ML20136E534
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E WASHINGTON, D. C. 20555
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016 NOV 0 61985 Docket No. 50-445/50-446 Ms. Billie Pirner Garde Citizens Clinic Director Government Accountability Project 1555 Connecticut Avenue, Suite 202 IN RESPONSE REFER Washington, DC 20036 TO F01A-85-59
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.(now referred to as the Comanche Peak Project) established in March, 1984.
We are now placing three boxes of records in the civil / structural category into the Public Document Room (PDR) in file F01A-85-59 under your name.
In partic-ular, with the exception of those records indicated by asterisks on the enclosed Appendix C, which are undergoing further staff review, we are placing documents numbered I through 642 in tho PDR. A separate set of the records identified on Appendix C is also being made available for public inspection and copying at the Loca! Nblic Document Room located at the University of Texas Library, 701 South Cooper Street, Arlington, Texas.
We will communicate with you further as staff review progresses on the remaining documents subject to your request.
Si Jerely, f.xf
. M. Felton, Director Division of Rules and Records Office of Administration
Enclosure:
As stated 9
v 95tle/ o 6JO R/k k
+
APPENDIX C DOCUMENTS IN PDR FILE FOLDER F01A-85-59 Box 1 - Civil Structural Cat.1. Inadequate material used in concrete Paces
- 1.
2/15/83 Letter to alleger 1
Re: Attempts to comunicate 2.
6/8/79 NRC ltr to TUGC0 RE:
2 Transmittal of Inspection Report RIV 79-09 3.
5/24/82 Partial Testimony 3
4.
3/16/83 Appendix B 13 RIV 83-03 A 83-01 5.
No date Allegations and/or 2
Investigations Sumary 6.
No date Allegation Review I
7.
No date Civil / structural 1
Allegation Review Categories
- 8.
No date AC-27 84-006 1
(Partial testimony) 9.
No date Star - Telegram Article 1
" Utility Official denies Comanche Peak Charges" Cat. 2 Concrete Placement 10.
3/7/84 Telephonic Interview 1
11.
3/7/84 Partial Testimony AC-?3 1
12.
No date Allegations and/or 1
Investigations Sumary
2 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 13.
No date Civil / Structural Al-1 legation Review Categories Cat. 3 Poor Weather Conditions Placement of Concrete
- 14.
5/16/83 ASLB Hearing AC-35 8
Partial Testimony 15.
2/16/77 RIV IR# 50-445/77-01 9
& 50-446/77-01 16.
5/10/79 RIV #50-445/79-11 12
& 40-446/79-11 17.
6/12/79 TUGC0 ltr from Gary 3
to NRC W.C. Seidle Re:
Response to NRC Inspection No. 79-11 18.
7/5/79 RIV 50-445/Rpt. 79-11 1
& 50-446/Rpt. 79-11 NRC letter to TUGC0 Re:
Response to Notice of Violation 19.
9/17/79 TUGC0 Ltr from Gary 1
to NRC Siedle Re: Followup Response to NRC Inspection Report No. 79-11 20.
10/10/79 RIV 50-445/Rpt. 79-11 1
& 50-446/Rpt. 79-11 NRC ltr to TUGC0 Re:
Response to Notice of Violation 21.
5/24/82 AC-24 Partial Testimony 2
22.
03/11/83 Case Ltr pages 2 & 18 2
(listing of documents) 23.
8/19/83 RIV 50-446/83-24 17 50-446/83-24 (Cat. 42)
E 3
Re: F01A-85-59 APPENDIX C (Contir.ued)
Pages 24.
No date Civil Structural 1
Allegation Review Categories 25.
No date Allegations and/or 2
Investigations Sunnary Cat. 4 Concrecc Voids / Cracked / Crumbled
- 26.
3/7/84 Testimony AC-25 5
84-006 27.
12/20/77 RIV IR#
7 50-445/77-13,50-446/
77-13 ( Cat. 12) 28.
1/30/78 RIV IR# 50-445/78-01 8
& 50-446/78-01 29.
7/19/78 Brown & Root NCR 23 C-1034R1 30.
10/11/79 NRC Memorandum to 2
File
Subject:
Allegations Regarding Construction Deficiencies Comanche Peak Station, Units 1 & 2 DN 50-445; 50-446 31.
11/8/79 RIV IR# 50-445/79-26 7
& 50-446/79-25 32.
3/5/80 Brown and Root 32 NCR# C-1824, Rev. 2 33.
5/2/80 GH-6013 (Telefax from 2
TUSI to J.T. Merritt/R.E.
Heim/J.A. Allen)
Subject:
Stainless Steel Liner Repair 34.
8/22/80 RIV 50-445/80-16, 11 50-446/80-16 35.
5/24/82 Partial Testimony AC-41 1
l
4 Re: F01A-05-59 APPENDIX C (Continued)
Pages 36.
8/12/82 CPSES NCR C-82-01202, 7
Rev. 1 37.
1/12/83 TUSI ltr to Gibbs & Hill 2
Re: Voids Behind Unit 2 Reactor Cavity Liner 38.
1/20/83 GH-9413 TUGC0 Memo Re:
2 Voids behind Unit 2 Reactor Cavity Liner
- 39.
5/23/83 Report of Inquiry 3
Report No. Q4-83-0111 CPSES DCA 6663, R. 10 8
40.
7/16/84 41.
No date Civil / structural 1
Allegation Review Categories 42.
No date Allegations and/or 2
Investigations Sumary CP-82-13 43.
4 44.
10/20/82 TUGC0 Memorandum 4
Subject:
Comanche Peak Steam Electric Station CP-82-13 45.
10/26/82 TUSI memorandum 1
SDAR CP-82-13 Re:
Concrete void in Auxiliary Building 46.
10/29/82 TUGC0 ltr to NRC Re:
2 Auxiliary Building concrete floor slabs file No. 10110 CP-82-05 47.
- C-82-0058
5 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 48.
7/2/82 TUGC0 memorandum 2
to J.T. Merritt, Jr.
Subject:
Comanche Peak Steam Electric Station CP-82-05 49.
7/7/82 TUSI Memorandum 4
Kissinger from McBray Re: Concrete Void in Steam Gen &rator Wall Ref: TUQ-1314 50.
7/15/82 TUSI letter CPPA-21, 2
292 Re: Concrete void in Steam Generator Wall Ref:
1 SDAR-CP-82-05 2 NCR-C-82-00858 51.
7/20/f,2 TUSI memorandum 3
to Tolson from McBray Re: Concrete void in Steam Generator wall SDAR-CP-82-05 52.
7/23/82 TUGC0 Itr TXXX-3549 2
Re: Unit #2 Steam Generator Compartment wall 53.
8/24/82 TUGC0 ltr TXXX-3561 1
Re: Steam Generator Compartment Wall 54.
2/24/83 CASE's Provisional 3
Proposed findings of fact 55.
8/31/84 Documentation of tele-1 phone conversation by:
John X. Devers
\\
i 6
Re: FOIA-85-59 APPENDIX C (Continued)
~
Cat. 5 Concrete Pages 56.
9/28/83 NRC Inspection Report 7
50-445/83-27 57.
3/16/83 Report 50-445.83-03 13
& 50-446/83-01 J
- 58.
8/29/83 NRC Report of Inquiry 7
Q4-83-021
Subject:
Comanche Peak Steam Electric Station alleged intimidation of Electrical Craft Personnel 59.
10/4/79 RIV IR f 50-445/79-20 9
& 50-446/79-20 (Inspection Report) 60.
5/20/83 A4-83-005 A-19 9
Partial Testimony 61.
11/18/83 Memo for W.J. Dircks 1
from John T. Collins
Subject:
OIA Report 62.
No date Civil / Structural 1
Allegation Review Categories 63.
No date Partial Testimony 2
64.
No date A.26c page 40 (Taylor 1
NRCStatement) 65.
No date Allegations and/or 5
Investigations Sumary 66.
1983 Utility official denies 2
Comanche Peak charges Cat. 6 Rebar improperly installed or omitted
- 67.
4/10/79 Interview with alleger 2
I Re: FOIA-85-59 APPENDIX C (Continued)
Pages
- 68.
5/23/83 Report of Inquiry 3
- Q4-83-0111 (alleged poor construction practices)
- 69.
No date Testimony of alleger 4
70.
11/2/77 CPSES - Design change 1
2323-S-0757 71.
11/3/77 Brown & Root (NCR) 5 1
2323-SS-10 Rev. 3 72.
8/3/78 Brown & Root Memo 12 IM-15084
Subject:
Recall of WES-12 Revision 1 73.
4/30/79 Gibbs and Hill ltr 37 GTN-36257 Re: Reinforcing steel specification 74.
9/10/79 CPSES Design change 2
- 5536 - Rev. 1 75.
9/13/79 Gibbs & Hill 23 Dwg. changes
Subject:
TUSI: Containment wall 76.
9/18/79 NRC letter Docket No.
1 50-446 to TUGC0 Re:
TUGC0 Plans to resume the placement of concrete in the Reactor containment building exterior wall &
dome on September 18, 1979.
77.
11/9/79 NRC ltr to TUGC0 Re:
6 Transmittal of Inspection Report RIV IRf 50-445/Rpt.6
& 50-446/Rpt. 79-18 78.
11/8/79 RIV IRf 50-445/79-25 7
Inspection Report
1 8
)
Re: F0IA-85-59
)
APPENDIX C (Continued)
Pages 79.
3/14/83 Brown and Root Memo 37 IM-25214
Subject:
Recall of welding Procedure Specifications 80.
No date List of Drawings 2
Handwritten
- 81.
No date Partial Handwritten 3
AQC-12 82.
No date Interview AC-37 2
83.
No date Dwg. 2323-52-0505 1
Rev. 5 84.
No date Dwg. 2323-S1-0572 1
Rev. 4 85.
No date Dwg. 2323-5-0757 1
Rev. 12 86.
No date Dwg. 2323-52-0505 1
Rev. 5 V. TUGC0 QA Non Conforwence C.P. 6 87.
CP-77-6 88.
6/23/77 TWI-762 TUGC0 Memo 1
Subject:
Unit 1 Containment Rebar 89.
6/27/77 GTT-1259, TWX - 762 1
Gibbs & Hill memo
Subject:
Reactor Cavity Rebar 90.
7/6/77 Gibbs and Hill GTN -
2 19823 to TUGC0 Re:
Reactor Cavity Rebar Ref: 1 TWX-762 2 GTT-1259
9 Re: F0!A-85-59 APPENDIX C (Ccatinued)
Pages 91.
12/14/77 Gibbs & Hill TWX-2 1154 Re: Resolving changes to Drawings 92.
12/19/77 Gibbs and Hill DAX 1
123 to TUGC0
Subject:
Reactor Building No.1 Internal Structure Interface Problems
~
93.
12/28/77 CPSES Design change 1
832 94.
12/28/77 TUGC0 GRSE-DAX-133 1
To Gibbs & Hill
Subject:
Reactor Building No. 1 Internal Structure Interface Problems 95.
3/5/80 TUSI SITE GTT-4837 memo 1
Subj: Reactor Bldg. Unit
- 1 cavity rebar 96.
3/6/80 CPSES Design change 9
No. 6918 97.
No date Dwg. 2323-S1-0572 1
Rev. 3 Sh. I 98.
No date Dwg. 2323-S1-0572 1
Rev. 2 Sh. 1 99.
No date Dwg. 2323-S1-0572 1
Rev. 4 Sh. I 100.
No date Dwg. 2323-S1-0575 1
Rev. 4 Sh. 4 101.
No date Dwg. 2323-51-0574 1
Rev. 2 Sh. 3 102.
No date Dwg. 2323-S1-0574 1
Rev. 3 Sh. 3 103.
No date Dwg. 2323-51-0574 1
Rev. 4 Sh. 3 f
1
10 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 104.
No date Dwg. 2323-S1-0753 1
105.
No date Dwg. 2323-0785 1
Rev. 7 Sh. I 106.
No date Dwg. 2323-5-0751 1
Rev. 15 Sh. 7 107.
No date Dwg. 2323-5-0718 1
Sh. 1 108.
No date DC-003 Category 1
(Handwritten notes) 109.
No date TUGC0 TUQ-357 office 2
memo
Subject:
Comanche Pesk Steam Electric Station 1980-82 2300 MW Installation QA Non-confonnance Report CP-77-6 V TUGC0 QA Non-Confonnance CPP-77-10 110.
10/31/77 CPSES Design Change 1
477 111.
10/31/77 TUGC0 Conference 3
memorandum
Subject:
Construction Omission of Reinforcing Steel 112.
CP-77-10 113.
10/31/77 TUGC0 Memo 7
TUQ-425
Subject:
QA Non Conformance Report CP-77-10 114.
10/31/77 TUGC0 GTT-1697 4
to Gibbs and Hill
Subject:
Omitted Rebar in wall columns
11 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 115.
11/9/77 Brown and Root Memo 8
IM-11728
Subject:
Corrective Action for Installation and Inspection of Reinforcing Steel 116.
12/13/77 TUSI TUS-5032 1
Office Memo
Subject:
Comanche Peak Steam Electric Station 1981-83 2300 MW Installation QA Non conformance Reports Ref: TUS-5027 file: 10110 117.
No date Dwg. 2323-S1-0572 1
Rev. 4 118.
No date Dwg. SAB-00711 Sh. 2 1
l 119.
No date Dwg. SAB-00711 Sh. I 1
120.
No date Dwg. SAB-00711 Sh. 4 1
j 121.
No date Dwg. SAB-00711 Sh. 3 1
122.
No date Dwg. SAB-00718 Sh. 3 1
V TUGC0 QA Non-Conformance CP-77-11 123.
5/16/77 CPSES Design change 4
No 518 l
124.
10/6/77 Gibbs and Hill 4
GHF-2032 memo
Subject:
Construction Joints 125.
11/2/77 TUGC0 TUQ-429 memo 4
i
Subject:
QA nonconformance l
Report CP-77-11 126.
11/3/77 Brown and Root NCR 3
2323-55-10 i
1
,.____m
- - - - - - " - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ' - - - - - ~ ~ - - - - ^ - - - - - - - - - - - - - ~ - - - - - - - - -
- - ~ ^
12 Re: FOIA-85-59 APPENDIX C (Continued)
P_a ges 127.
11/4/77 Brown and Root IM-11688 2
Memo
Subject:
Corrective Action for Installation and Inspection of Rein-forcing Steel 128.
11/8/77 CPSES Design change 1
No. 518 129.
11/9/77 Brown and Root IM-11728 8
Memo
Subject:
Corrective Action for installation and Inspection of Reinforcing Steel 130.
11/10/77 TUGC0 TUQ-433 Office 6
memo
Subject:
QA non conformance Report CP-77-11 Revision 1 131.
11/11/77 Handwritten notes 1
132.
11/28/77 Brown and Root BRQ-0748 9
To TUGC0 Re: Possible Deficiencies Documented in NCR-C-806, 809, 810, 811 and 815 133.
No date Dwg. 2323-51-0622 Rev. 20.
I 134.
No date Dwg. 2323-51-0622 Rev. 19.
I 135.
No date Dwg. 2323-S1-0622 Rev. 17.
I 136.
No date Dwg. 558-1065 Sh. 2 1
137.
No date Dwg. SAB-0278 1
138.
No date Dwg. 2323-51-0622 1
Cat. 7. Concrete /Rebar Undocumented Activity / Rework
- 139.
No date Testimony of alleger 2
13 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 140.
9/9/76 Gibbs and Hill 13 SAB-103C Design changes
Subject:
Auxiliary Building Slab Reinforce-ment 141.
4/16/82 Memorandum for Roger 13 Fortuna from John T.
Collins,
Subject:
Alleged Coverup of Weld Defects at CPSES 142.
No date Dwg. BRHL-CC-1-137-2 700-E63R 143.
No date Slab Analysis 2
Cat. 8 False / wrong Documents
- 144.
No date Contact with A11eger 3
145.
3/14/77 Brown and Root De-8 ficiency and Disposition Report C-488, Rev. 1 146.
3/23/77 NRC ltr to TUGC0 Re:
20 Transmittal of Inspection Report RIV IR# 50-445/Rpt.
77-02 & 50-446/Rpt. 77-02 147.
5/10/79 RIY IR# 50-445/Rpt.
17 78-07 & 50-446/Rpt.
78-07 Inspection Report
- 148, 6/6/79 RIV IRf 50-445/79-09 37 50-446/79-09 Inspection Report 149.
5/24/82 Partial Testimony AQC-3 2
150.
5/24/82 Partial Testimony ACQ-2 2
151.
5/24/82 Partial Testinony ACQ-1 3
152.
No date Partial Testimony AQC-7 2
14 Re: F0!A-85-59 i
APPENDIX C (Continued)
Pages Available in Public Library 153.
4/4/79 Fort Worth Star-Tele-10 gram, " Area N-plant tests questioned."
Cat.9 154.
3/23/77 NRC ltr to TUGC0 Re:
6 Transmittal of Inspection Report RIV IR# 50-445/Rpt.
77-02, 50-446/Rpt. 77-02 155.
5/11/77 NRC ltr to TUGC0 6
Re: Response to apparent items of non-compliance RIV IRf 50-445/Rpt. 77-02 50-446/ Rpt. 77-02.
i 156.
5/27/77 NRC ltr to TUGC0 8
Re: Transmittal of i
Inspection Report RIV IRf 50-445/Rpt. 77-02, 50-446/Rpt. 77-06 157.
6/6/79 RIV IRf 50-445/79-09; 33 50-446/79-09 Inspection Report 158.
5/24/82 Partial Testimony 2
Cat.10 159.
6/8/79 NRC ltr to TUGC0 39 Re: Transmittal of Inspection Report RIV IRf 50-445/ Rpt. 79-09 4
l 160.
5/24/82 Partial Testimony 2
- 161.
5/23/83 Report of Inquiry 4
04-83-011
Subject:
Alleged Poor Construction Practices l
1
15 I
Re: FOIA-85-59 j
APPENDIX C 1
(Continued) i 162.
No date Standard Test Method 2
163.
No date Allegations and/or 2
Investigations Sumary Cat. 11 Seismic Design / Construction 164.
9/6/77 TUGC0 GTT-1543 Memo 1
^
Re: Expansion Joint Gaps s
165.
10/7/77 TUGC0 TUS-5012 office 1
)
memorandum
Subject:
i Elastic Joint filler 166.
12/20/77 RIV IRf 50-445/77-13; 7
50-446/77-13 Inspection Report 167.
1/30/78 RIV IRf 50-445/70-01; 8
50-50-446/78-01 Inspection j
Report i
168.
1/30/78 Gibbs and Hill Inc.
1 GHF-2390 Re: Elastic joint filler removal i
169.
2/13/78 Brown and Root QA Dept.
4 Inspection of Elastic Joint Filler Material Pemoval Check list 170.
2/19/78 Brown and Root memo 3
)
IM-12939
Subject:
CPSES Job 35-1195 Inspection of Rotofoam removal from the Seismic gap between the i
Auxiliary Building and Containment Building #1 171.
7/18/78 Brown and Root memo 1
IM-14835
Subject:
- CPSES, i
Job 35-1195 Deletion of various Quality Control Procedures and Inspections i
i 16 Re: F0!A-85-59 APPENDIX C (Continued)
Pages 172.
10/3/78 CPSES Inspection Report 3
Spec. No. 2323-554 173.
4/13/83 CPSES (NCR) 12 No. C-83-01067 174.
4/26/83 TUSI Separation of 1
Seismic Category I 175.
8/24/84 Gibbs and Hill GTN 83 69378
Subject:
Rotofoam filler between buildings 176.
No date Quality Instruction 5
CP-QCI-2-4-9 (Inspection of Elastic Joint Filler Material Removal) 177.
No date Brown and Root 5
Plan View 178.
No date Allegations and/or 1
Investigations Summary 179.
No date Cat. 11. Reference 1
Documents Cat.12. Concrete Structural Deviations / Tolerances 180.
No date Allegation Summary 1
Cat.13. Crack in Concrete pad beneath Reactor Vessel
- 181.
4/11/83 NRC Report of 5
Inquiry #Q4-84-016 AC-44 Cat. 14. Control Room 182.
1/28/76 Gibbs and Hill 14 GTN-6624 Acoustical Treatrent 183.
7/18/77 Gibbs and Hill 12 GTN-20085 Lowered Ceilings
j 17 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 184.
2/2/82 Calculations Gibbs 81
& Hill #SCS-171C Set 2 185.
3/16/82 Gibbs and Hill 17 Calculations
- SCS-174 C Set 2 186.
6/10/82 Gibbs and Hill 135 Calculations 187.
2/3/83 Processing CMC's 23 and DCA's 188.
10/4/83 Gibbs and Hill 22 Calculations SCS-174C Set 3 189.
1/5/84 Design Review.
22 Record From (Support Loads for Aircraft Cable Conduit System Due to Seismic Load.)
190.
1/16/84 Gibbs and Hill 217 Calculations SCS-226C Set 1 191.
8/3/84 TUSI memo to file 9
CPPA-40, 224 Control Room Ceiling 192.
No date Gibbs and Hill 13 No. DC-7, Rev. 9 193.
No date Documents Given by 1
Licensee i
Cat.15. Rebar improperly drilled 194 8/13/82 Brown and Root 9
from Dodd to Rice Re: Brown and Root Investigation l
l l
1 18 l
l Re: FOIA-85-59 APPENDIX C (Continued)
Pages 195.
2/3/83 Affidavit of 9
alleger
~
- 196.
3/28/83 IR-83-03 Statement 3
- 197.
5/20/83 Statement of 23 AC-40
- 198.
5/20/83 Alleged Improper 3
Construction Practices A4-83-005 199.
9/28/83 RIV 50-445/83-27 22
- 200.
8/2/84 Technical Interview 38 of alleger
- 201.
9/19/84 Technical Interview 26 202.
9/5/78 Gibbs and Hill ltr 24 GTN-29823 Rebar cutting 203.
11/10/78 CPSES CMC 2889 2
204.
2/2/80 Gibbs and Hill
Subject:
15 F. B. Base Anchorafie for Radweste solid'fication system 205.
6/17/81 Brown and Root Procedure 5
CCP-47. Requests for Rebar cutting 206.
8/11/81 TUGC0 CP-QP-11.2-1 4
Rev. 2 Surveillance and Inspection of Concrete Anchor Bolt Installation 207.
7/1/82 TUGC0 CP-QP-11.2-1 6
Rev. 9, Concrete Anchor Bolt Installation i
208.
2/23/83 Change verification 11 checklist for CMC's and DCA's
19 i
Re: FOIA-85-59 APPENDIX C l
(Continued)
Pages
- 209.
3/7/84 AQC-13 Telephonic 8
Interview 210.
No date Unauthorized rebar 82 cutting number of bits l
ordered.
l 211.
No date Hand notes - start 24 of new crew and new operation Rebar cutting Detail 212.
No date Category 15 Reference 2
Documents
~
213.
No date
. Allegations and/or 3
Investigation Sunnary 214.
No date Dwg. FSC-1000 1
215.
No date Dwg. 2323-5-0800, Rev. 5.
1 216.
No date Dwg. 2323-S-0801. Rev. 6.
1 217.
No date Dwg. 2323-5-0820, Rev. 6 1
Cat.16. Excavation / Backfill I
1
]
218.
2/9/82 Facsimile Transmittal 12 Sheet Region !Y I
- 219, 2/23/82 Regulatory Infonnation 18 Distribution Sheet l
l 220.
No date Affidavit of Robert C.
4 Stewart (NRC) l 221.
No date Affidavit concerning 7
l Contention 7 222.
No date Comanche Peak SES 1
i final Safey Analysis Report Units 1 and 2 fracture map Unit 1 foundation.
l
20 Re: FOIA-85-59 APPENDIX C (Continued)
Pages Cat. 17. Concrete Sampling 223.
3/28/83 In Reply to 50-445/
16 83-03,50-446/83-01
- 224.
3/31/83 Affidavit of alleger 5
Allegations 225.
7/29/83 Proposed initial 61 decision
- 226.
9/7/83 Assistance to In-2 spection Report A-4-83-005
- 227.
1/9/84 Report of Inquiry 9
Q4-84-001 228.
3/15/84 Clarification of open 21 issues Docket Nos.
50-445/50-446
- 229.
8/17/84 Civil / structural 2
Allegation Review Categories l
230.
7/28/82 Atomic Safety and Licensing 289 Board Hearing Transcript Allegations Contained in Star (Fort Worth. Texas)
I
- 231.
Undated Allegations Contained 4
Fort Worth Star Telegram
- 232.
4/5/79 Preliminary Notification 4
of Event or Unusual Occurrence
- 233.
4/10/79 Interview with alleger 2
- 234.
4/16/79 Contacts with allegers 2
235.
4/17/78 Inspection Plan 3
i i
m m-.
.m
21 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 236.
5/10/78 NRC ltr to TUGC0 13 Re: RIV 50-445/Rpt.
78-07, 50-446/Rpt.
237.
6/13/78 Memo from R.G. Taylor 2
to W. G. Seidle
Subject:
CPSES Response to deviation 78-07 238.
6/20/78 NRC ltr TUGC0 Re:
4 Response to Notice of Deviation RIV 50-445/Rpt.
78-07, 50-446.Rpt. 78-07 239.
4/4/79 Investigation Plan 6
Available in Public Library 239a.
4/4/79 Star Telegram, "N-site 6
tests questioned" 239b.
4/4/79 Star Telegram, "Arta 4
N-Plant tests questioned" 239c.
4/8/79 Star Telegram, "What's 3
happening at Comanche Peak?"
Researcher CASE (CPSES)
- 240.
8/28/79 Researcher Investigation 1
- 241.
No date Researcher Investigation 1
- 242.
8/14/79 Pending action control 4
Form 243.
No date Conclusions (draft) 2
- 244.
No date Fonn Letter 1
Available in Public Library 244a.
4/4/79 "N-site test questioned" 9
22 Re: F01A-85-59 APPENDIX C (Continued)
Pages CPSES Misc.
4
- 245.
4/4/79 NRC memorandum for file 2
~
from W.G. Hubaceek
Subject:
Allegations of 4
Construction Irregularities at the CPSES
- 246.
4/9/79 NRC memorandum for file 7
Subject:
Allegations of Construction Irregularities of CPSES, DN 5-0-445; 446
- 247.
4/16/79 Interview with alleger 4
248.
4/13/79 Travel Report Form 1
1 1
i 1
i i
4
)
i l
ll
_=
23 i
APPENDIX C DOCUMENTS IN PDR FILE FOLDER F01A-85-59 Box 2 - Civil Structural PAGES
- 249.
11/30/84 Sumary of status of alleger interviews 2
in the civil / structural area
- 250.
No date Contracts List 2
- 251.
No date List of allegers 38
]
- 252.
No date Sumary of the status of alleger interviews 2
and the actions in the Civil / Structural Area
- 253.
04/10/79 Interview with alleger 2
- 254.
6/28/84 ShiERDraft,Rev.2 13
- 255.
08/03/84 Interview with alleger 16
- 256.
10/31/84 Interview with alleger 13
- 257.
02/01/85 SSER Draft DCP11 9
- 258.
No date SSER Draft - Handwritten 5
- 259.
No date SSER Draft 6
- 260.
7:o date SSER Draft C/S 18 8
- 261.
No date SSER Draft 4
- 262.
No date SSER Draft 3
263.
02/01/74 Gibbs & Hill letter GTN-832 from R. E. Hersperger 58 to Robert W. Claude Texas Utilities Services Inc.
RE: Transmittal of two marked copies of specification 2323-55-9 " concrete" 1
264.
03/18/75 CPSES construction procedure 35-1195-CPP-10 12 Revision 0 Concrete Batch Plant Operations 265.
03/21/75 Gibbs & Hill letter GTN-2751 with attachment 80 to Texas Utilities Generating Company RE:
1980-82-2300 MW Installation G & H Project 4
No. 2323 05209 concrete 266.
04/21/75 CPSES Construction Procedures 35-1195-CPP-10, 28 Revision 1 267.
08/07/75 Concrete Batch Plant Operations Gibbs and Hill 48
24 Re: F01A-85-59 i
APPENDIX C (CONTINUED)
PAGES 268. 09/04/75 Concrete Specification No. 2323-55-9 Revision 2 20 Texas Utilities letter THG 1000 to Gibbs and Hill Document Status CPSES 269. 02/21/76 Information Copy - Monthly tests, General Labwork 11 270. 02/21/76 Information Copy - Beam Placement 5
3 271. 03/01/76 Gibbs & Hill letter-7185 to TUGCO-1900-82 62 2300 MW Installation G&H Project No. 2323 05209 1
Concrete 272. 03/09/76 CPSES Construction Procedure 35-1195-CPP-10 28 Rev. 2 Concrete Batch 273. 08/25/76 G&H Spec 2323-55-9 Rev. 4 219 274. 10/27/76 Comanche Peak Steam Electric Station Construction 27 Procedure 3S-1195-CPP-10 Rev. 3 Concrete Batch Plant Operations 275. 03/25/77 CPSES-Construction Procedure 35-1195-CCP-10 26 i
Rev. 4 276. 04/07/77 ConcreteTestingDataManagement(A-127) 23 277. 04/07/77 ConcreteTestingDataManagement(A-120) 14 278. 10/18/77 ConcreteTestingDataManagement(A-127) 35 279. 10/18/77 ConcreteTestingDataManagement(C-301) 12 280. 10/18/77 ConcreteTestingDataManagement(A-127) 39 281. 10/18/77 ConcreteTestingDataManagement(A-127) 30 d
282. 10/18/77 ConcreteTestingDataManagement(A-121) 20 283. 10/18/77 ConcreteTestingDataManagement(A-121) 14 284. 10/18/77 ConcreteTestingDateManagement(A-120) 35 285. 10/18/77 ConcreteTestingDataManagement(A-120) 30 286. 10/18/77 ConcreteTestingDataManagement(A-120) 25
25 Re:
F01A-85-59 APPENDIX C (CONTINUED)
PAGES 287. 10/18/77 ConcreteTestingDataManagement(A-121) 25 288. 10/18/77 Concrete Testing Data Management (A-126) 27 289. 10/18/77 Concrete Testing Data Management (A-126) 23 290.11/14/77 ConcreteTestingDataManagement(A-126) 28 1
291. 11/14/77 ConcreteTestingDataManagement(A-127) 43 292. 12/20/77 Concrete Testing Data Management (C-306) 20 293. 07/12/78 TUGC0 'QI-QP-11.0-1 Cadweld Inspection Activities 13 294. 07/12/78 TUGC0 QI-QP-11.0-6 Inspection of Grouting 4
295. 07/12/78 TUGC0 QI-QP-11.0-4 Sumer Concrete Curing Inspection 4
J
]
296. 07/12/78 TUGC0 QI-QP-11.0-2 Reinforcing steel, miscellaneous 3
steel and embedded item placement inspection 297. 07/12/78 TUGC0 QI-QP-11.0-5 Inspection of Concrete Repair 5
298. 08/03/78 Distribution Log QI-QP-11.0-3, Rev. 0 5
Concrete Placement Inspection 299. 08/16/78 TUGC0 QI-0P-11.02 Reinforcing steel, miscellaneous 4
steel and embedded item Placement Inspection 4
300. 08/17/78 TUGC0 QI-QP-11.0-5 Inspection of Concrete Repair 5
301. 10/10/78 Distribution Log QI-QP-11.0-4 Sumer Concrete Curing 5
i Inspection 302. 11/10/78 TUGC0 QI-QP-11.0-1 Cadweld Inspection Activities 21 303. 12/04/78 Brown & Root CPSES 35-1195-CPP-10 Concrete Batch 23 Plant Operations 304. 01/16/79 TUGC0 QI-QP-11.0-6 Rev. 1 Inspection of Grouting 5
305. 03/07/79 TUGC0 QI-QP-11.0-1 Rev. 2 Cadweld Inspection 21 Activities 306. 03/15/79 Concrete Testing Data Management (C-302) 10
26 Re:
FOIA-85-59 APPENDIX C (CONTINUED) 4 i
PAGES 4
307. 03/23/79 Gibbs & Hill letter GTN-35206 to TUGC0 RE:
69 transmittal of specification 2323-55-9, Rev. 5 308. 05/14/79 RIV Inspection Report 79-11 19 309. 06/25/79 QI-QP-11.0-1, Rev. 3 Cadweld inspection activities 21 i
310. 07/05/79 QI-QP-11.0-6, Rev. 2 Inspection of Grouting 4
311. 07/05/79 QI-QP-11.0-3, Rev. 1 Concrete or mortar 4
Placement Inspection 312. 07/05/79 QI-QP-11.0-4, Rev. 2 Summer concrete or mortar 4
curing inspection 313. 11/08/79 Investigation Summary Report No. 50-445/79-26; 7
50-446/79-25 314. 03/05/80 QI-QP-11.0-5, Rev. 2 Inspection of concrete 5
Rebar 315. 04/29/80 Gibbs and Hill letter to TUGC0 RE: Unit 2 138 Refueling Pool corewall concrete GTN-45945 including attachments 316. 10/28/80 QI-QP-11.0-6, Rev. 3 Cadweld inspection activities 3
317. 01/14/82 QI-QP-11.0-1, Rev. 4 Inspection of. grouting 22 318. 04/08/82 QI-QP-11.0-6, Rev. 4 Inspection of grouting 4
319. 06/04/82 QI-QP-11.0-6, Rev. 5 Inspection of grouting 4
4 320. 06/07/82 TUGC0 office memo
Subject:
Re-numbering / deletion 13 of procedures / instructions l
321. 07/02/82 QI-QP-II.0-3, Rev. 2 concrete or mortar placement 6
inspection 322. 07/02/82 QI-QP-11.0-2, Rev. 2 Reinforcing steel, misc. steel 5
and embedded Item placement Inspection J
323. 07/02/82 QI-QP-11.02, Rev. 4 Inspection of Cold weather 44 l
concrete operations
27 Re: FOIA-85-59 APPENDIX C (CONTINUED)
PAGES 324. 07/07/82 RIV IRf 82-11 Inspection Report 9
325. 07/21/82 QI-QP-11.0-3, Revision 3 Concrete or mortar 7
Placement Inspection 326. 11/23/82 Change Verification checklist for CMC's and DCA's -
11 G&H Job no. 2323 327. 12/13/82 QI-QP-11.0-5, Rev. 3 Inspection of Concrete Repair 7
328. 12/13/82 QI-QP-11.0-6, Rev. 6 Inspection of grouting 6
329. 12/13/82 QI-QP-il.0-5, Rev. 3 Inspection of concrete 7
repair 330. 12/13/82 QI-QP-11.0-3, Rev. 4 Concrete or mortar placement 7
inspection 331. 05/18/83 NRC letter to TUGC0 RE: Response to notice of 4
violation RIV IRf 50-445/83-03 & 50-446/83-01 332. 08/24/83 RIV IR# 83-24 & 83-15 Inspection Report 17 333. 09/28/83 RIV IRf 83-27 Inspection Report 22 334. 12/07/83 QI-QP-11.0-2, Rev. 3 Reinforcing steel, misc.
5 steel and Embedded Item Placement Inspection 335. 12/27/83 QI-QP-11.0-12, Rev. 12 Inspection of Grouted 44 Base Plates and Equipment Bases 336. 06/05/84 Technical Review Team Guidance 19 337. 06/18/84 Deletion of QU-QP-11.0-10 8
338. 07/12/84 TUGC0 office memo to Dan Hicks
Subject:
Historical 32 files audit 339. 07/24/84 QI-QP-11.04, Rev. 3 Sumer Concrete or Mortar 5
Curing Inspection 340. 08/17/84 Civil / Structural Allegation 2
28 Re: F0!A-85-59 APPENDIX C (CONTINUED)
PAGES 341. 09/10/84 QI-QP-11.0-9, Ray. 2 Soil Backfill 16 342. 09/11/84 QI-QP-11.0-2, Rev. 4 Reinforcing steel, misc. steel 5
and embedded Item Placement Inspection 343. 09/12/84 QI-QP-11.0-8, Rev. 2 Cnnerete Production 14 344. 09/24/84 QI-QP-11.0-1, Rev. 5 Cadweld Inspection Activities 22 345. 09/26/84 GAP letter to NRC (Darrell G. Eisenhut) 8 Regarding: Request for Further Information 346. 10/11/84 Memora'ddum to George Lear from David C. Jeng, 2
Subject:
status on resolution of Comanche Peak Civil / Structural allegations 347. 10/16/84 NRC letter to G. Lear from L. C. Shao
Subject:
1 Need for D. C. Jeng's continued effort in Comanche Peak Allegation Resolution 348. 10/23/84 NRC Memo to Darrell G. Eisenhut from R. C. Tang 64 RE: TRT Draft 349. 11/07/84 Meeting with CASE Volume I 91 350. 11/07/84 Meeting with CASE Volume II 91 351. 11/09/84 NRC letter from Vincent S. Noonan to Darrell G.
31 Eisenhut
Subject:
Comanche Peak Action Items -
Schedule Update 352. 01/08/85 NRC Memorandum from B. J. Youngblood to Distribution, 81
Subject:
Review and comments on Comanche Peak SSER #5 relating to CYGNA Independent Assessment Programs 353. 01/21/85 SRI X-FE-108-1 Schmidt Hammer test on concrete at 8-the Comanche Peak Steam Electric Station 354.01/29/85 NRC memo to Distribution from Vincent S. Noonan, 32
Subject:
Freedom of Information Act Request 355. 01/30/85 File List 2
356. 02/01/85 Category #18 Reinforcing steel 13
29 Re:
F0!A-85-59 APPENDIX C (CONTINUED)
PAGES 357. 02/07/85 Routing slip with attached note concerning (FOIA) 2 358. 02/25/85 Memorandum for Vince Noonan from L. C. Shao 13
Subject:
Review of Design QA Draft 1 359. No date CPSES Organization Charts 7
360. No date NRC Technical Review Team 2
361. No date Connents from J. Scinto, OELD 15 362. No date Defici,e,nt data sheets 16 363. No date Suggest Areas for Contention V Panel Site Visit 2
in Civil / Structural Discipline 364. No date BNL Review of Texas Utilities Generating Company 24 Comanche Peak Steam Electric Station Upper Lateral Restraint Beam - Stear. Generator 365. No date NRC memorandum for Comanche Peak Team Members from 5
Vince Noonan
Subject:
Freedom of Information Act Requests Drawings 366. No date SCB-10512, Rev. 1 Sheet 3 1
367. No date SCB-10512, Rev. 1, Sheet 4 1
368. No date SCB-10519, Rev. 8, Sheet 1 1
369. No date SCB-10519, Rev.10 Sheet 2 1
370. No date SCB-10519, Rev. 4. Sheet 4 1
371. No date SCB-10519, Rev. 6, Sheet 5 1
372. No date SCB-10519, Rev. 4, Sheet 6 1
373. No date SCB-10519 Rev. 1, Sheet 7 1
374. No date SCB-10519, Rev. 6, Sheet 3 1
375. No date SCB-10522, Rev. 3, Sheet 1 1
30 Re: FOIA-85-59 APPENDIX C (CONTINUED)
PAGES 376. No date SCB-10522, Rev. 1, Sheet 2 1
377. No date SCB-10522, Rev. O, Sheet 3 1
378. No date SCB-10522, Rev. 8, Sheet 5 1
379. No date SCB-10525, Rev. 10, Sheet 1 1
380. No date SCB-10525, Rev. 7, Sheet 2 1
381. No date SCS-10528. Rev. 4. Sheet 1 1
~
382. No date SCB-10530. Rev. 6, Sheet 1 1
383. No date SCB-10530. Rev. 5 Sheet 2 1
384. No date SCB-10534, Rev. 6. Sheet 1 1
385. No date SCB-10534, Rev. 3, Sheet 2 1
386. No date SCB-10545, Rev. 1, Sheet 1 1
387. No date SCB-10545, Rev.1, Sheet 2 1
388. No date SCB-10545, Rev. 1, Sheet 3 1
389. No date SFB-00800, Rev. 7, Sheet 1 1
390. No date SFB-00800, Rev. 4, Sheet 2 1
391. No date SFB-00800, Rev. 5, Sheet 3 1
392. No date SFB-00800, Rev. 5, Sheet 5 1
393. No date SFB-00803, Rev. 6, Sheet 1 1
394. No date SFB-00803, Rev. 3, Sheet 2 1
395. No date SFB-00806, Rev. 2 Sheet 1 1
396. No date SFB-00806, Rev. 2, Sheet 2 1
397. No date SFB-00809, Rev. 1, Sheet 1 1
398. No date SFB-00809, Rev. 2, Sheet 2 1
- -. ~ _
31 Re: FOIA-85-59 APPENDIX C (CONTINUED)
PAGES 399. No date SFB-00811, Rev. 3, Sheet 1 1
400. No date SFB-00815, Rev. 6, Sheet 1 1
401. No date SFB-00815, Rev. 8. Sheet 2 1
402. No date SFB-00815, Rev. 4, Sheet 3 1
403. No date SFB-00815, Pev. 3, Sheet 4 1
404. No date SFB-00815 Rev. 4, Sheet 5 1
405. No date SAB-00700, Rev. 6 Sheet 1 1
406. No date SAB-00700, Rev. 9. Sheet 2 1
407. No date SAB-00700, Rev. 13, Sheet 3 1
408. No date SAB-00700, Rev. 2. Sheet 4 1
409. No date SAB-00703, Rev. 6, Sheet 1 1
410. No date SAB-00703, Rev. 14, Sheet 2 1
411. No date SAB-00703, Rev. O Sheet 3 1
412. No date SAB-00703, Rev. 4. Sheet 4 1
413. No date SAB-00711, Rev. 10, Sheet 1 1
414. No date SAB-00711, Rev. O, Sheet 2 1
415. No date SAB-00711, Rev. 11, Sheet 3 1
416. No date SAB-00711, Rev. 10, Sheet 4 1
417. No date SAB-00714, Rev. 5, Sheet 1 1
418. No date SAB-00714, Rev. 9, Sheet 2 1
419. No date SAB-00718, Rev. 3, Sheet 1 1
420. No date SAB-00718, Rev. 9, Sheet 3 1
421. No date SAB-00718, Rev. 5, Sheet 4 1
32 Re: FOIA-85-59 APPENDIX C (CONTINUED)
PAGES 422. No date SAB-00721, Rev. 6, Sheet 1 1
423. No date SAB-00721, Rev. 9, Sheet 2 1
424. No date SAB-00721, Rev. 4, Sheet 3 1
425. No date SAB-00728, Rev. 4, Sheet 1 1
426. No date SAB-00728, Rev. 5. Sheet 2 1
427. No date SAB-00728 Rev. 3, Sheet 3 1
428. No date SA3-00728, Rev. 2, Sheet 4 1
429. No date SAB-00731, Rev. 3, Sheet 1 1
430. No date SAB-00731, Rev. 8, Sheet 2 1
431. No date SAB-00735. Rev. 3, Sheet 1 1
432. No date SAB-00735, Rev. 5 Sheet 2 1
433. No date SAB-00738, Rev. 3, Sheet 1 1
434. No date SAB-00738, Rev. 2, Sheet 2 1
435. No date SAB-00741, Rev. 3 Sheet 1 1
436. No date SSW-001107, Rev. 13 Sheet 1 1
437. No date SSW-001107, Rev. 4, Sheet 2 1
438. No date SSW-001107, Rev. 3, Sheet 3 1
439. No date SSW-001107, Rev. 6 Sheet 1 1
440. No date SSW-001107, Rev. 2, Sheet 4 1
441. No date FSC-00295, Rev. O, Sheet 1 1
442. No date SPT-10318, Rev. 1, Sheet 1 1
443. No date SPT-10318, Rev. 1, Sheet 2 1
444. No date SPT-10318 Rev. 2, Sheet 3 1
33 Re: F01A-85-59 APPENDIX C (CONTINUED)
PAGES 445. No date FN-PS-35, Rev. 1, Sheet 4 1
446. No date FN-PS-36, Rev. 2 Sheet 5 1
447. No date FN-SCR-37, Rev. 2, Sheet 4-1 1
448. No date FN-SCR-39, Rev. 2, Sheet 4-3 1
449. No date FN-SCR-40, Rev. 2, Sheet 4-4 1
450. No date FN-SCR-42 Rev. 2. Sheet 4-6 1
451. No date FN-SCR'-44, Rev.1, Sheet 4-8 1
452. No date FN-SCR-48, Rev. 1, Sheet 4-12 1
453. No date FN-SCR-49, Rev. 1, Sheet 4-13 1
454. No date FN-SCR-59, Rev. 2, sheet 4-23 1
455. No date FN-SCR-71, Rev. 1, Sheet 4-35 1
456. No date FN-SCR-72 Rev. 1, Sheet 4-36 1
457. No date 2323-51-0519, Rev. 4 1
458. No date 2323-51-0520. Rev. 4 1
459. No date 2323-51-0521, Rev. 3 1
460. No date 2323-51-0541, Rev. 4 1
461. No date 2323-51-0600, Rev. 12 1
462. No date 2323-51-0602, Rev. 13 1
463. No date 2323-51-0604, Rev. 11 1
464. No date 2323-51-0605, Rev. 11 1
465. No date 2323-S1-0607. Rev. 7 1
466. No date 2323-51-0608, Rev. 5 1
467. No date 2323-S1-0610, Rev. 3 1
34 Re: FOIA-85-59 APPENDIX C (CONTINUED)
~
PAGES 468. No date 2323-S1-0611, Rev. 6 1
~ ' - -
469. No date 2323-S1-0613, Rev. 2 1
470. No date 2323-51-0614, Rev. 4 1
471. No date 2323-S1-0616, Rev. 4 1
472. No date
?.323-S1-0617, Rev. 3 1
473. No date 2323-S1-0618. Rev. 10 1
474. No date 2323-St-0622, Rev. 20 1
475. No date 2323-S1-0623, Rev. 18 1
476. No date 2323-S1-0624, Rev. 21 1
477. No date 2323-51-0625, Rev. 21 1
478. No date 2323-S1-0626, Rev. 21 1
479. No date 2323-S1-0627, Rev. 13 1
480. No date 2323-51-0628. Rev. 13 1
481. No date 2323-51-0629, Rev. 5 1
482. No date 2323-S-0785, Rev. 7 1
483. No date 2323-S-0786, Rev. 9 1
484. No date 2323-S-0802, Rev. 3 1
485. No date 2323-5-0819, Rev. 5 1
486. No date 2323-S-0821, Rev. 4 1
487. No date 2323-S-0826, Rev. 6 1
488. No date 2323-S v835, Rev. 6 1
1 489. No date 2323-5-1118, Rev. 4 1
490. No date 2323-S-0107, Rev. 6 1
35 APPENDIX C DOCUMENTS IN PDR FILE FOLDER FOIA-85-59 Box 3 - Civil Structural 1975 Pages 491.
3/11/75 Structural Calculations 18 (SSB-106C2) 492.
6/01/75 R. W. Hunt " Inspection &
27 Testing Procedural Manual for CPSES Rev.2 493.
7/31/7'S Deficiency & Disposition 3
Report (DDR)#C-67 494.
8/8/75 Deficiency & Disposition 2
Report DDR f C-66 495.
9/02/75 Deficiency & Disposition 2
Report DDR #C-106 496.
9/02/75 Deficiency & Disposition 1
Report DDR #C-107 497.
9/10/75 Ltr from NRC (Madsen) to 12 TUGC0 (Brittain)RE: TUGC0 pro-posed corrective actions relative to items of non-compliance 498.
10/6/75 Deficiency & Disposition 1
Report DDR #75 499.
11/24/75 Structural Calculations 96 1976 500.
2/4/76 Hand Calculations 13
" Safeguards Bldg"
36 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 501.
2/12/76 NRC Ltr from Madsen 16 to TUGC0 Brittian RE:
acknowledgement of ltr responding to corrective actions 502.
2/17/76 Deficiency & Disposition 1
Report DDR f C-219 503.
2/17/76 Deficiency & Disposition 19 Report DDR #C-220 504.
2/20/76 Deficiency & Disposition 3
Report DDR f C-222 505.
2/28/76 Deficiency & Disposition 7
Report DDR f C-225 506.
3/4/76 Cover Sheet Set 12 55 Structural Calculations 507.
3/4/76 Cover Sheet Set 13 36 Structural Calculations 508.
3/9/76 Deficiency & Disposition 6
Report G-233 509.
3/16/76 Deficiency & Disposition 6
Report C-239 510.
3/16/76 Deficiency & Disposition 7
Report C-225, Rev. 1 511.
3/17/76 Cover Sheet Set 16 22 Structural Calculations 512.
3/23/76 Deficiency & Disposition 2
Report DDR C-246 513.
4/20/76 I.F. Inspection Report 5
Nos. 50-445/76-04 446/76-04
37 Re: FOIA-85-50 APPENDIX C (Continued)
Pages 514.
5/27/76 Cover Sheet Set 17 33 Structural Calculations 515.
7/27/76 Deficiency & Disposition 13 Reports DDR C-219. Rev. 1 516.
10/11/76 Concrete Testing Data 20 Management CP244 517.
10/11/76 Individual Test Data 2
28 day compressing strength (PSI) 518.
12/2/76 Deficiency & Disposition 12 Report DDR C-440 519.
12/9/76 Deficiency & Disposition 7
Report DDR C-446 520.
12/15/76 Deficiency & Disposition 50 Report DDR C-449 4
521.
12/27/76 Deficiency & Disposition 1
Report DDR C-443 R1 522.
12/28/76 Deficiency & Disposition 33 Report DDR C-457 1977 523.
2/14/77 Deficiency & Disposition 84 Report DD C-499 524.
2/28/77 Deficiency & Disposition 2
Report DDR C-515 525.
3/3/77 Deficiency & Disposition 2
Report DDR C-518 526.
3/11/77 Deficiency & Disposition 89 Report DDR C-529 527.
3/14/77 Deficiency & Disposition 8
Report DDR C-488, R1 l
38 Re: F01A-85-59 APPENDIX C (Continued)
Pages 528.
4/6/77 NCR Case Exhibit 528 2
- 529.
4/15/77 Sampson & Associates 1
(Sampson) Ltr to Robert H. Hunt Company (Kinkade) RE: Certification of Level I Inspectors 530.
4/26/77 Brown & Root NCR 13 Case Exhibit 12 531.
4/26/77 Brown & Root NCR 11 Case Exhibit 11 532.
4/26/77 Brown & Root NCR 5
Case Exhibit 9 533.
4/26/77 Brown & Root NCR 15 Case Exhibit 10 534.
4/26/77 Brown & Root NCR 12 NCR Case Exhibit 8 j
535.
5/24/77 Brown & Root NCR 78 C-642. R2 536.
5/24/77 Brown & Root NCR 86 Case Exhibit 529 C642 R1 537.
6/22/77 Brown & Root NCR 8
2323-SS-10-Rev3 538.
7/27/77 Brown & Root NCR 4
2323-SS-9 539.
7/29/77 Brown & Root NCR 9
2323-5S-10, Rev 3 540.
8/25/77 Brown & Root NCR 6
ACI-301-para-12.4
39 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 541.
10/13/77 Brown & Root NCR 6
2323-SS10 Rev3 542.
11/2/77 Brown & Root NCR 5
2323-55-10, Rev3 543.
11/2/77 Brown & Root NCR 1
2323-SS-10, Rev 3 544.
11/4/77 Brown & Root Memo 3
IM-11688 54 5.
11/8/77 Brown & Root NCR 3
2323-SS-10 Rev 3 546.
11/28/77 Brown & Root Ltr 9
BRQ-0748 547.
12/6/77 Brown & Root NCR 3
35-1195 CCP-13 Case Exhibit 487 548.
12/20/77 Individual Test Dates 2
28 Day Compressive Strength (PSI) 549.
12/28/77 Brown & Root NCR 5
2323-SS-10 Rev 3 1978 550.
4/15/78 Brown & Root NCR 10 2323-S5-9 Case Exhibit 490 E51.
7/18/78 Brown & Root Memo 12 IM-14835
Subject:
Delegation of various quality control procedures and instructions
4J 4
Re: FOIA-85-59 APPENDIX C (Continued)
Pages 552.
7/27/78 CPSES/FSAR 278 2.5-1 Amendment 2 553.
9/22/78 Brown & Root 13 NCR C-1168 554.
9/26/78 Brown & Root 4
NCR C-1170 1979 555.
1/5/79" Brown & Root NCR 6
NCR i C-1303 556.
1/17/79 Brown & Root NCR 8
C-1314 557.
2/5/79 Brown & Root NCR 3
C-1335 558.
3/13/79 Brown & Root NCR 10 i
C-1367 559.
3/15/79 Individual Test Data 2
28 Day Compressive Strength (PSI) 560.
4/4/79 Brown & Root NCR 6
C-1389 561.
4/23/79 Brown & Root NCR 6
C-1418 1
Case Exhibit 523 562.
6/8/79 NRC ltr to TUGC0 RE:
transmittal of inspection report RIV 50-445/Rpt.79-09 38 50-446/Rpt 79-09 563.
7/18/79 Brown & Root NCR 4
C-1534 i
41 Re: F01A-85-59 APPENDIX C (Continued)
Pages 564.
8/30/79 Brown & Root NCR 4
C-1653 565.
10/2/79 Brown & Root NCR 3
C-1741 566.
10/30/79 Brown & Root NCR 2
C-1781 567.
11/9/79 RIV NRC LTR to TUGC0 9
RE: Transmittal of Inspection Report 50-445/Rpt. 79-26 50-446/Rpt. 79-25 1980 568.
3/13/80 Brown & Root NCR 2
C-1784, Rev. 1 569.
4/2/80 NRC Ltr to R.S. Gary 25 TUGC0 RIV 50-445/80-08 50-446/80-08 RE: Trans-mittal of Inspection Report 570.
4/8/80 Brown & Root NCR 1
l C-1784 Rev 2 571.
5/6/80 Brown & Root NCR 44 C-1784, Rev. 3 572.
5/6/80 Brown & Root NCR 20 C-1766 Rev 2 573.
7/2/80 TUGC0 Ltr from 16 R.J. Gary to W. C.
i Seidle RE: Inspection Report No. 50-445/80-11; 50-446/80-11 containing proprietary information
42 Re: F01A-85-59 APPENDIX C (Continued)
Pages 1981 574.
5/21/81 CPSES/FSAR 2
3.78-59 and 3.78-60 575.
l C-81-00243 1982 576.
C-82-00475 577.
6/2/82 Case's Response to 8
Applicants Motion for Summary Disposition of Case's Contention 5 578.
C-82-01079 579.
C-82-01432 580.
C-82-01432 R1 581.
C-82-01432 R2 1983 582.
2/24/83 Cases Provisional 3
Proposed Findings of Fact 583.
7/29/83 Proposed Initial Decision 61 (Concerning aspects of construction quality control.
Emergency Planning and Board questions)
ASLBP No. 79-430-060 Docket No. 50-446-OL, Docket No. 50-445-OL
43 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 1984
- 584.
10/22/84 Draft 6 SSER AQC-10 3
with written notes
- 585.
10/22/84 Draft 6 SSER AQ-64 3
- 586.
10/22/84 Draft SSER AE-17 6
587.
4/12/84 Tugco CPSES QI-QP-3 11.1-2 Rev 1 Disposition of concrete cylinders on placements of less than 50 cu. yd.
588.
4/12/84 TUGC0 CPSES 15 QI-QP-11.1-1 Rev 6 Laboratory testing and test frequency matrix 589.
8/14/84 Motions Regarding 267 ANI Documents /w attachments 590.
9/18/84 Technical Review 21 Team Briefing I
591.
9/21/84 Computer Printout 12
" Pours 2/10 to 3/30 592.
9/28/84 Comanche Peak Response 5
to10CFR50.57(c) 593.
10/5/84 Comanche Peak Response 32 Team Action Plan Item II.a Reinforcing steel in the reactor cavity 594.
10/5/84 Comanche Peak Response 5
Team Action Plan Item Number: I.C Electrical conduit supports
44 Re: F0!A-85-59 APPENDIX C (Continued)
Pages 595.
10/8/84 TUGC0 Ltr to Eisenhut RE:
4 Program Plan and Issue-specific action plan in response to the request for additional infomation by NRC staff Technical Review Team 596.
10/8/84 Program Plan and 19 Issue-Specific Action Plans 597.
10/19/Q4 TUEC Meeting with 64 NRC(staffagenda) 598.
10/29/84 Draft 4 CPSA 17 2.1.1-2.1.1.4 599.
11/1/84 NRC Memorandum for 9
Joseph Scinto from Vincent S. Noonan Subj:
Comanche Peak Technical Review Team (TRT)andRegionIVinput to the staff's response to Texas Utilities' 50.57(C) motion to load fuel and conduct certain precritical testing 600.
11/7/84 CPSES Plant Information 83 System (Printout) NIMS 601.
11/7/84 CPSES Plant Information 29 System (Printout) NIMS 602.
11/7/84 CPSES Plant Information 5
System (Printout) NIMS 603.
11/7/84 CPSES Plant Information 85 System (Printout) NIMS 604.
11/14/84 Case Ltr with attachments 56 (correction to attachment D to Case's answer to applicants motion for sumiry disposition relating to Richmond inserts)
45 Re: F01A-85-59 APPENDIX C (Continued)
Pages 605.
11/21/84 TUGC0 Ltr from Spence to 32 (NRC) Eisenhut RE: Transmittal of Revision 1 to Program Plan 606.
12/19/84 NRC Memorandum for 5
Vince Noonan from James P.
Knight,
Subject:
Comanche Peak Steam Electric Stations 1 and 2 Resolution of open items from Cygna hearings
- 607.
12/19/84 Ltr to Noonan from Shao 9
Subject:
Comments on Action items identified B.J. Ellis of Case during November 7, 1984 meeting 608.
12/28/84 Tenera Corporation Ltr 7
to Vince Noonan RE: TRT Civil / Structural Issues Summary of Planned Inspection and Testing Activities 609.
12/29/84 Ltr from Robert E.
3 Philleo, P.E., to David Jeng Regarding: Safety Issues 1985 610.
1/7/85 Agenda NRC-TRT 66 Meeting 611.
1/7/85 Memorandum for 8
Docket File from D.C. Jeng,
Subject:
Minutes of TRT/TUGC0 meeting held on January 7, 1985, at Comanche Peak Site 4
46 l
Re: F0!A-85-59 APPENDIX C (Continued)
Pages 1
612.
1/15/85 NRC Ltr to TUGC0 18
Subject:
Comanche Peak Response team action plan 613.
1/17/85 NRC Memorandum for 6
I Noonan from Parr
Subject:
Staff Position on Spent Fuel Pool Liner 614.
2/13/85 NRC Memorandum for 2
Mizuno from Shao
Subject:
Response to Case's Requests for Admissions Available in Public Library J
614a.
1/10/85 Washington Bureau 2
of News" 110 Potential Hazards of Comanche Peak cited in report" 614b.
1/10/85 Dallas Times 2
"NRC Blasts Comanche Quality Control" 614c.
1/10/85 Fort Worth Star-Telegram 2
" Reactor blasted in Report"
]
Undated Docur,ents 615.
Undated Pour Number Identification 1
Index 616.
Undated C-5 Data Sheet Instructions 6
l 4
617.
Undated CPSES Contention 5 Data Base Input Sheet 3
618.
Undated Request for Additional Information 14 c
I
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yr_
r., _ _ _ _ _,_ _. - -.-._.,
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47 Re: FOIA-85-59 APPENDIX C (Continued)
Pages 619.
Undated Case's Attachment D & Related 37 Infonnation 620.
Undated Assorted infonnation 24 concerning Rebar Cuts and amount of Rebar at CPSES 621.
Undated Assessment of Civil /
1 Structural Construction
- 622.
Undated Action Items in the C/S area 5
identified from Case 12/01/84 letter and 11/07/84 meeting in Texas 623.
Undated List of allegations 2
(handwritten) 624.
Undated Listofallegations(typed) 1
- 625.
Undated Summary of Civil / Structural 10 Group QA/QC-related findings 626.
Undated Schmidt Hamer Requirements 5
627.
Undated Evaluation of Item No. II b 1
628.
Undated Case Coments on Applicant's 16 10/8/84 program plan to respond to 9/18/84 Technical Review Team (TRT) Report 629.
Undated BNL Review of Upper Lateral 25 Restraint Beam-Steam Generator 630.
Undated Report showing why C/S SSER 6
was or was not referred to QA/QC TRT
48 Re: FOIA-85-59 APPENDIX C (Continued)
Pages
- 631.
Undated Report showing the following 9
per SSER: (1) Potential QA/QC Breakdown; (2)
Characterization of violation; (3) TEUC Action Required 632.
Undated Deficiency Data Sheet 4
633.
Undated Report on "A Conservative 4
Assessnent of the Structural Significance of unauthorized Rebar cutting at Comanche Peak Steam Electric Station" 634.
Undated Comanche Peak Open Issue 54 Action Plan DCPI -
Action Plan 17 635.
Undated Appendix K Status of Staff 20 Evaluation and resolution of technical concerns and allegations relating to civil and structural and miscellaneous issues regarding construction and plant readiness testing at Comanche Peak Steam Electric Station 4
Units 1 and 2
- 636 Undated Allegation Sunnary Cat.1-17 33 637.
Undated ANSI / ASTM C 172-71 3
Standard Method of Sampling fresh concrete 638.
Undated CPSES/FSAR 3.2 Classification 23 of Structures.
Components, and Systems
49 Re: F01A-85-59 APPENDIX C (Continued)
Pages 639.
Undated Major Contracts 9
640.
Undated Rebar Estimate CPSES 128 641.
Undated Crack in Base Mat of Unit 1 3
642.
Undated Case Exhibit 7:
1 Sketch of Base Mat Concrete Placement e
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G l/ERNMENT ACCOUNTADILITY PROJECT 1555 Connecticut Avenue. N.W., Suite 202 -
Washington. D.C. 20036 (202)232-8550 G:85 :104 January 21, 1985 FREEDOM 0F INFORMATION ACT REQUEST fiEdDifJ UF NOhMAliON ACT. REQUESI,
Fora-/S47 Director
~
Office of Administration Nuclear Regulatory Commission Washington, D.C.
20555 To Whom It May Concern:
Pursuant to the Freedom of Information Act ("F0I A"), 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 re-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 of ficial, " 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 surrounding records, including but not limited to a list of all records which have been or are destroyed and/or removed, a description of the 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 Feak nuclear power plant.
/ J't /
' lc w L.
(
~
)
Director Office of Ac.....ustration Page Two We are requesting the abov'e information as part of an ongoing monitoring project on the adequacy of the NRC's efforts to protect public safety and health at nuclear power plants.
For any documents or portions that you deny due to a specific FOIA exemption, please provide an index itemizing and describing the documents or portions of The index should provide a detailed jusitfication of your documents withheld.
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, b'
i0m~
Billie Pirner Garde Citizens Clinic Director 9
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U. S. NUCLEAR :tEGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 50-445/79-11; 50-446/79-11 Decket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 2
Investigation at: Comanche Peak Steam Ele::tric Station, Glen Rose, Texas Investigation Conducted: April 2-3 and April 13-23, 1979 Inspectors:
8//e[7f n-
. G. Taylor, Resicent Reactor Inspector, Projects Date Section k
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- 0. P. Tomlinson, Reactor Inspector Engineering Cate/
Support Section ( April 13, 1979, Interview )
(bckf hA. B. Beach, Reactor Inspector, Engineering Date /
f 4
Support Section (April 23, 1979, Interview)
Approved:
((/4[/f W. A. Crossman, Chief, Projects Section Da'te S obf R. E/ Hall, Chief. Engineering Support Section Dite /
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Investigation Sucrnary:
Investigation on April 2-3 and Aoril 13-23, 1979 (Recort 50-445/79-11; 50-446/79-11)
Areas Investigated: Special investigation of allegations received indi-4 cating snat concrete had been placed on the Unit 1 dome during a rainstorm l
in January 1979, without QC or documentation; that pipe with sandblasted-l off markings was being used in Unit 1; that steam system pipe was damaged l
t by a handling accident and covered up; and that welders were not being properly qualified.
The investigation involved thirty-six inspector-hours by the Resident Reactor Inspector and three inspector-hours by two Region IV based inspectors.
Results: The allegation relative to the concrete placement was confirmed (noncompliance - failure to implement the QA program - infraction).
No items of noncompliance or deviations were identified relative to the balance of the allegations.
1 i
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. ~ _ _ _
N INTRODUCTION i
j Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, are under construction in Somerville County Texas, near the town of Glen Rose, Texas. Texas Utilities Generating Company is the Construction Permit i
holder with Brown and Root Inc., as the constructor and Gibbs and Hill, Inc., as the Architect / Engineer, j
REASON FOR INVESTIGATION l
The Region IV Reactor Construction and Engineering Support Branch office received a telephone call from a fomer CPSES employee who reported i
several allegations indicating a potential breakdown in the CPSES Quality Assurance program.
j j
SUPNARY OF FACTS 1
1 On March 30, 1979, the Region IV Reactor Construction and Engineering Support Branch received a telephone call from a party who identified himself as a fomer CPSES employee. The call was taken by an on-duty Reactor Tnspector in the Projects Section who in turn provided the l
information to the assigned Resident Reactor Inspector at CPSES on i
April 2,1979. The allegations, as received on March 30, 1979, were:
4
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During a concrete pour on the Unit I containment dome in January j
1979, a rain occurred which washed away part of the concrete.
4\\j The affected area was repaired by the use of grout. Workers involved were requested to " keep it quiet." Two workers, who are still at the site, have knowledge of this occurrence.
2.
The identity of a lot of "Q" and "non-Q" pipe (6" or less) being used for Unit 1 has been lost due to obitteration of heat numbers i
i by sandblasting and loss of identifying tags. Workers are guessing as to the proper identification of the pipe.
2 A steam pide intended for the Unit 1 turbine fell off of a truck 1
3.
j and struck a railroad track.
It was taken back to a storage area 4
and hidden.'
4.
Third class helpers are being qualified in less than three months and are being used for safety related welding on Unit 1.
l 4
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On April 13, 1979, the Resident Reactor Inspector assigned to CPSES and accompanied by another Region IV inspector interviewed the alleger in an effort to obtain additional information on the allegations.
The additional information is summarized as follows:
1.
The concrete used for the repair was not grout as originally indi-cated but was known to contain gravel.
The concrete came from the batch plant where it was mixed on the ground and carried in a bucket to a tower crane at the Unit 1 Containment Building and hoisted to the dome area.
The work was accomplished sometime during the middle of the second shift, possibly around 10:00 to 10:30 p.m. (January 1979, no day specified).
2.
The pipe in question was not prefabricated pipe but rather bulk pipe joints. Sometimes, the pi occurrence not identified)pe is sandblasted on the outside (rate of which removes all of the heat marking used for traceability.
3.
The steam pipe was being moved during the second shift from the "Dodd's Spur" storage area to the plant area when it was dropped off the truck. A couple of the large " cherry-picker" type cranes were dispat:hed to the indicent to pick up the pipe and place it back on the truck.
The crew with the truck decided instead to put the pipe back into the storage area and leave it there for another shift to pick up and perhaps be blamed for: damaging the pi pe. The alleger did not know if the pipe had actually suffered any damage. He was aware the pipe in question was "non-Q" but expressed a concern that if the craft could get away with a cover-up on "non-Q," they probably are also doing it on the "Q" pipe and other equimpment.
4.
The alleger indicated he was concerned with what must be incompetent welders working on "Q" welds, since they could not have very much experience and still only be considered third class labor.
CONCLUSIONS Research of various records and interviews with both craft labor and Brown & Root QC personnel produced the folicwing conclusions:
1.
The allegation relative to the concrete placement on the deme of Unit 1 is essentially correct and is evidence of a breakdown in the licensee's Quality Assurance program. The incident will be considered an item of noncompliance..
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2.
The allegation relating to the loss of pipe traceability markings could not be confirmed. The Resident Reactor Inspector's finding was that on occasion the sandblasting, with attendant loss of readily visible markings, probably does occur through human error, but that there are other means which will re-establish the identity of the pipe without guessing on the part of the craft labor force.
3.
The piping in the "Dodd's Spur" storage area is for the turbine portion of the plant and is not safety related from a nulcear standpoint and is therefore not within the furisdiction of the NRC inspection program. The more generalized concern of cover-up cf improper handling practices is not consistant with the obser '
vations of the Resident Reactor Inspector and other NRC inspectors made during the course of routine inspections.
The allegation.
cannot be verified or refuted at this time, but should subsequent observations verify that the alleged situation is occurring, appropriate action will be taken.
4.
Welders are qualified in accordance with the provisions of the ASME Boiler and Pressure Vessel Code,Section IX, " Welding and Brazing Qualifications," as required by NRC regulations and the licensee's commitments as contained in the Safety Analysis Report submitted to obtain a Construction Permit.
The labor classifica-tion, and therefore the pay, of the welders is not an element of the ASME Code welder qualification program, only the ability of the person being tested to weld on a specified weld coupon.
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DETAILS 1.
Persons Contacted Non-Licensee or Contractor Persons The alleger is a former employee of Brown & Root (the site general contractor).
The person identified himself as a former equipment operator and foreman of equipment operators.
Principal Licensee Employees Construction Manager, Texas Utilities Generating Co.
Supervisor of Product Assurance, Texas Utilities Generating Co./
Gibbs & Hill Brown & Root, Inc.
Project General Manager Construction Project Manager General Foreman, Building Department Superintendent, Building Department I
Quality Control Inspector, Civil
)
2.
Preliminary Investigation - Aoril 2-3, 1979 i
a.
Allegation 1: The Resident Reactor Inspector (RRI) initiated a preliminary investigation of the allegation as soon as received. The RRI was aware that a number of concrete place-ments had been necessary to complete the dome area of Unit 1 and that a substantial portion of these placements occurred in January 1979. Schedule completion data indicated that five of the total of thirteen dome placements occurred in January 1979. Rainfall data for January was then obtained from the licensee's meteorology unit which indicated rain had fallen on Janaury 15,1979 (with the rainfall totalizer reset to zero) and again in the period between January 15 and 22,1979, when the totalizer was again zerced.
The data suggested that placement 101-8805-013, the final placement on the dome, was a
the most likely candidate since 2.72 inches of rain had occurred about the placement date of January 18, 1979.
The RRI then examined the QC inspection records for the placement which stated,
" Pour stopped at 8:00 p.m.1/18/79 due to inclement weather.
Pour was topped out all but to a 30' radius which was cleaned up and finished 1/19/79."
The RRI then interviewed the QC inspector of record for the placement and was informed that the placement had started under good weather conditions on January 18, but that the -
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weather subsequently developed into a light mist and drizzle 1
which did not interfere with the placement.
By late evening, the weather deteriorated further and became a full rainstorm with thunder and lightning.
By 7:30 p.m. or so it was decided that the placement would have to be stopped for reasons of personnel safety. The placement area was covered to keep the rain off the fresh concrete and the second shift was instructed to water blast and clean up the area so the placement could be resumed the following day.
b.
Allegations 2, 3 & 4: No attempt was made to perform a pre-liminary investigation of these allegations since the infor-mation was too vague.
l 3.
Licensee / Contractor Report of Allegations During the course of the above preliminary investigation, personnel of the licensee's management and QA organizations approached the RRI and stated that they too had received an allegation relative to the dome placement.
It was stated that licensee management had received a telephone call on or about March 19, 1979, on the subject and that licensee management had visited the alleger at his home on March 20,1979, to ascertain the facts of the allegation.
The alleger then was invited to visit the site and discuss the allegation, which the alleger is reported to have done on March 26, 1979. On the basis of these interviews, the licensee's Product Assurance personnel under-took an investigation which concluded that the allegation had no merit.
4.
Interview with Alleger by NRC Personnel The Region IV office made several attempts to establish contact with the alleger during the period following March 30, 1979, when the allegation-was received, through April 12, 1979, when the interview date and location were established.
The RRI and another NRC inspector met with the alleger and a friend on April 13, 1979.
The alleger provided the following information about himself:
a.
He had been employed by Brown & Root at CPSES for 2-1/2 to 3 years and had quit in mid-March because he was dissatisfied with how the night shift equipment operators were being dispatched and supervised.
b.
He had been an equipment operator, primarily on cherry-pickers, and also a foreman for equipment operators at an earlier time.
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He stated that he had made the a'llegations to licensee management and Brown & Root management earlier but had not been at all sat-isfied with the answers he had received to his allegations.
The alleger provided the following additional information relative to each of the allegations:
Allegation 1: The incident occurred well after the time that the placement had been stopped.
He could not be sure of the time but thought it was probably 10:00 to 10:30 p.m. when some equipment was dispatched to the concrete batch plant to bring down a bucket of concrete to Unit 1 and thought it strange.
The concrete was taken to the dome by a tower crane. He was sure that the concrete was not batched by the batch plant and certainly was not delivered by the usual concrete mix truck.
Allegation 2: The alleger made it clear that he was not referring to completed pipe spools but rather to bulk pipe.
The cherry-picker operators routinely move the pipe from one location to another on the site and that the pipe involved was bulk pipe or joints.
He stated that the pipe was sometimes sandblasted in such a way as to obliterate the heat number markings or tags and that he was pretty sure that there was a lot of unidentified pipe in the safety systems in Unit 1.
This sandblasting sometimes happened to various steel forms used to make supports.
Allegation 3: The alleger described being dispatched with his equipment out to "Dodd's Spur" to pick up a length of pipe that had fallen off a truck after being loaded.
The pipe had fallen on the spur railroad track. The RRI was not familiar with the tem "Dodd's Spur." The alleger stated it was the area were the turbine components are stored. When he (the alleger) arrived at the site of the incident, he was told not to reload the pipe on the truck but to take it back into the storage area and put it down.
The pipe crew indicated to him that they hoped that a day shift crew would come for the pipe and would probably be blamed for any damage that might have occurred to pipe when it fell.
He stated that he did not know if the pipe had been damaged. He stated that he knew it was "non-Q" pipe but thought the NRC should be aware that such things were going on at the site.
4 5.
Final Investigation - Aoril 16-23, 1979 a.
Allegation 1.
The RRI obtained the craft labor time sheets for both shifts for January 18 and 19,1979.
Review of the time sheets for the day shift on January 18 indicated that a portion of that shift worked on placement 101-8805-013.
The records indicated that the day shift was teminated at approximately..
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8:30 p.m. relative to the placenent as were the personnel at the concrete batch plant. The batch plant has no second shift operators.
The RRI found that a large number of people, well in excess of fifty, had then worked on the placement during a substantial portion of the second shift. One crew of twelve people was shown by the time sheets to have been placing concrete, a notation not consistent with the fact that the batch plant was closed during the shift. The RRI then utilized the time sheets to develope a list of persons to be inter-viewed in connection with the incident with special concentration on the persons listed on the time sheet indicating " placing concrete 101-8805-013." The B&R personnel office records indicated that eight of the ten names. included in this specific crew had been terminated at various times since January 18; the records did not suggest that any action was being taken to get rid of possible confinnatory personnel.
Late on April 17, 1979, two of the senior B&R construction manage-ment personnel very-informally asked the RRI bow the investigation of the allegations was coming along. The RRI responded that the on-site phase appeared to be complete and that NRC personnel would undertake the effort to locate and interview selected personnel innediately since it appeared that the allegation might be well founded. They asked the RRI if they could check with their people down to the General Foreman level as to the incident the night of '
January 18. The RRI indicated that such an inquiry on their part would probably not interfere with any future investigative action by the NRC.
On April 18, 1979, the licensee's Product Assurance Supervisor informed the RRI that he had information which indicated that the incident had occurred and that the craft General Foreman was the person responsible.
On April 23, 1979, the RRI, accompanied by another NRC Inspector, interviewed the General Foreman and his inmediate supervisor, the night shift B&R Building Department Superintendent.
These men related that on the night of January 18 the weather seemed to worsen and got to the point where the rain was so heavy that the people could hardly see.
The freshly placed concrete developed into a problem when the plastic cover could not take the rainfall water load.
Some of concrete began to sag back down the dome slope and one small area actually washed out and fell to the ground below.
These men related that they and their entire crew of up to about one hundred-fifty worked on into the night trying to save a very bad situation.
The sagged concrete was worked back into position and the crew protected it in any way they could to allow it to take a set.
-g-
The General Foreman went to the batch plant, got it open and operated the plant himself to make enough material to patch the washed out area. He stated that he found the design mix data used for the concrete on the dome and calculated the necessary weight of ingredients to prepare a half a cubic yard of cencrete.
The required data was put into the control system for the back-up dry batch plant, dropped into a skiff, and carried over to the quarter yard concrete mixer at the site test laboratory.
It was mixed in two batches and placed into a skiff and carried to the dome where most of the half yard was used as a patch in the washed out area.
Both the General Foreman and his Superintendent were aware that there were no Quality Control personnel around to observe any of these actions since they had all gone home when the weather got really bad. Both men related to the RRI a picture of almost panic proportions in which the presence or absence of Quality Control simply did not matter; they were going to save a concrete place-ment from what they considered a disasterous situation, regardless.
They indicated that while the night shift Assistant Construction Project Manager was generally aware of the situation on the dome that night, he probably was unaware of the fact that Quality Control personnel were not there or of the batching of the concrete under the conditions indicated.
In response to a question from the General Foreman as to "what happens now" the RRI stated that the NRC had no choice but to issue a Notice of Violation to the licensee since it had become very clear that the licensee's Quality Assurance program had broken down for the entire evening of January 18, 1979, and that a substantial amount of concrete on the dome was of an unknown quality.
b.
Allegation 2.
The RRI visited the paint shop sandblasting area during the course of the final investigation to ascertain if this allegation could reasonably happen.
The RRI interviewed a foreman of painters who is also in charge of the sandblasting activity and was told that three main categories of piping material routinely are sandblasted.
These are:
(1) Completed carbon steel spool pieces which are blasted on the outside prior to painting. The identity of these pieces is on an attached stainless steel band on which the identifying is encoded by stamping. Should the band come off, the spool piece identity can be re-established by the pipe fabrication shop since each spool is unique and is fully described by isometric drawings. -....
e e
(2) Carbon steel cut lengths, but otherwise in an unfabricated condition, are sent to sandblasting to have the inside cleaned prior to further fabrication. The outside, which usually carries the heat marking in paint is supposed to be untouched.
(3) Bulk carbon steel pipe materials used for making equipment stands and supports is blasted and painted prior to fabrica-tion. The material is used for such items as ir-trument supports.
The RRI found a number of examples of each of the above categories as well as steel shapes in the sendblast area.
During the tour of
-the area, the RRI did not find any material that could not be identified except that in category three. The RRI interviewed one of the sandblasting personnel and came to the conclusion that the person might make an occasional mistake on category 2 material since he seemed confused when asked what he was going to do with a number of pieces ready for him to work on.
It appeared that he might well blast the outside of a pipe when he should blast the inside.
Subsequent discussions with the paint shop foreman and with a Brown and Root Quality Control inspector in the pipe fabrication shop revealed that all cut, but unfabricated material, is trans-ferred to the paint shop by memo which details the size, schedule and length of the cut section and the pipe spool isometric drawing involved.
Should the outside of the pipe be inadvertently blasted, the piece can be raidentified relatively easy by measuring its size, schedule and length. The isometric drawing used to make the cut length is annotated with the pipe heat number prior to the cutting operation and verified by QC.
It appeared most unlikely to the RRI that two othenvise identical pieces but with different heat numbers would be inadvertently blasted within the same time period.
The RRI concluded that the allegers remark that " workers are guessing on the identity of pipe" might be true, but that there was an adequate cross-check system built into the quality assurance program to preclude untraceable pipe from being installed in the safety related systems.
All of the steel shapes used in safety related supports for pipe and cable tray that have been examined by the RRI and other NRC inspectors have been sufficiently marked to establish their origin.
These materials are also subject to a system of quality control verifications at various stages of fabrication sufficient to make it very unlikely that any improperly identified or unidentified material is used and installed..
9 e
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Allegation 3: Based on the interview with the alleger, no further action was taken to investigate the specifics of the allegation since the pipe in question was clearly not safety related and therefore not within the jurisdiction of the NRC inspection pro-gram. The more general concern that the pipe handling incioent was a possible indicator of the general attitude of the craft personnel, particularly the riggers and pipefitters, appeared to be unfounded. The RRI has observed durin the past nine months (since August 1978) g many plant tours over that the material hand-ling activities of the craft personnel have been accomplished under well controlled conditions in so far as they relate to safety related equipment and materials. An allegation of possible 4'
cover-up of improper actions by the craft personnel in behalf of other craft personnel is almost impossible to either confirm or canpletely refute.
d.
Allegation 4: No further investigation was made into the charge that third class welders are being used to perform safety related piping system welds on the basis that the welders are all qualified under a program prescribed by the ASME Soiler and Pressure Vessel Code Section IX, " Welding and Brazing Qualificatien." The applica-tion of the Section IX program has been reviewed a number of times by the RRI and other NRC inspectors since it'was implemented at CPSES. The implementation has been found to be consistent with the requirements.
These requirements, however, do not. address themselves to the experience or inexperience of the person seeking qualification as a welder, but rather to whether he can accomplish i
a weld in one or more of the Code prescribed positions that will pass the test criteria imposed by the Code.
The terminology " third class,' as it applies to the labor force, relates primarily to the pay category in which a person is hired and previous experience is a factor in this determination.
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TEXAS CTILITIES GENERATING CO.TIPANY June 12, 1979 TXX-2998
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Mr. W. C. Seidle, Chief Reactor Construction and Engineering Support Branch U. S. Nuclear Regulatory Commission Office of Inspection & Enforcement 611 Ryan Plaza Dr., Suite 1000 Docket Nos. 50-445/Rpt. 79-11 Arlington, Texas 76012 50 446/Rpt. 79-11 COMANCHE PEAK STEAM ELECTRIC STATION 1981-83 2300 MW INSTALLATION RESPONSE TO NRC INSPECTION REPORT NO. 79-11 DOCKET NOS. 50-445 & 50-446 FILE NO. 10130
Dear Mr. Seidle:
We have reviewed the report dated May 14, 1979 relative to the inspection conducted by Messrs. R. G. Taylor ard other members of your staff, of the activities authorized by NRC Construction Permit Nos. CPPR-126 and 127 for the Comanche Peak facility. We have responded to the finding listed in Appendix A of that report.
To aid in the understanding of our re:,ponse, we have repeated the requirement and your finding followed by our corrective action (s).
We believe the attached information to be responsive to the Inspectors' finding.
If you have any questions, please advise.
Very truly yours, R. J. Gary RJG:dla Attachment 1
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Inspection Report No.
3-11 TXX-2998
, Page 2 Appendix A NOTICE OF VIOLATION Based on the results of the NRC investigation conducted during the periods April 2-3 and April 13-23, 1979, it appears that certain of your activities were not conducted in full compliance with the conditions of your NRC Construction Permit No. CPPR-126 as indicated below:
Failure to Imolement the Ouality Assurance prooram For Civil Construction 10 CFR 50, Appendix B, Criterion II requires that a quality assurance program be established and implemented for the construction of the structures important to safety of the nuclear plant. The Texas Utilities Generating Company Comanche Peak Steam Electric Station Quality Assurance Plan affirms the intention to fulfill this require-ment. The CPSES " Civil Inspection Manual" provides a body of inspection and testing procedures required to implement the Quality Assurance Plan.
Contrary to the above:
On January 18, 1979, personnel of the civil construction labor force placed an undetermined amount of concrete of an unknown 44 quality on the dome of the Unit 1 containment without the knowledge b
of your Quality Assurance organization and without benefit of required inspections and testing of the concrete.
This is an infraction.
Corrective Steos Which Have Been Taken And The Results Achieved TUSI Engineering has retained the services of an established Materials and Concrete Consultant for the purpose of evaluating the inplace condition of that portion of the dome in question. The results of the Consultant's investigation and evaluation coupled with the Architect / Engineer's review of the investigation will be utilized to formulate required corrective action.
Corrective Stecs Which Will Be Taken to Avoid Further Noncoraaliance l
Construction concrete supervisory personnel have been informed that if a similar situation occurs, no additional concrete shall be batched or placed without prior notification to senior construction management and suitable arrangements for implementation of the established quality control program.
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i Insp;ction Report No.
9-11
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TXX-2998 Page 3 Date of Full Comoliance Corrective actions were initiated on May 23, 1979.
The results of the Consultant's evaluation and A/E's review are scheduled for completion by September 3, 1979.
Preventive measures were established verbally on April 17, 1979 with the personnel involved in the incident and extended to other personnel through supervisory meetings subsequent i
to that date.
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l July 5, 1979 In Reply P4fer To:
RIV Docket No. 50-445/Rpt. 79-11 50-446/Rpt. 79-11 l
Texas Utilities Generating Company ATTN:
Mr. R. J. Cary, Executive Vice President and General Manager 2001 Bryan Tower Dallas, Texas 75201 Centlemen:
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Thank you for your letter of June 12,1979,'in response to our letter dated May 14, 1979, and the attached Notice of Violation. We have no further questions at this time and we will review your corrective action during a 3
future inspection.
t Sincerely.
Origins! s'gied by l
Q W. C. Seidic b'
W. C. Seidle, Chief Reactor Construction and l
Eggineering Support Branch bec to Reproduction Unit bec to DAC:ADM I
for distribution 7/5/79 for distribution 7/5/79 AD/RCI (Rein =uth)
Central Files I
Standards Development LPDR i
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TEXAS UTILITIES GENERATING CO.ilPANY 2003 SRYAN TO% EM
- DALt A1 TEXAS 76204 September 17, 1979
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TXX-3043 Mr. W. C. Seidle, Chief Reacter Construccion and Engineering Support Branch U. S. Nuclear Regulatory Commission Office of Inspection & Enforcement 611 Ryan Plaza Dr., Suite 1000 Docket Nos. 50 445/Rpt. 79-11 Arlington, Texas 76012 50-446/ Rat. 79-11 COMANCHE PEAK STEAM ELECTRIC STATION 1981-83 2300 MW INST 4'.LATION FOLLOWUP RESPONSE TO NRC INSPECTION REPORT NO. 79-11 DOCKET NOS. 50-445 & 50-446 FILE NO. 10130
Dear Mr. Seidle:
As stated in our response on the subject inspection recort concerning the infraction for failure to implement the Quality Assurance Prcgram for Civil Construction, we have completed our Evaluation of the Unit 1 Containment Dome. As a result of a Material and Concrete Consultant's investigation and evaluation with the Architect Engineer's review, we plan no additional corrective action as the questioned concrete for the Unit 1 Containment Dome satisfies design requirements.
Records suoporting this determination are available for review at the CPSES site.
Very truly yours,
R.
. Gary RJG:dla
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l October 10, 1979 l
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Docket No. 50-445/Rpc'. 79-11 50-446/Rpt. 79-11
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Texas Utilities Generating Conpany ATTN:
Mr. R. J. Gary, Executive Vice l
President and General Manager 2001 Bryan Tower Dallas, Texas 75201 j
Gentlacien:
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Thank you for your letters of July 5 and September 17, 1979, in response to our letter dated May 14, 1979, and the attached Notice of Violation. We l
have no further questions at this time and we vill reviev your corrective i
action during a future inspection.
i s
l Sincerely, 1
Original signed by W. C. Seidje W. C. Seidle, Chief Reactor Construction and 0, h Engineering Support Branch n
Wh" bec's v/ite dtd 9/17/79, R. J. Gary to W. C. Seidle bec to Reproduction Unit bec to DAC:ADM l
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1 APPENDIX U. S. NUCLEAR REGULATORY CCfMISSION REGION IV I
NRC Inspection Report: 50-445/83-24 50-446/83-15
(
l Docket: 50-445 Ca tegory: A2 50-446 Licensee:
Texas Utilities Generating Company (TUGCO) 2001 Bryars Tower Callas, 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:
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8//f/83 R. G. Taylor, Senior Resident Inspector Cete' Construction (SRIC)
Aoproved:
b Y) n,&W 8l/ F/83
- 0. M. Hunnicutt, Chief Cate '
Reactor Project Section A Insoection Sumary l
Inscection Conducted March through July 1983 (Recort 50-445/83-24 and 83-446/83-15)
Areas Insoected: Special inspections, announced and unannounced, related to l
allegations mace to various flRC persons including the Atomic Safety and l
Licensing Board in their procedings regarding the operating license for Comanche l
Peak Station. The inspections involved 449 inspector-hours by one flRC inspector.
Results: The inspection confirmed the need to issue four violations initially icentified by the Construction Appraisal Team (CAT) (NRC Inspection Report l
50-445/33-13;50-446/83-12).
These involved the areas of HVAC, Ecuipment Installation, Document Control, and Storage of Eouipment.
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2 De tails 1.
Persons Contacted Princioal Licensee Emoloyees
'R. G. Tolson, Site QA Supervisor
- C, T. Brandt, Non-ASME QC Supervisor
- J. R. Merritt, Engineering, Construction and Startup ttanager
- J. B. George, Project General hanger l
- D. N. Chapman, QA Manager
'3. R. Clements, Vice-President, Nuclear Brown & Root (B&R) l
'G. R. Purdy, Project QA Manager
- 0. Frankum, Construction Project Manager l
The SRIC also interviewed many other licensee, B&R, and subcontractor personnel during the course of the inspection.
I
- Denotes those persons who attended one or more management interviews with l
the SRIC.
2.
Licensee Action on Previous Insoection Findines I
(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 i
water lines located several feet outside of the ASt1E code boundry point.
Tne device was engineered by the licensee's A/E and manufactured by NPS Industries.
Due to the overall si:e of the structure, it has been nick-named " George Washington Bridge" by the site labor and quality forces.
Tne licensce had reported the finding of the defects as a potential 50.55(e) item to the SRIC on September 30, 1982, which was subsecuently stated not reportable in a letter dated Cecember 27, 1982. An NRC inspector follcwed uo en 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 tne defects involved with the exception of two cracked welds that had not been analy:ed at the time of the inspection.
The engineer has recently analy:ed these two defects and has detemined that had they not been detected, tne l
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 tr,e functional application of the structure.
Since the structure has no centinucus service application and is essentially subject to a cne-time loacing, the cracks would not have the potential for further crocagatten, i
Further, the cracks are at points in the structure that would receive rela-tively low stresses in tne one-time impact based on tneir small si:e in relation to the members being welded.
It appears that the cracks formed due to the stresses develooed during the tightening of hign strength bolting in l
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1 the innediate vicinity of the welds during the site assembly of the structure, Taken in conjunction with the earlier documented review of the engineers i
calculations and the SRIC's review of thest cracks, the SRIC has concluded I
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-82-12) 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 nonconfomance report. The final disposition and closure of the NCR will be evaluated during future routine inspections.
3.
Review of Licensee Self-Evaluation (Usino INPO 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 INPO.
The evaluation was perfomed in behalf of the licen-see by personnel in the employnent of Sargent & Lundy, an architect-engineer fim with substantial nuclear power involvement. A copy of the report was furnished to the NRC, and subsequently, to the Atemic 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 determine 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 cf 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 deficiencies in design, engineering,(e). testing that would constitute conditions reportable under or 10 CFR 50.55 4
Car Wash In Containment During the limited appearance statement portion of the Atomic Safety and l
Licensing Board hearing on t'ay 16, 1983, a person stated at transcriot 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 develcped at about the same time that there was a meeting at the 0/Pd 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 tne meeting.
During this entire period, the Unit i reactor system was undergoing what is referred to as " Hot Func.
tional Testing".
This particular test is an accurate simulation cf ne coeration of tne reactor system and its appurtenances but withcut a reactor core being in olace.
The heat and pressure in the syste~. is generated oy l
the reactor coolant pumps in conjunction with the che-ical anc volume cen-l trol system charging pumps.
The test could readily te construed to be a pressure test out in fact is an operational test at pressure.
This parti-cular test extended overall for about 90 days beginning late in February L
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i 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 Containment 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 addition, there was a considerable amount of condensation crippino from the reactor coolant pump motor cooling coils.
This was caused by the cold water in the coils condensing the humidity from the atmosphere witnin the building 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 such a test and is one of the reasons for performing the test.
5.
':esien of the HVAC System Succorts By letters, both dated March 11, 1983 Citizens Association for Sound i
l Energy (CASE) notified the NRC's Offices of Inspection and Enforcement and i
the Executive Legal Director of a concern that the HVAC system for Ccmanche Peak had not been properly supported, nor had it been properly considered i
in regard to seismic load conditions or its treatment as potential mis-siles.
CASE specifically states that from their review of the FSAR, it appears that the licensee has not analyzed the HVAC supports for a seismic lead 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 en the HVAC l
systems within the containment.
CASE also states tnat all HVAC components and supoorts inside containment should be treated as missiles under Cri-terion 4 of the General Cesign Criteria for Nuclear Pcwor Plants, 10 CFR 50, Appendix A.
Sheet 21 of Table 17A of the FSAR lists the containment ventilation sys-tems as being Seismic Category II. Apparently, it has been assumec ::y CASE that this category excludes seismic loading in the design.
This assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seis..ic Category !! as being those portions of systems or C0mponents anose
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 P.egulatory 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 earthquake (y II and are designed and constructed so that SSE) will not cause such 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 whien could also mean that it might be impossible for workers 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.
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 consultant.
It was determined that all permanent HVAC supports are analyzed for seismic loading.
Two methods are utilized:
Zero Peak Accleration (IPA), 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 suoport 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 utili:ing two computer programs entitled FEASA-20 and FEASA-3D.
The acronym stands for frame eigenvalue and stress analysis.
The -2D version is used on the transverse supports and the -30 version is used on the axial suoports.
The inclusion of equivalent weights frem both up and downstream transverse supports and accesories such as vol-ume dampers and vane turns in the design of the axial supports was verified.
This inspection verified the adequacy of the siesmic design techniques being utili:ed for the design of HVAC supports at Cceanche Peak.
The concerns expressed by CASE have been found to be without eerit.
Dersons contacted during the course of the inspection at Corporate Censulting
t 6
& Development Company, LTD. were:
J. Roland Yow President & Chief Executive Officer Gary Hughes, Vice-President for Operations David Lindley, Principal Engineer Stephen Lehrman, Seismic Department Manager Daryl Hughes, Project Engineer 6.
Heating, Ventilation, and Air Conditioning System (HVAC)
During the CAT inspection (NRC Inspection Report 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 welding code requirements.
The dominate deficient condition 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 inch 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 2
.and gaskets between ducting sections were loose and/or missing.
'The CAT inspectors also found that some support 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 centractor.
Further,
a review of the licensee s audit program indicated that the licensee j
was unaware of these several problems in the fabrication, installation, and inspection of the HVAC systems.
Based upon the CAT inspectors' findings and his cwn observations, the SRIC recomended 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 Ecuipment The CAT inspectors noted during their inspecticns of certain major items of equipment that there were several variables in how the equipment was fastened to the building equipment pads.
In some instances, tanks for example, CAT inspectors fcund that tnere were two nuts (double nuts) en the embedded bolts securing tne equipment, other bolts had one nut, (single nut) and so e had a combination of f
both single nuts and dcuble nuts on one piece of equipment.
The CAT personnel also noted that certain heat exchangers had sletted holes in one of tne mounting bases to allcw for tnermal expansion during coeration.
The holddewn nuts appeared to be installed too tightly and may have prevented freedom of movement.
The SRIC ebtained the design and installation drawings for two of tne referenced heat excnangers identified in the CAT report.
Both were found to be horizontal Utube heat exchangers wncse 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 ccnstruction drawings for the mounting pedestals had a flat steel plate en ene
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 thes'e documents did not note or address the slotted details, the plate, or the fact the bolts 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 that each stud was double nuted and when the top nut was loosened, the second nut was approximately one flat (about 1/6 of a turn) from being fully tight.
This degree of looseness should allcw 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 nor were there any apparent provisions made to lock nuts where they must be deliberately lef t loose. This was considered overall to be a violation of Criterion 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 Ecuicment in Outdoor Storace Areas The CAT found that a considerable amount of equipment such as pipe support struts, clamos, 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 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 INPO Self-Evaluation Report at jage 111 describes essentially the sage finding.
This situation sas detemined to be a violation of Criterion XIII of 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 evicence that any items which indicated substantial deteriorstion from such stcrage conditions have in fact been installed in the nuclear power block.
It would accear that the various items involved i
have been cleaned and restored prior to installation such that tney can perform the required function.
9.
Cbsolete and/or Illecible Crawines In The Field The CAT inspectors found a group of drawings in one particular area adjacent to the control room that were found to be out of date by up to several issues and further, that some drawings in other areas were incomplete in the title and revision blocks.
The SRIC discussed
8 the finding with supervisory personnel of the licensee's central document centrol 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 distributien 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. A11eoations Relative Tq Imorecerly Sucoorted 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 souret to the effect that:
a.
There is field run conduit above the centrol room supported eniy 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 centrol room area and has examined the pertinent engineering drawings depicting both in relation to these allegatiens with the following findings:
a.
There is a censiderable amount of both safety-related and nensafety 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 sus-pcrts that are typical of these that the SRIC has cbserved in large cpen factories and are not designed to seismic standards.
In eacn 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 nonnal 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 fem a part of tne sleping wall area below the control room observation recm.
These cry-wall materials have been securely fastened to a metal frame work (metal batten) which in turn is succorted by conventienal and ncn-seismic straps and wires to the concrete primary builcing.
The 'ri e work is also attached to a system of stainless steel c3 Dies wnicn in turn also attach to the primary structure sucn tnat if normal su:-
ports fail during a seismic event, the weignt of the framing and drywall will be assumed by the cabling thus preventing the materials fran falling.
9 c.
The lighting fixtures in the control room are supported from an intermediate substructure of "unistrut" by light-weignt conduit.
The substructure is likewise supported by the same type cf 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 I
if the conduit or its attachment were to fail.
In the case of the actual light fixtures, the cable is attached to the light i
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 af ter a seismic event as Seismic Category I.
Items not having a safety function but whose failure could damage components which have a safety function j
or cause injury to the occupants of the control room during an event are referred to as Seismic Category !!.
In the case of the items involved in this allegation, all are Seismic Category !! since their falling could cause injury to the centrol operators.
The cabling system described can be expected to prevent such a fall even though the normal supports could pessibly 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 identified during this special inspection etfort.
i 11.
Placement and Curino of Concrete Curino Freezine Weather i
Curing the limited public appearance portion of the At::mic Safety and 1.icensing Board (Beard) hearing cenducted on May 15, 1983, tnere were two references to the placing of concrete in freezing weather at the Ccmanche Peak Statien which in turn lead to a question from the Board to the NRC staff as to whether there were any NRC personnel present with kncwledge of the matter.
The two references are at 6106 and 6134 of the hearing transcript while tne Board Question is at 6109.
Also at 6109, an uni-dentified voice responded to the Board that the matter had been recorted in IE inspectiet rep orts.
Researen of the NRC inspection recorts revealed that there had been such a discussion in NRC Inspection Recort 50 445/77 01 wnien was categorized as an unresolved item pending tne licensee's review l
and action on their finding of the problem.
The unresolved item was i
further discussed in NRC Inspection Report 50 445/77-04 with tne closure of the item by an improvement in the OA procedures.
10 l
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 !!,
paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had been prepared approximately 2 weeks earlier than the inspection j
period covered by the inspection report.
The first of the licensee's reports (C-134-17) was directed specifically to findings by a licensee inspector that the surface temperature of Concrete Placement 101-2808-001 some 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 Decerreer 30, 1976, being completed at approximately 6:00 p.m.
t.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 of 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 fro:en even though the surface temperatures were well below freezing.
The records also reflect that in order to resolve the issue, swiss hammer tests were run on the suspect areas af ter the concrete had fully cured.
These tests indicated that the suspect areas had attained strengths comparable to kncwn properly cured areas, indicating that the concrete had not been damaged even thcugh the possibility exists that it had been frozen for a perioc of time.
The records reflect that good concrete curing temperatures, i.e., above 400F were established and maintained shortly af ter the licensee's inspector's observation.
For the record, the SRIC would note that Placement 101-280f-001 took olace jh in the Unit 1 Reactor Building. The placement became the open area floor yJ at tne lowest full floor in the building.
This floor area, wnile suppor-A ting some equioment, serves primarily as a walk area.
As such, it is fully topped with an architural concrete making the structural concrete no lenger accessable.
NRC Inspection Report 50-445/77-01 also discussed ccm arable events to that documented on Surveillance Recert C 135-77. Cne of these events was tecu-ented by Surveillance Report C 063-76 on January 7,1976, and on St.R ceficiency/discosition reports (now titled nonconformance re: orts).
These documents indicate that on January 7,1976, the surface te":erature of placement 105 2773-001, the foundation basemat for tne Unit 1 Safeguards Building, were found frozen as evidenced by fro:en wet burtao over certain areas that were not covered by insulating blankets.
The receres also
11 reveal that the reported finding took place almost 7 days af ter the place-ment of the concrete. Althouoh 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 "u3e-as-is" en 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 free:e even though the burlap was froze.
This conclusion is considered valid by the SAIC based on his review of publications of the /cerican 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 l
allowed by the project specifications, swiss hammer tests were perfomed.
The swiss ham.er tests indicated the concrete placement had full specified s treng th.
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 HRC Inspection Report 50-445/77-01 included 00R-C 460 which in turn discussed low temperatures during the curing per.
iod of three separate placements that were made during the late Cecember time period of 1976.
In each case, the records reflect that the placements were accepted "use-as-is" since the least amount of cure time was 9 days, again witn good conditions until the cold weather occurred.
The NRC inspector involved in NRC Inspection Repurt 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 bout the quality of the placements involved.
Tne SRIC would note for the record that there are no regulatory or industry 1
prohibitions on placing concrete in cold weather conditions.
The /merican Concrete Institute and the Bureau of Reclamation both indicate that if the 0
fresh concrete is above 40 F at the time of placement, the chemical process of hydration will generate sufficient heat to prevent the concrete from free:ing provided that precautions are taken to prevent heat loss.
In mass concrete applications, the greatest danger to the concrete is en the exposed surface areas, particularily at corners and other edges of the place ent.
It would be exceedingly rare for the mass of tne concrete to free:e and sustain damage.
These publications also indicate that even if fro:en, the concrete will nor-ally cure to full design strengths if tem:eratures cen.
cucive to the hydration process are restored.
12.
Allegations pelative To *he As. Built verification and Cesign Verificatien Activities.
Curing Acril 1983, NRC personnel received allegations to the effec that
12 l
l the QA group performing as-built verifications were not measuring support member dimensions and therefore, the " Vendor Certified Orawings" of the supports would not be accurate. A second allegation from the same person l
indicated tMt 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 first allega-tien, the SRIC found that the allegation was and is factual.
The allegation, hcwever, does not appear to have any significant impact en safety in that the as-built inspection was not developed to assure that the " Vendor Cer-tified Crawing" 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 perfoming the final pipe stress analysis.
The responsibil-ity for assuring that the support members and other characteristics of the individual support reflect the design drawing requirements reside in other QA groups associated with the fabrication and installation efforts.
To also perfom these functions in the as-built verification inspection would be a redundant inspection that would not contribute significantly to the safety function of any given support.
Renarding the second allegation, the SRIC found that it too was factual but cnly at the specific time the allegation was made. When making the allega-tion, the alleger provided the NRC personnel with a reference to a OC inspecticn report which he said would fully display his concern.
This report, identified as IR DCV-00421, was found to contain notation that the verification was based en a computer tabulation and that the report was being ccr.pleted at the direction of the inspector's supervisor.
The original report was dated April 4,1983. The pemanent 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 notatien of having been based upon a cceputer tabulation and without notation of apparent protest of directicas 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 OC design verification program effort.
The inspector stated that the attemot to use the ccecuter based data in the perfor-ance of the assigned task was in error from the beginning because of errors by persons generi-ting the cceputer data. The interviewee stated that cnly the ene verifica-tien effort had been done using the ccmouter based data and tnat all price and subsequent verifications have been done by tne assignec insoectors directly and personally examining the existent quality records in compli-ance with applicable OC procedures for the task.
He stated that the only
13 l
procedural deviation was the one instance stated in the allegation. Dis-cussions between the group supervisor at the time the allegation was 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 l
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-l dures.
No violation to NRC requirements were revealed during this special inspection effort.
13.
Improcerly Certified Licuf d Penetrant Examination Materials The CAi! informed the Atomic Safety and Licensing Board by a letter dated May 18, 1983, of a potential problem with the liquid penetrant meterials in 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 Dalla: 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
(
cn the date of the letter, it would seem that the problem arose between approximately May 8 to May 11,1983.
The situation bears clcse resemblance to the situation outlined bcginning 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 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 November 1982 and documented in NRC Inspection Report 50-446/82-11, the inspector found that previous corrective actions were not adequate and fur-ther that the " pan 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 Cecember 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 inspecticn pro-cedures had been implemented to prevent repetition.
NRC Inspection Report 50-445/83 10;50-446/83-05 docurented verification that the license *'s actions were acceptable and the matter was closed.
It accears that the situation outlined in the CASE letter ::arallels the NRC findings in all details except for the dates wnich crecaoly arose as a result of misunderstood or ince.mplete comunicatiens between the 1
1 1
14 F4gna-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 probl em?
Answer:
The roles of repor ability were effectively reversed in that the NRC identified the problem and notified the
- iicensee, i
A need for further NRC action on this matter has not been identified and l
the matter is considered closed.
14 Penetration Seals This special inspection was undertaken to ascertain the validity and sig-l nificance of allegations received initially by an NRC Headquarters. Duty l
Officer on or about March 22, 1983, which were confirmed and added to during l
a telephone interview with the alleger on Furch 23, 1913, 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.
l 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,,1egion IV l
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 l
production and quality procedures applicable to the work at CPSES as well l
as some that are not.
The alleger specifically mentioned that the NRC l
should review Procedures QC-507, SP-504, SP-505, SP-505-1, and SP-505-1 in regard to the flexible boot overlap problem.
Eacn of the acove procedures was in the backs offered to the SRIC for review.
A brief discussion fol-lows as to the contents of these procedures:
a.
OCP 507:
This procedure covers the final inscection of installed I
15 flexible coots.
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 formation 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 of the pipe and/or sleeve to which the boot will be attached.
Tnus, 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 skehed and the seam overlap will be something other than 2-inches.
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 oe spread over a 3-inch depth from the fabric edge pt-f or to fitting up the fabric where it is to be i.nstalled.
Although not so stated, it appears that the 3-inch width of adhesive is to provide sufficient area of adhesive in the event the above men-tioned cone skewing eccurs.
d.
SP-505-1 and SP-505-2: Tnese 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-tecticr/ 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 979 and that his homeoffice engin-eering had then changed the seal seam' detail. Tne 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-MS-3SF) recuires 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 hign that even a reduction of 1/3 in the area of the overlap woulc have the effect of changing the safety factor from 22:1 to approximately 1::1.
It is the SRIC's conclusion that while the allegation relative to tne reduction in seam from 3 to 2 inches is correct, the reduction woulc have no significant effect on the performance of the boot in service at CPSES and that, therefore,'the allegatjon has no technical merit.
!(
.~
16 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-fomance of the fabric in one application which is understood to involve the tearing of the fabric after being punctured.
It is understood that the 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 separation of the radiation seal aggregate material from the carrier material, the SRIC can only conclude that the al-l 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 SISCO procedures require i
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 Solicing 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 licensee has committed in FSAR Section 8.1 to comply with IEEE 420,
" Trial-Use Guide for Class 1E 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 i-order to ascertain whether they are acceptable -items, items of ncn-compliance, or deviations.
One such item, disclosed during the inspection, is discussed in paragraoh 15 above. Tnis item is identified as " Splicing of Electrical Cables in Cabinets."
(8324-01)
17
- 17. Manacement Interviews 4
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, i
)
l l
l
.