turned over to the Comanche Peak Project Director for followup. (c) Review of Quality Records The reviewer examined quality records documenting construc-tion (site erection) and QC inspection of whip restraint numbers M22, M25, and M-37 on elevation 900 of the Unit I reactor building. These records included weld travelers, QC inspection of structural steel bolting, QC inspection of welding, and as-built drawings showing as-built dimensions, elevation and location for the restraints. The reviewer noted that inspections for installation of Jam nuts required per DCA 14813, R2, was not documented in the inspection packages. There was no resolution of this item during the review, therefore, this item will be refered to the Comanche Peak Project Director for followup and resolution. The reviewer did not examine the CB&I whip restraint fabrication records. (3) Review of Nonconformance (NCR) 10453 The reviewer examined NCR 10453 which was written to document and disposition a problem which developed during field erection of four moment limiting component supports on the feedwater lines in the Unit' 1 Safeguards Building. The supports, which are ASME components, are similar to pipe whip restraints. The purpose of the supports, which were erected around the feedwater lines, is to limit movement of the pipes during pipe break accidents. The restraints are constructed from heavy beams and columns which were fabricated offsite by CB&I. During field erection of the restraints (which was accomplished by Brown and Root) cracks developed in welds which attached small (6 inch by 9 inch) gussett plates to the columns and beams when the bolts in the beam-column connections were torqued. The reviewer examined the NCR and discussed the corrective action with QC inspection personnel. Review of the NCR disclosed that it had been revised five times. Some of these revisions resulted from changes to the corrective action af ter further evaluation of the problem. Other revisions were as a result of changes to the administrative handling of the NCR, e.g., to repair all four restraints under one NCR is lieu of writing a separate NCR for each restraint. These types of revisions are normal during disposition of NCRs. Review of the NCR and discussions with responsible inspectors disclosed that the problem was resolved by removal of the damaged gusset plates (i.e., the plates where welds
50 had cracked) from the beam and columns, non-destructive examina-tion (NDE) of the base metal in the beams and columns at the points where the gusset plates had been attached, fabrication of , new gusset plates, and rewelding of the new gusset plates to the beam columns. The reviewer examined selected quality records associated with repairs of one of the restraints, including weld travelers, PT inspection report number 19059 and 19054 and design documents including CMC 96060 and Brown and Root drawing number MSB-0683-CBI. The corrective action to resolve this NCR was completed in March 1984. (4) Unit 1 Cable Tray Supports (a) Review of Quality Control and Construction Procedures 1 The reviewer examined specifications, drawings, and quality control procedures for construction and inspection of cable tray supports. Documents examined by the reviewer were as follows: G&H Drawing Numb.er 2323-El-0713-01-S, Cable Tray' Support Plan, EL 792'-0"& 790-6", Aux & Elect. Control Bldgs. G&H Drawing numbers 2323-5-0901, 0902, and 0903, Cable Tray Support Details, Sheets 1-3 , G&H Specification number 2323-55-160, Structural Steel (Category I) Brown and Root drawing number FSE-00185, Sheets 1-3, Reference Drawing for Cable Tray Hangers Brown and Root drawing number FSE-00159, sheet numbers 527, 537, 557, 2895, 2898, 2904, 2905, 2908, 12580, 12600, 12608. These are the fabrication drawings for the cable tray hanger supports. The sheet number corresponds with the hanger number. _ l
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l The reviewer also examined the outstanding (enincorporated) design changes against the above G&H drawing. Triere were 344 CMCs and 19 DCAs against drawing 0713-01-S, 6 CMCs and i 9 DCAs against drawing 0901, 4 CMCs and 10 DCAs against ! drawing 0902, and 26 CMCs and 29 DCAs against drawing 0903. l The reviewer examined the document packages maintained in DCC i Satellite 306 for the above drawings and verified that they were complete and contained the latest (current) revisions of 1 the design changes. During examination of the design changes the reviewer noted that the majority of them were originated as a result of minor construction problems. For example, most of the design changes to drawing 0713-01-S, which is the
n-51 1 cable tray support location p '. an , were as a result of interferences encountered dur'.ng construction and were requested by construction personnel. These interferences required relocation of soine of the supports shown on this drawing. Often the relocated supports were only moved a few inches. (b) Field Inspection of Cable Tray Supports The reviewer, accompanied by a QC inspector, examined randomly selected cable tray supports located on eleva-tions 790'-6" and 792"-0" of the electrical control building. The supports and the acceptance criteria utilized by the reviewer appear in the table below. TABLE Support Applicable Number
- Support Type Design Change 527 B-2 (Dwg 0901) CMC 8250 537 0-1 W/ Brace -
(Dwg 0901) 557 A-1 (Owg 0901) CMC 94628 DCA 1946 OCA 2687 2895 SP-2 (Dwg 0903) CMC 50474 7198 SP-2 (Dwg 0903) CMC 4521 CMC 2646 2904 SP-2 (Dwg 0903) CMC 52473, R2 DCA 3494 2905 SP-2 (Dwg 0903) DCA 6299-R7 CMC 2646 2908 B-2 (Dwg 09C3) - 12580 B (Dwg 0601-015) CMC 61731 12600 A (Dwg 0500-04-5) CMC 67033 12608 SP-7 (Dwg 0903) CMC 68393 CMC 1969 DCA 19973
- Support number and location shown on B&R drawing number FSE-00185
o 52 During .the field inspection, the reviewer verified the followin'g. were in accordance with requirements specified or ; design ' drawings: method of attachment to wall and/or ' ceiling, dimensions, elevation of support, proper size of structural steel members, joint connection details, and
~
configuration'of support. The reviewer also walked down other areas in the auxiliary and electrical control building and examined cable tray supports for general configuration .and quality of workman-ship. During examination of supports in the Unit I cable spreading room, the reviewer noted. that six and eight inch siderails had been added to four inch deep trays. The practice of increasing the height of siderails on cable trays' and its effect on the design of cable tray supports ' was examined by the reviewer. Details of this revi6w are discussed in paragraph G.b.(7).c below. (c) Review of Quality Records . The reviewer examined quality records documenting constr ;c-tion and QC inspection of the. cable tray supports listed .in the paragraph above. These, records included construction travelers,. weld filler material. logs, and cable tray l inspection reports for installation .cf cable tri.y hangers, - [ cable tray clamps, and instal.lation of expansion. anchors or i Richmond Inserts. Based on review nf the records, and the l walkdown inspection discussed above,- the reviewer' concluded f that the cable tray supports were constructed and inspected in accordance with the requirements of the construction drawings. The records were neat, legible, complete, and
; retrievable.
(5) Inspection and Testing of Richmond Inserts (a) Review of Program for Verification of Installaticn of Richmond Insert Bolts Duri'g n review of records, the licensee determined that documentation of QC inspections were incomplete for installa-tions of Richmond Insert bolts. In order to verify that bolts' of the proper length were installed in the Richmond Insert sleeves, the licensee carried out a reinspection program for the Richmond Insert bolts. The reviewer examined TUGC0 procedure number QI-QP-11.14-8, Verification of
. Installt. tion of Richmond Insert Bolts, which was used to control the reinspection program. During the reinspection e
e _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ A-
.t 53 program,. QC inspectors verified the length of the bolts either through ultrasonic testing or physical measurement, and checked bolt diameter, minimum embedment length, and
" snug tight" condition of the bolts. The reviewer discussed the reinspection program with mechanical QC inspectors responsible for its implementation in the electrical control building. Based on review of the procedures and discussions with the QC personnel, the reviewer concluded that the reinspection program to verify installation of the Richmond Insert bolts was comprehensive.
(b) Observation of Testing of Richmond Inserts The licensee .is performing extensive onsite testing of the Richmond Inserts to confirm the strength values used in design of structures using this type of anchorage. The reviewer examined TUGC0 Engineering Instruction number CP-EI-13.0-13 which specifies the method of installation of test specimens, and describes the test apparatus and specifies the technique used in application of the test loads. The reviewer examined the testing apparatus and verified that the test equipment had current calibration stickers. The reviewer observed the tension test of specimen 28, a 1 inch EC-2W Richmond Insert, and the shear-tension test of specimen 6, a 1 inch EC-6W Richmond Insert. During the tests, the reviewer verified that application of the test load was accomplished in accordance with the procedure requirements and .that the test data was accurately recorded. Following completion of the above tests, the reviewer examined the results of tension and shear-tension tests that had been previously completed and noted that those results were consistent with the results of the tests witnessed by the reviewer. The majority of the modes of failure resulted in failure of the high strength bolts, not the concrete or insert sleeve. The reviewer also examined the concrete cylinder unconfined compressive test data to verify the strength of the concrete was recorded for use in evaluation of the test results. (6) Program for Application of Protective Coatings in the Unit 1 Containment Building (a) Review of Specification and Quality Control Inspection Procedures The reviewer examined specifications and quality control procedures for application and inspection of Service Level I protective coatings, for steel structures, including the polar crane and liner plate, inside the Unit I reactor building. Acceptance criteria utilized by the reviewer
54 appear in ANSI Standard N101.2-1972 and FSAR Section 3.8.1.6.5.g. Procedures examined were as follows: G&H Specification 2323-AS-31, Protective Coating, TUGC0 procedure wmber CP-QP-11.4, Inspection of Protective Coating; TUGC0 Procedure number QI-QP-11.4-1, 11.4-5, 11.4-17, 11.4-22, 11.4-26, and 11.4-28. These procedures cover inspection of storage and handling of protective coating materials, surface preparation, application of the primer and finish coats, and when necessary, coating repairs. TUGC0 Procedure Number QI-QP-11.4-23 and 11.4-29. These procedures cover reinspection and testing of coated steel for which inspection documentation was incomplete. (b) Observation of Protective Coatings Work Activities The reviewer witnessed application and inspection of protective coatings on steel structure inside the Unit I reactor building. During this onsite review the bulk of tha protective coating application work in progress consisted of repairs to the primer and finish coats, and surfac.e prepara-tion for application of coatings. The reviewer verified environmental conditions were being monitored and were acceptable in the reactor building at time of application of coating. The reviewer observed that application of the coatings and QC inspection of the coatings were being performed in accordance with NRC and procedure requirements. (7) Onsite Civil Design Activities (a) General Onsite civil design activities are performed by Gibbs and Hill (G&H) civil-structural engineers who work under the direction of the G&H lead civil-structural engineer who reports to the TUGC0 Nuclear Engineering Manager. The onsite G&H engineers have access to the FSAR, codes, standards and design criteria, and copies of the original design calcula-tions. The bulk of the design work presently being performed onsite relate to review and approval of design changes (CMCs and DCAs). Many of the design changes are originated at the request of construction personnel and involve minor changes, usually due to construction interferences. 1 i l 1 l 1
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55 (b) Review of the Design Change Program The reviewer examined G&H Project Guide-29, Site Review of CMCs, DCAs and S-0910s. This procedure establishes the guidelines under which onsite design change reviews are performed. Acceptance criteria examined by the reviewer were ANSI N45.2.11 and NRC requirements (Criteria III to Appendix B, 10 CFR 50). The reviewer discussed the design change program with license engineers. These discussions disclosed that when a request for design change is made by construction craft or QC personnel, the design change is prepared by civil project engineer. During preparation for the design change request, the civil project engineer usually performs some preliminary calculations in order to arrive at a feasible and workable solution to the problem. After the design change request is prepared, it is transmitted to the G&H onsite design t engineers and to construction. Construction personnel implement the design change "at risk." That is, if the G&H design engineers do not approve the design change, a removal notice is issued and the work affected by the design charge is either removed or reworked in order to comply with the approved design change request. D.iscussions with licensee engineers disclosed that approximately 99 percent of the design changes are approved by the G&H design engineer without revisions and therefore, do not require rework after they are implemented by construction. After receiving the design change request, G&H civil engineers perform a detailed review. Approval of the design changes consists of a detailed review by an engineer, followed by an independent review by another engineer serving as a checker. If the
, design change does not meet the requirements of the design criteria, it is revised as necessary. After it is reviewed and approved, the design change is dictributed per procedural requirements.
The reviewer examined randomly selected design changes which had been made to drawing number 2323-El-0713-01-S, Cable Tray Support Plan. These included two which were currently being reviewed by the G&H design engineers, (CMC 8229, R12 and CMC 8235, R3), several which had recently been reviewed and approved by the G&H design engineers, and several others which had been reviewed by G&H engineer since 1979, the last date drawing 0713-01-S had been revised. Based on this limited review of the design change control program implemented at the site, the reviewer concluded that design changes are being properly reviewed and that design changes are being accomplished in accordance with NRC requirements.
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56 (c) Review of Cable Tray Loading Y As discussed in paragraph G.b.(4) above, the reviewer noted during field walkdown inspections that siderails had been raised on some cable trays in order to accommodate additional electrical cables. The reviewer also noted that fire barrier materials, commonly known as thermolag, were being added to 4 the cable trays (electrical raceways). The reviewer examined the design controls used to verify the structural adequacy of the cable trays from the increase in loadings due to the addition of thermolag and/or addition of cables to the trays. i
! Details of the review are discussed below.
Evaluation of Effect of Thermolag Fire Barriers on Structural Adequacy of Cable Trays / Supports The reviewer examined TUGC0 engineering procedure CP-EI-4.0-49, Evaluation of Thermolag (TSI) Fire Barrier Material on Class 1E Electrical Raceways. This procedure i outlines the program to be implemented to verify that cable trays and supports meet sehmic design criteria after installation of the thermolag is completed. The program will 4 verify that the combination of the weight- of the cables in the trays, the dead weight of the trays, and the weight of the thermolag will not exceed the maximunr design allowable load of 35 psf. The procedure outlines steps to be followed a when the allowable design load is exceeded. The reviewer - discussed this program with licensee engineers who stated that the "as-building" of the cable trays to account for the installation of the thermolag will begin in the near future. After the as-building program is completed, the evaluation of the effect of additional weight of the thermolag on the cable trays will be performed per procedure CP-EI-4.0-49 require- 1 ments. This area is being referred to the Comanche Peak Project Director for followup. I - Evaluation of Increases to Height of Cable Tray Side Rails During the field walkdown discussed above, the reviewer randomly selected for review three four-inch cable trays in the Unit I cable spreading rcom which had 6 or 8 inch side rails. These were tray numbers T-13-0CC-Q07, T-13-GCC-M10, and T-13-GCC-M33. The above trays are 30 inches wide. The reviewer examined sheets 1 and 12 of drawing number 2323-El-0712, and the 133 DCAs against sheet I and 4 DCAs against sheet 12. These drawings , detail the layout and size / type of the above cable ' trays. The reviewer also examined the document packages 4
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57 maintained in DCC Satellite 307 for the above drawings and verified that they were complete and contained the latest (current) revisions of the design changes. From review of the design change documents, the reviewer verified that addition of the 6 or 8 inch side rails to the 4 inch deep trays was authorized by DCAs. For example, the addition of 8 inch side rails to cable tray 13-0CC-Q07 was authorized by DCA 15207. The reviewer discussed the effect that raising the side rails of cable trays has on the tray and support design load of 35 psf with project civil and electrical engineers. These discussions disclosed that the side rail depths were increased because cable extended above the side rails of the 4" deep trays. This often occurs at intersections (TEES) of trays and is a result of cable pulling problems. The engineers stated that whenever the height of siderails is increased, the total loading of the trays is checked to verify it is below the design allowable of 35 psf. The cable load for each tray is documented in the G&H Cable Raceway Schedule, 2323-E-1-1700. Various other schedules maintain the identity of' each cable in each tray and the weight of each cable. The raceway schedule expresses capacity of the trays as percent filled. Review of the schedules -
, disclosed the data shown in the Table below:
TABLE Tray Numoar Number of Cables Percent Filled T-13-0CC-Q07 198 28 T-13-GCC-M10 288 31 T-13-GCC-M33 217 28 From review of the cable schedule, the reviewer deter-mined that the average weight of the cables in tray T-13-0CC-Q07 was approximately 0.11 pounds per linear foot. Therefore the cable load in this tray is (number of cables)(w +/ cable} = (198)(.11) pound /ft = 8.8 PSF Width of tray = 2.5 Ft This is well below the design allowable load value. Based on review of the above schedules and discussions with responsible engineers, the reviewer concluded that the design values used to determine the structural adequacy of cable tray supports are conservative.
58 The reviewer conducted informal interviews with nine civil and six mechanical QC inspectors. Subjects covered during the interviews were the inspector training program, ability to discuss their safety concerns with their management and/or the NRC, cooperation between craft and QC personnel, and availability of technical assistance from engineering personnel. From the interviews, the reviewer concluded that the QC inspectors felt freedom to express their safety concerns to management and/or the NRC, that the inspectors felt that craft personnel were aware of the require-ments to do the work properly, and that the craft recognized the importance of QC inspection activities and cooperated with the inspectors. The inspectors stated that engineering assistance in resolution of problems was available whenever they requested it. The interviews alto disclosed that the licensee has an extensive training program which the inspectors are required to complete prior to becoming certified and being able to inspect and accept work. The training program involves classroom training, on the job training, and passing written and practical exams (the exams contain essay type questions, not multiple choice). The training program for the inspectors performing inspection of structural steel protective coating involved 40 hours of classroom training rm d
.. a ifficult at times, but most saio wor ing in an area for a period of time they became familiar'with the changes and were able to overcome this problem.
- c. Conclusions (1) The licensee has effectively implemented the QA program require-ments in the areas examined by the reviewer.
(2) QC inspectors are knowledgeable of their inspection requirements and perform their inspection in accordance with the licensee's QC procedures. (3) The licensee's QC inspector training program is comprehensive. (4) The licensee's present document control system is good. Though the number of unincorporated design changes against some drawings is large, the availability of a package containing a complete set of the documents made review of the documents possible without too much diffculty to an experienced inspector. The licensee's new unique DCC system (use of computers) exceeds NRC requirements in the area. 1
59 (5) The quality records examined by the reviewer were neat, legible, retrievable, and complete. (6) '- - -
* +.- - -
of unin -- . --.. -.. __ . -
-sults in a . * - . -
- 07
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. es a' - -- " **-14MM e *reve' - -
- s.
The ra a - - - . , . .
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m
. . . . w... -o in paragraph G.b.(2) above.
(7) The design change process is controlled and complies with NRC H. Review of Heating Ventilation and Air Conditioning Systems (HVAC)
References:
Drawings, standards, and specifications applicable to this equipment are as follows: i Hanger Dwg. SG-790-2J-IR, Rev. O Hanger Dwg. SG-790-2J-1V, Rev. O Hanger Dwg. SG-790-2J-R13, Rev. O Hanger Dwg. SG-790-1J-RIL, Rev. 1 Hanger Dwg. SG-790-1J-10C, Rev. 0 . Hanger Dwg. SG-790-1H-RIG, Rev. 0 . l
; Hanger Dwg. RB-832-1E-1A, Rev. O Hanger Dwg. RB-832-1E-1L, Rev. O ;
, Dwg. 2323-M2-0651-HAN, Rev. 2 : ! Dwg. 2323-M2-0651-HBSC, Rev. 1 l Dwg. 2323-M1-0651-HAN, Rev. 6 Dwg. 2323-M1-0651-BSC, Rev. 6 Dwg. 2323-M1-0551-BSC, Rev. 10
)
l l Dwg. 2323-M1-0551-HAN, Rev. 9 i Dwg. 2323-M1-0554-BSC, Rev. 12 Dwg. 2323-M1-0554-HAN, Rev. 7 Dwg. FCUS-0010-HAN, Rev. 5 l Dwg. 2323-S1-0600, Rev. 17 Dwg. MC-134-680C Dwg. MC-143-689C Dwg. DCA 3262, Rev. 1 Dwg. ANS D1.1 Specification 2323-MS-85, Rev. 3 Procedure WP-TUSI-001, Rev. O Procedure DFP-TUSI-003, Rev. 8 l. l l
60
- a. General The reviewer conducted tours of containment, auxiliary building, safeguards building, and control building for both units to generally observe quality, work in progress, material control, and protection of HVAC equipment, as well as weld rod control. Discussions were held with craft and inspection personnel during these tours relative to plant qua.lity.
- b. Review Effort Previous discrepancies identified by NRC regarding. HVAC installation served as a driving force for this review effort. A review was made of evaluations and calculations performed as a result of the previously identified problems. In addition, the reviewer observed HVAC ducting and supports for conformance to applicable drawings, specifications, and standards.
The reviewer generally observed ducting in various areas of the containments, auxilicry building, safeguards buildings, and control building for both units for proper bolting, proper gaskets, and structural integrity. In addition, the inspector observed duct and equipment supports for conformance to requirements. Supports reviewed included unit 2 duct hangers 2J-1R, 2J-IV, and 2J-RIB; Unit 1 duct hangdrs 1J-RIL, IJ-10C, IE-1A,1E-1L, and 1H-RIG; floor mount of Unit 1 Train A Containment Spray Pump Room fan coil unit; and the two unit 1 Safety Injection Pump Room Fan Coil unit hangers.
- c. Conclusion No significant problems were identified relative to ducting. Only minor problems, well within previous discrepancies evaluated, were analysed during the review indicating that these hangers were accept-able. Several minor drawing errors were also noted which were corrected during the review. The evaluations and corrective actions performed as a result of previously identified problems with HVAC installation appear to be adequate.
I. Formal Interviews of QA/QC Personnel
- a. Formal interviews were conducted of QA/QC personnel in order to assist in assessing site quality and management support of site quality. It was felt that discussions with inspection personnel would give a good conservative insight into whether or not the plant was being const-ructed properly. Interviews of five management personnel M Inspectors were selected at random NNepTToh. "'iTeTfricaT~ inspectors were primarily selected from a group of inspectors which had recently been involved in a personnel
61 incident involving a dress code "(Tee Shirt)" issue in order to assess whether these persons had significant technical concerns. In addition, two electrical inspectors indicated a desire to talk to NRC and were interviewed. Several additional electrical inspectors were chosen in addition to inspectors in various other disciplines. The group included inspectors working for eight different supervisors. Experience of these personnel ranged from persons who had been in QC less than a year, to persons who had been at Comanche Peak from early construction (mid 1970s). Most had some previous experience such as site craft, non-nuclear industry or military experience. Some had worked at other nuclear facilities. The major thrust of the interviews was to determine if the personnel had any plant safety or quality concerns. Concerns in these areas were solicited from all those interviewed. Discussions of other subjects were also held with most of the individuals interviewed. These subjects included Wi + 4 support for identifying problems, ability to have prorsems evaluated and corrected as necessary, feedback on evaluation of problems, adequacy of training program, and relation-ship with NRC. All but two inspectors stated they felt the plant would be safe which meant they had no significant quality problems which they felt would compromise safe operation. One inspector, who was not sure of the plant's safety, stated he was assigned to an area which was less controlled than he was used to, e.g., non-ASME code work versus ASME code work (which has the most stringent requirements), and was uncomfortable with the leeway allowed in this area. This person also incicated he had doubts about QA at nuclear plants in general. The other individual who was unsura of plant safety indicate he was satisfied with quality with one exception. This involved a specific problem which he was not sure was adequately evaluated. This item was described to the NRC:RIV Senior Resident Inspector for followup. Two inspectors who stated they had decided en their own that they wanted to talk to NRC, expressed very strongly that the plan'. quality was
" excellent" and there was no plant safety concern. Another inspector, with over twenty-years' experience, who was at his fifth nuclear plant said Comanche Peak was the "best" plant he had seen.
Seven inspectors expressed one or more specific concerns. These concerns involved questions on whether a particular procedure require-ment or whether a particular technical evaluation was appropriate, documentation problems not involving quality of construction, questions whether certain personnel transfers were discriminatory, h in ons, and concerns whic a re een brought up and were yet to be evaluated by the licensee. All concerns have been forwarded to the Comanche Peak Project Director for followup for review and evaluation as necessary. Several concerns were given to NRC:RIV personnel during this inspection and followup showed th'at there was no technical problem identified. m
62 The NRC Resident Inspector was familiar with one of the concerns and had already evaluated the condition as technically acceptable. Several additional concerns were given to RIV personnel verbally on the last day of this inspection for timely followup. The special team interviewer reviewed the concern regarding transfers of six of seven individuals mentioned in the personnel transfer concerns. These transfers appeared to be non-discriminatory. It should be noted that in all cases of concerns involving specific hardware discrepancies these discrepancies had been identified to appropriate licensee personnel and had been or were being evaluated. All in:pectors questioned (21) as to their ability to identify problems such as via NCF.s, indicated no suppression in this area. Several tors indicated that J1CR written evaluat W ~""~a -l a w F4 addbmLmar.a.z: ding problems, such as via explanations of NCR evalua-tions, was considered good by 19 of the individuals questioned. One individual indicated he did not always receive complete feedback but these items did not involve significant technical concerns. Two individuals stated they felt uncomfortable with s - "" - -4 " NCR evaluations. One stated that more feedback was ne u as to re ons for procedure changes. Many of the inspectors indicated that communications were improving and the assignment of the new site QA manager vas a positive step in improving communications. It was clear that some communications problems had existed in the past and rapport between inspectors and their management had been strained previously in some areas. Communi-cations in the ASME code construction area appeared to be exceptionally positive. All but a few inspectors were questioned regardi
- by craft No significant problems were identified a w v 3 . incivi-Generally, the rapport between craft and inspection appeared to be very good.
Adequacy of the training program was discussed with approximately half of the inspectors. Several indicated that d
.e., tougher (not necessarily more ex . u ormal traini g, plus on-the-job training was adequate to perform the inspection functions. Many stated that the training was excellent.
Twenty inspectors felt no hindrance at all to talking with NRC and indicated that the freedom to talk with NRC has been continually stressed by management. Several indicated some apprehension about talking with NRC which appeared to be a natural fear of the position
63 NRC holds. Several were under the impression for a short while that they must have their "act together" if they were going to see the NRC, but now appear to feel no hindrance. Most indicated they saw NRC inspectors regularly in the field but a majority indicated that they had not talked directly with NRC in the field. Interviews of management indicated they were very supportive of inspectors and sensitive to inspector concerns. There appeared to be a strong encouragement for personnel to come forward with any concerns, as evidenced by a memorandum dated March 22, 1984, to all QA/QC personnel from the Site QA Manager. Postings indicating management support for inspectors and other personnel in identifying problems were prominently displayed along with NRC Form 3, NRC Information Notice 84-07 and 10 CFR 21 information. In summarj, although some concerns were expressed requiring further review, these concerns did not appear to be excessive in number or serious and would be normally expected during the interview process. Generally, the most experienced inspectors had a high confidence in the quality of the plant. Past problems in communication and some past apprehens. ion about management support had existed but there :: ems to have been a marked improvement in this area. No one indicated that past communication problems had caused them to not perform inspections properly or not to identify problems when found. Inspector freedom to identify problems and freedom to talk with NRC has apparently been strongly stre,ssed. Management appeared to be sensitive to employee concerns and appeared to be seriously evaluating existing concerns.
- b. In addition to forpal interviews, numerous informal discussions were held between the NRC team personnel and site managers, craft, inspec-tors, engineers, and office personnel as irdicated previously in other sections of this report. The comments received from these individuals were consistent with those received during the formal interviews.
These discuss 1ons covered topics such as plant quality, training, 2 management support, and document control. Appendix A, which follows, is a sanitized listing of concerns raised by individuals during the interview process. The concerns are only those which will require followup by the Comanche Peak Project Director. The interviews were sanitized only so far as confidentiality is related.
64 APPENDIX A Inspector Name: A-1 Date Interviewed: General
Background:
Interviewee Comments: Uncomfortable with less structured program for non-ASME versus ASME; e.g. , seem to change dwg. when structure doesn't meet original, can add welds in field and he doesn't think it gets incorporated into dwg., QC lead can approve changes to travelers for non-ASME structures, not much QA involvement in this area. Specific: Procedure QIQP 1114-12, electrical mounting backfit, craft complained so procedure was revised to reduce number of inspections, 4 revisions made to delete requirements (bolt tight-ening,etc.) H.as the impression that QA has been generally deficient at nuclear plants and QC has not been supported at Comanche Peak in the past. I'ndicated main problem is probably him being able to adjust to non-ASME work: is not aware of code violations taking place.
.s o.
65 Inspector Name: A-2 Date Interviewed: General'
Background:
Interviewee Comments: l Has some concern with use-as-is NCR situations, use-as-is seems particularly prevalent when using Specification ES-100. Specific Technical Concern: NCR was writter when cable damage occurred during Biso Seal removal using a threaded rod. This occurred in Auxiliary Building, elev. 832' . NCR said no damage was done to cable but some insulation had been scraped off by rod. Feel further evaluation may be in order for these cables and there may be similar problems elsewhere. Specific: Wrote 2 NCR's regarding traceability of fuse blocks. Blocks were not marked "Q". NCR said OK as-is because no non-Q blocks were purchased via order MS-605. Feels other similar non-Q blocks have been purchased via different purchase order and could have been installed as Q. Thinks this a possible paperwork problem. . Specific: W' rote recent NCR (not yet evaluated) on GE Motor Control Centers. Compression lugs have bends as much as 180 degrees (more than normally done done by site construction). Don't think GE can violate requirements and may be a prcblem elsewhere in GE MCC's. Also have some broken wire strands which we are fixing as we find.
' Specific: Had previous paperwork conflict problem in solving rework of terminal blocks. 6 page RFIC involved and Proc. SAP-6 involved. Wrote 2 NCR's. NRC inspectors Creek and Johnson were aware, Creek told NRC inspector Taylor, Taylor told to have an answer. Never got fee,dback as to results.
Specific: Repaired a solenoid, shortly after coming to Comanche Peak in craft, without paperwork. Don't know if it was safety related. Not concerned with solenoid technically - did a good job. Notes: The specific concerns were given verbally to the SRI - Construction on 4/12/84 for further. followup. It was indicated during the interview he would get more specifics for SRI. MCC problem was still being evaluated. I suggest allowing the licensee to evaluate and then followup for adequacy of corrective action.
, , y- _
66 Inspector Name: A-3 Date Interviewed: General
Background:
Interviewee Comments: Generally concerned with finding numerous problems during past construction inspection and procedure being changed to delete inspection, e.g., loose terminations found in lighting. Some NCR's are answered simply that the problem is not addressed in Specification ES-100. Recent NCR written because restraint cable (lighting) crimp gages were worn & therefore, inspection was inadequate. This is still being evaluated. Wires of two different gages were terminated at some lugs and many terminations are loose. Have more pressure not to write NCR's during turnover. Found loose LB's (elbow termination fittings) @ East & South ends of Unit 1 Diesel Generators, wrote two NCR's, was accepted.as is. Found cables not trained (routed) in workmanlike manner in Unit 1 Cable Spread Room 0 junction boxes 1058 and 1059. NCR said OK because cable radius was OK but did not admit workmanship problem. Feels post construction inspectors were transferred to Unit 2 as retaliation for finding problems. Heard second hand that IR's (inspection reports) were being written falsely (without reinspection) to clear IIRN's (discrep-ancy report) on cable trays. Heard from lady in Paper Flow Group (PFG) and lady in vault. Said he would get back to NRC with more specifics. Notes: Some review of the lighting termination issue and post check procedure was conducted by team member Ruff. The site inspector indicated he had told of most of these issues and QA was evaluating. I forwarded concern relative to 1058 & 1059 junction boxes to RIV: Martin and he indicated he inspected these boxes and sees no technical problem. Resident Inspector: Smith partic-ipated in most of the interview and indicated he was aware of the D/G loose fittings and sees no technical problem. I evaluated reasons why 6 personnel including were transferred to Unit 2 and this move does not appear to be discriminatory.
. .. j' 0; ,
67 Inspector Name: A-4 Date Interviewed: General
Background:
Interviewee Comments: Uncomfortable with some use-as-is situations, e.g., cable separation problem found in fuel building during walkdown did not meet procedure but was evaluated as use-as-is. He can show someone where it is. Wrote NCR on lack of 5-thread engagement on a conduit fitting poor evaluation in that they simply said that couldn't see it; a second NCR was written on this area for cable damage, seemed to be looking for a way to buy this area off, took two tries to get everything evaluated, knows about this but didn't get back to him on fact that NCR's were poorly handled, i.e., non-tech-nical aspects.' Feels discriminated against in that he was transferred to Unit 2 where there is no overtime. Got grilled on cable damage NCR at the same time as being counseled on a personnel issue so it sppeared that his transfer had something to do with NCR. Management is aware of this concern. Note: I did not review this person's trar.sfer situation. b
'*' ^
68 Inspector Name: A-5 Date Interviewed:
' General
Background:
Interviewee Comments: Had problems with post check, e.g'., loose lighting terminations and junction boxes. Took lighting out of procedure and made it more difficult to look at junction boxes. Management was made aware of these concerns. (Has no significant safety concern) More tendency toward use-as-is when pressure is on (safety requirements are being met, however) Has had some fear of talking with NRC, didn't think reporting on-site would ever get off-site, doesn't have NRC RIV phone number Feels discriminated against by being transfarred to Unit 2 Some NCR evaluations are inaccurate or unclear, e.g., statement that workmanship was not compromised when in fact workmanship was poor but the item was technically acceptable Notes: I reviewed the transfer situation; appears to be reasonable but not as clear as reasoning on other 5 transfers. NRC Form 3 appears well posted so I'm not sure why he doesn't nave. the number. He does not appear to fear talking with NRC now. Although, he stated he does not have significant safety / quality concerns, his comment on NCR answers is interesting. Similar general comments were received from other inspectors and this could indicate a need for better answers on NCR's. An example would be that if a workmanship question was not addressed properly then ~perhaps needed retraining of personnel as preventive action would not get performed. Perhaps the licensee needs to improve in this area. e e s
a ( ', a , e , o; . .
'.'
- 69 Inspector Name: A-6 Date Interviewed:
General
Background:
Interviewee Comments: Added higher sides to some cable trays to keep cables in trays Also there may be cable density / compaction problem in this area It's tough to keep people off trays to keep from damaging them Have had problems with clearance of pipe and cables, have to notch insulation, place metal between insulation and trays There is alot of rework to get proper separation Notes: This man was questioned primarily to get input for RIV review of cable spread room as to where there could be problems. He personally has little problem with plant quality. RIV - Martin was at the interview and verbal feedback on the first two items indicated that the situations were acceptable. e
Y - e-70 Inspector Name: A-7 Date Interviewed: General
Background:
Interviewee Comments: Had problems with Paper Flow Group (PFG), when first implemented, with completeness of packages. Getting'better and does not know of safety problem involved Some inaccurate NCR answers Site has problem with lost records, 2 people are assigned full time in the vault, NCR's are not written on lost records, reinspect when record is lost but this reinspection may be very difficult or very impractical. He has no evidence that reinspections are not getting done. This problem could relate to competance of PFG people, i.e. , maybe they lost records. Note: Various special team members looked quite extensively at records. Results are in the team report. O _}}